Multi-purpose deployer for ITER in-vessel maintenance

Energy Technology Data Exchange (ETDEWEB)

Choi, Chang-Hwan, E-mail: Chang-Hwan.CHOI@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Tesini, Alessandro; Subramanian, Rajendran [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Rolfe, Alan; Mills, Simon; Scott, Robin; Froud, Tim; Haist, Bernhard; McCarron, Eddie [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, OXON (United Kingdom)

2015-10-15

Highlights: • ITER RH system called as the multi-purpose deployer (MPD) is introduced. • The MPD performs dust and tritium inventory control, in-service inspection. • The MPD performs leak localization, in-vessel diagnostics maintenance. • The MPD has nine degrees of freedom with a payload capacity up to 2 tons. - Abstract: The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke

1999-02-01

An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

Distribution of the In-Vessel Diagnostics in ITER Tokamak

Energy Technology Data Exchange (ETDEWEB)

González, Jorge, E-mail: Jorge.Gonzalez@iter.org [Rüecker Lypsa, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Clough, Matthew; Martin, Alex; Woods, Nick; Suarez, Alejandro [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France); Martinez, Gonzalo [Technical University Of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Stefan, Gicquel; Yunxing, Ma [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France)

2017-01-15

The ITER In-Vessel Diagnostics have been distributed around the In-Vessel shell to understand burning plasma physics and assist in machine operation. Each diagnostics component has its own requirements, constraints, and even exclusion among them for the highly complex In-Vessel environment. The size of the plasma, the requirement to be able to align the blanket system to the magnetic centre of the machine, the cooling requirements of the blanket system and the size of the pressure vessel itself all add to the difficulties of integrating these systems into the remaining space available. The available space for the cables inside the special trays (in-Vessel looms) is another constraint to allocate In-Vessel electrical sensors. Besides this, there are issues with the Assembly sequences and surface & volumetric neutron heating considerations that have imposed several additional restrictions.

Progress and Achievements on the R&D Activities for ITER Vacuum Vessel

Energy Technology Data Exchange (ETDEWEB)

Nakahira, M. [Japan Atomic Energy Research Institute (JAERI); Koizumi, K. [Japan Atomic Energy Research Institute (JAERI); Takahashi, H. [Japan Atomic Energy Research Institute (JAERI); Onozuka, M. [ITER Joint Central Team, Garching, Germany; Ioki, K. [ITER Joint Central Team, Garching, Germany; Kuzumin, E. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Krylov, V. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Maslakowski, J. [Oak Ridge National Laboratory (ORNL); Nelson, Brad E [ORNL; Jones, L. [Max-Planck Institute, Garching, Germany; Danner, W. [Max-Planck Institute, Garching, Germany; Maisonnier, D. [Max-Planck Institute, Garching, Germany

2001-01-01

The ITER vacuum vessel (VV) is designed to be large double-walled structure with a D-shaped crosssection. The achievable fabrication tolerance of this structure was unknown due to the size and complexity of shape. The Full-scale Sector Model of ITER Vacuum Vessel, which was 15m in height, was fabricated and tested to obtain the fabrication and assembly tolerances. The model was fabricated within the target tolerance of 5mm and welding deformation during assembly operation was obtained. The port structure was also connected using remotized welding tools to demonstrate the basic maintenance activity. In parallel, the tests of advanced welding, cutting and inspection system were performed to improve the efficiency of fabrication and maintenance of the Vacuum Vessel. These activities show the feasibility of ITER Vacuum Vessel as feasible in a realistic way. This paper describes the major progress, achievement and latest status of the R&D activities on the ITER vacuum vessel.

Technical meeting on materials for in-vessel components of ITER

International Nuclear Information System (INIS)

Kalinin, G.; Barabash, V.

2000-01-01

The Technical meeting on materials for in-vessel components of ITER was held at the ITER Joint Work Site in Garching from 31 January to 4 February. The main objectives of the meetings were: 1. to summarize the requirements, 2. to review new data, 3. to discuss in detail the R and D program and to discuss the material assessment report

Investigation of vessel visibility of iterative reconstruction method in coronary computed tomography angiography using simulated vessel phantom

International Nuclear Information System (INIS)

Inoue, Takeshi; Uto, Fumiaki; Ichikawa, Katsuhiro; Hara, Takanori; Urikura, Atsushi; Hoshino, Takashi; Miura, Youhei; Terakawa, Syouichi

2012-01-01

Iterative reconstruction methods can reduce the noise of computed tomography (CT) images, which are expected to contribute to the reduction of patient dose CT examinations. The purpose of this study was to investigate impact of an iterative reconstruction method (iDose 4 , Philips Healthcare) on vessel visibility in coronary CT angiography (CTA) by using phantom studies. A simulated phantom was scanned by a CT system (iCT, Philips Healthcare), and the axial images were reconstructed by filtered back projection (FBP) and given a level of 1 to 7 (L1-L7) of the iterative reconstruction (IR). The vessel visibility was evaluated by a quantitative analysis using profiles across a 1.5-mm diameter simulated vessel as well as visual evaluation for multi planar reformation (MPR) images and volume rendering (VR) images in terms of the normalized-rank method with analysis of variance. The peak CT value of the profiles decreased with IR level and full width at half maximum of the profile also decreased with the IR level. For normalized-rank method, there was no statistical difference between FBP and L1 (20% dose reduction) for both MPR and VR images. The IR levels higher than L1 sacrificed the spatial resolution for the 1.5-mm simulated vessel, and their visual vessel visibilities were significantly inferior to that of the FBP. (author)

In-vessel tritium retention and removal in ITER-FEAT

International Nuclear Information System (INIS)

Federici, G.; Brooks, J.N.; Iseli, M.; Wu, C.H.

2001-01-01

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ∝350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

In-vessel tritium retention and removal in ITER-FEAT

Energy Technology Data Exchange (ETDEWEB)

Federici, G. [ITER Garching Joint Work Site, Garching (Germany); Brooks, J.N. [Argonne National Lab., IL (United States); Iseli, M. [ITER Naka Joint Work Site, Naka-gun (Japan); Wu, C.H. [EFDA Close Support Unit, Garching (Germany)

2001-07-01

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., {proportional_to}350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

In-Vessel Tritium Retention and Removal in ITER-FEAT

Science.gov (United States)

Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

ITER vacuum vessel structural analysis completion during manufacturing phase

Energy Technology Data Exchange (ETDEWEB)

Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F. [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Caixas, J.; Fernandez, E.; Zarzalejos, J.M. [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Kim, H.-S.; Kim, Y.G. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Privalova, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); and others

2016-11-01

Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

ITER vacuum vessel structural analysis completion during manufacturing phase

International Nuclear Information System (INIS)

Martinez, J.-M.; Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F.; Caixas, J.; Fernandez, E.; Zarzalejos, J.M.; Kim, H.-S.; Kim, Y.G.; Privalova, E.

2016-01-01

Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

ITER vacuum vessel, in vessel components and plasma facing materials

International Nuclear Information System (INIS)

Ioki, Kimihiro; Enoeda, M.; Federici, G.

2007-01-01

Design of the NB ports including duct liners under heat loads of the neutral beams has been developed. Design of the in-wall shielding has been developed in more details considering the supporting structure and the assembly method. The ferromagnetic inserts have previously not been installed in the outboard midplane region due to irregularity caused by the tangential ports for NB injection. Due to this configuration, the maximum ripple is relatively large (∝1 %) in a limited region of the plasma and the toroidal field flux lines fluctuate ∝10 mm in the FW region. To avoid these problems, additional ferromagnetic inserts are to be installed in the equatorial port region. Detailed studies were carried out on the ITER vacuum vessel to define appropriate codes and standards in the context of the ITER licensing in France. A set of draft documents regarding the ITER vacuum vessel structural code were prepared including an RCC-MR Addendum for the ITER VV with justified exceptions or modifications. The main deviation from the base Code is the extensive use of UT in lieu of radiography for the volumetric examination of all one-side access welds of the outer shell and field joint. The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. Development of new beryllium materials is progressing in China and Russia. The ITER limiters will be installed in equatorial ports at two toroidal locations. The limiter plasma-facing surface protrudes ∝8 cm from the FW during the start-up and shutdown phase. In the new limiter concept, the limiters are retracted by ∝8 cm during the plasma flat top phase. This concept gives important advantages; (i) mitigation of the particle and heat loads due to disruptions, ELMs and blobs, (ii) improvement of the power coupling with the ICRH antenna

Development of in-vessel neutron flux monitor equipped with microfission chambers to withstand the extreme ITER environment

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Masao, E-mail: ishikawa.masao@jaea.go.jp; Takeda, Keigo; Itami, Kiyoshi

2016-11-01

Highlights: • The in-vessel components of MFC system must withstand the extreme ITER environment. • To verify this, the thermal cycle test and the vibration tests were conducted. • Both tests were conducted under much severer conditions than ITER environment. • Soundness verification tests after the tests indicated that no problemswere found. • It is shown that the in-vessel component is sufficiently robust ITER environment. - Abstract: Via thermal cycling and vibration tests, this study aims to demonstrate that the in-vessel components of the microfission chamber (MFC) system can withstand the extreme International Thermonuclear Experimental Reactor (ITER) environment. In thermal cycle tests, the signal cable of the device was bent into a smaller radius and it was given more bends than those in its actual configuration within ITER. A faster rate of temperature change than that under the typical ITER baking scenario was then imposed on in-vessel components. For the vibration tests, strong 10 G vibrational accelerations with frequencies ranging from 30 Hz to 2000 Hz were imposed to the detector and the connector of the in-vessel components to simulate various types of electromagnetic events. Soundness verification tests of the in-vessel components conducted after thermal cycling and vibration testing indicated that problems related to the signal transmission cable functioning were not found. Thus, it was demonstrated that the in-vessel components of the MFC can withstand the extreme environment within ITER.

In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

Energy Technology Data Exchange (ETDEWEB)

L. C. Cadwallader

2013-01-01

The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

Design and issues of the ITER in-vessel components: ITER Joint central team and home teams

International Nuclear Information System (INIS)

Parker, R.R.

1998-01-01

This paper surveys the status of the design of the in-vessel components for ITER, in particular the major components, namely the vacuum vessel, blanket and first wall, and divertor, and the interface of selected ancillary systems such as those used for RF heating and current drive, and for diagnostics. The vacuum vessel is a double-walled structure constructed from two toroidal shells joined by ribs. The space between the skins is filled with shield plates directly cooled by water. The structural material is 316 LN IG (ITER grade). Toroidal supports joining the vessel midplane ports with the TF structure limit possible differential toroidal displacements, as might occur due to seismic or vertical displacement events (VDEs). A variety of load conditions corresponding to normal and off-normal loads have been considered and in all cases peak vessel stresses are within allowables. The blanket system consists of approximately 700 modules, each weighing ∝4 t. The integrated first wall consists of a beryllium-tiled copper mat bonded to the water-cooled SS shield block. The copper mat functions as a heat sink and has imbedded in it an array of SS tubes providing water cooling. The modules are mechanically attached to a toroidal backplate. Loads due to centered disruptions are reacted via hoop stress in the backplate, whereas net vertical and horizontal loads such as those arising from VDEs are transferred through the backplate and divertor supports to the vessel. (orig.)

Computed tomography depiction of small pediatric vessels with model-based iterative reconstruction

Energy Technology Data Exchange (ETDEWEB)

Koc, Gonca; Courtier, Jesse L.; Phelps, Andrew; Marcovici, Peter A.; MacKenzie, John D. [UCSF Benioff Children' s Hospital, Department of Radiology and Biomedical Imaging, San Francisco, CA (United States)

2014-07-15

Computed tomography (CT) is extremely important in characterizing blood vessel anatomy and vascular lesions in children. Recent advances in CT reconstruction technology hold promise for improved image quality and also reductions in radiation dose. This report evaluates potential improvements in image quality for the depiction of small pediatric vessels with model-based iterative reconstruction (Veo trademark), a technique developed to improve image quality and reduce noise. To evaluate Veo trademark as an improved method when compared to adaptive statistical iterative reconstruction (ASIR trademark) for the depiction of small vessels on pediatric CT. Seventeen patients (mean age: 3.4 years, range: 2 days to 10.0 years; 6 girls, 11 boys) underwent contrast-enhanced CT examinations of the chest and abdomen in this HIPAA compliant and institutional review board approved study. Raw data were reconstructed into separate image datasets using Veo trademark and ASIR trademark algorithms (GE Medical Systems, Milwaukee, WI). Four blinded radiologists subjectively evaluated image quality. The pulmonary, hepatic, splenic and renal arteries were evaluated for the length and number of branches depicted. Datasets were compared with parametric and non-parametric statistical tests. Readers stated a preference for Veo trademark over ASIR trademark images when subjectively evaluating image quality criteria for vessel definition, image noise and resolution of small anatomical structures. The mean image noise in the aorta and fat was significantly less for Veo trademark vs. ASIR trademark reconstructed images. Quantitative measurements of mean vessel lengths and number of branches vessels delineated were significantly different for Veo trademark and ASIR trademark images. Veo trademark consistently showed more of the vessel anatomy: longer vessel length and more branching vessels. When compared to the more established adaptive statistical iterative reconstruction algorithm, model

Computed tomography depiction of small pediatric vessels with model-based iterative reconstruction

International Nuclear Information System (INIS)

Koc, Gonca; Courtier, Jesse L.; Phelps, Andrew; Marcovici, Peter A.; MacKenzie, John D.

2014-01-01

Computed tomography (CT) is extremely important in characterizing blood vessel anatomy and vascular lesions in children. Recent advances in CT reconstruction technology hold promise for improved image quality and also reductions in radiation dose. This report evaluates potential improvements in image quality for the depiction of small pediatric vessels with model-based iterative reconstruction (Veo trademark), a technique developed to improve image quality and reduce noise. To evaluate Veo trademark as an improved method when compared to adaptive statistical iterative reconstruction (ASIR trademark) for the depiction of small vessels on pediatric CT. Seventeen patients (mean age: 3.4 years, range: 2 days to 10.0 years; 6 girls, 11 boys) underwent contrast-enhanced CT examinations of the chest and abdomen in this HIPAA compliant and institutional review board approved study. Raw data were reconstructed into separate image datasets using Veo trademark and ASIR trademark algorithms (GE Medical Systems, Milwaukee, WI). Four blinded radiologists subjectively evaluated image quality. The pulmonary, hepatic, splenic and renal arteries were evaluated for the length and number of branches depicted. Datasets were compared with parametric and non-parametric statistical tests. Readers stated a preference for Veo trademark over ASIR trademark images when subjectively evaluating image quality criteria for vessel definition, image noise and resolution of small anatomical structures. The mean image noise in the aorta and fat was significantly less for Veo trademark vs. ASIR trademark reconstructed images. Quantitative measurements of mean vessel lengths and number of branches vessels delineated were significantly different for Veo trademark and ASIR trademark images. Veo trademark consistently showed more of the vessel anatomy: longer vessel length and more branching vessels. When compared to the more established adaptive statistical iterative reconstruction algorithm, model

Project management techniques used in the European Vacuum Vessel sectors procurement for ITER

Energy Technology Data Exchange (ETDEWEB)

Losasso, Marcello, E-mail: marcello.losasso@f4e.europa.eu [Fusion for Energy (F4E), Barcelona (Spain); Ortiz de Zuniga, Maria; Jones, Lawrence; Bayon, Angel; Arbogast, Jean-Francois; Caixas, Joan; Fernandez, Jose; Galvan, Stefano; Jover, Teresa [Fusion for Energy (F4E), Barcelona (Spain); Ioki, Kimihiro [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lewczanin, Michal; Mico, Gonzalo; Pacheco, Jose Miguel [Fusion for Energy (F4E), Barcelona (Spain); Preble, Joseph [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Stamos, Vassilis; Trentea, Alexandru [Fusion for Energy (F4E), Barcelona (Spain)

2012-08-15

Highlights: Black-Right-Pointing-Pointer File name contains the directory tree structure with a string of three-letter acronyms, thereby enabling parent directory location when confronted with orphan files. Black-Right-Pointing-Pointer The management of the procurement procedure was carried out in an efficient and timely manner, achieving precisely the contract placement date foreseen at the start of the process. Black-Right-Pointing-Pointer The contract start-up has been effectively implemented and a flexible project management system has been put in place for an efficient monitoring of the contract. - Abstract: The contract for the seven European Sectors of the ITER Vacuum Vessel (VV) was placed at the end of 2010 with a consortium of three Italian companies. The task of placing and the initial take-off of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. A stringent quality controlled system with a bespoke Vacuum Vessel Project Lifecycle Management system to control the information flow, based on ENOVIA SmarTeam, was developed to handle the storage and approval of Documentation including links to the F4E Vacuum Vessel system and ITER International Organization System interfaces. The VV Sector design and manufacturing schedule is based on Primavera software, which is cost loaded thus allowing F4E to carry out performance measurement with respect to its payments and commitments. This schedule is then integrated into the overall Vacuum Vessel schedule, which includes ancillary activities such as instruments, preliminary design and analysis. The VV Sector Risk Management included three separate risk analyses from F4E and the bidders, utilizing two different methodologies. These efforts will lead to an efficient and effective implementation of this contract, vital to the success of the ITER machine, since the Vacuum Vessel is the biggest single work package of Europe's contribution to ITER and

Design of the ITER vacuum vessel

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

1995-01-01

The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement.The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2-2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4-0.7 m.The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200-300 MN due to VDE and to allow for differential thermal expansion.The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology. (orig.)

Design and development of ITER high-frequency magnetic sensor

NARCIS (Netherlands)

Ma, Y.; Vayakis, G.; Begrambekov, L. B.; Cooper, J.J.; Duran, I.; Hirsch, M.; Laqua, H.P.; Moreau, Ph.; Oosterbeek, J.W.; Spuig, P.; Stange, T.; Walsh, M.

2016-01-01

High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in

MELCOR ex-vessel LOCA simulations for ITER+

International Nuclear Information System (INIS)

Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

1995-01-01

Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H

2001-11-01

Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

International Nuclear Information System (INIS)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

2001-01-01

Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

Design and development of the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Koizumi, K.; Nakahira, M.; Itou, Y.; Tada, E. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan); Johnson, G.; Ioki, K.; Elio, F.; Iizuka, T.; Sannazzaro, G.; Takahashi, K.; Utin, Y.; Onozuka, M. [ITER Joint Central Team (JCT), Garching (Germany); Nelson, B. [US Home Team, Oak Ridge National Laboratory (United States); Vallone, C. [EU Home Team, NET Team, Garching (Germany); Kuzmin, E. [RF Home Team, Efremov Institute, City (Russian Federation)

1998-09-01

In ITER, the vacuum vessel (VV) is designed to be a water cooled, double-walled toroidal structure made of 316LN stainless steel with a D-shaped cross section approximately 9 m wide and 15 m high. The design work which began at the beginning of the ITER-EDA is nearing completion by resolving the technical issues. In parallel with the design activities, the R and D program, full-scale VV sector model project, was initiated in 1995 to resolve the design and fabrication issues. The full-scale sector model corresponds to an 18 sector (9 sub-sector x 2) and is being fabricated on schedule. To date, 60% of the fabrication had been completed. The fabrication of full-scale model including sector-to-sector connection will be completed by the end of 1997 and performance tests are scheduled until the end of ITER-EDA. This paper describes the latest status of the ITER VV design and the full-scale sector model project. (orig.) 3 refs.

Conceptual design studies of in-vessel viewing equipment for ITER

International Nuclear Information System (INIS)

Obara, Kenjiro; Oka, Kiyoshi; Taguchi, Hiroshi; Itoh, Akira; Tada, Eisuke; Shibanuma, Kiyoshi

1996-03-01

In-vessel viewing systems are essential to inspect all surface of in-vessel components so as to detect and locate damages, and to assist in-vessel maintenance operations. The in-vessel viewing operations are categorized into the three cases, which are 1) rapid inspection just after off-normal events such as disruption, 2) scheduled inspection, and 3) supplementary inspection during maintenance operations. In case of the rapid inspection, the viewing systems have to be operated in vacuum (ca. 10 -5 Pa) and high temperature (ca. 300degC) under a gamma ray dose rate of 10 7 R/h. On the other hand, the latter two cases are anticipated to be under atmospheric inert gas, 150degC and 3x10 6 R/h. Accordingly, the in-vessel viewing systems are required to have sufficient durability under those conditions of all cases as well as precision of the vision to all of in-vessel surface. Based on those requirements, scoping studies on various viewing concepts have been performed and the applicability to the ITER conditions have been assessed. As a result, two types of viewing systems have been chosen, which are a periscope type viewing system and a image fiber type viewing system with a multi-joint manipulator. Both systems are based on radiation hard optical elements which are being developed. In this report, the design features of both viewing systems are described, including technical issues for ITER application. Finally, a periscope type viewing system is recommended as a primary system and the following specifications/conditions are proposed for the further engineering design. (1) Unified type periscope with a movable mirror at the tip (2) Integrated lighting device into the periscope (3) Accessed from top vertical ports located at 7.3m from the machine center (4) Proposed configuration with a total length of around 27m and a diameter of 200mm. (author)

Iteration scheme for implicit calculations of kinetic and equilibrium chemical reactions in fluid dynamics

International Nuclear Information System (INIS)

Ramshaw, J.D.; Chang, C.H.

1995-01-01

An iteration scheme for the implicit treatment of equilibrium chemical reactions in partial equilibrium flow has previously been described. Here we generalize this scheme to kinetic reactions as well as equilibrium reactions. This extends the applicability of the scheme to problems with kinetic reactions that are fast in regions of the flow field but slow in others. The resulting scheme thereby provides a single unified framework for the implicit treatment of an arbitrary number of coupled equilibrium and kinetic reactions in chemically reacting fluid flow. 10 refs., 2 figs

Structural materials for ITER in-vessel component design

Energy Technology Data Exchange (ETDEWEB)

Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

1996-10-01

The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

Assessment and selection of materials for ITER in-vessel components

Science.gov (United States)

Kalinin, G.; Barabash, V.; Cardella, A.; Dietz, J.; Ioki, K.; Matera, R.; Santoro, R. T.; Tivey, R.; ITER Home Teams

2000-12-01

During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)-IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti-6Al-4V alloy and two copper alloys, CuCrZr-IG and CuAl25-IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

Progress of ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Ioki, K., E-mail: Kimihiro.Ioki@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bayon, A. [F4E, c/ Josep Pla, No. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Le Barbier, R.; Martinez, J.-M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Pathak, H. [ITER-India, A-29, GIDC Electronic Estate, Sector 25, Gandhinagar 382025 (India); Preble, J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sa, J.W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Terasawa, A.; Utin, Yu. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); and others

2013-10-15

Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.

Progress of ITER vacuum vessel

International Nuclear Information System (INIS)

Ioki, K.; Bayon, A.; Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B.; Kim, B.C.; Kuzmin, E.; Le Barbier, R.; Martinez, J.-M.; Pathak, H.; Preble, J.; Sa, J.W.; Terasawa, A.; Utin, Yu.

2013-01-01

Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure

Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

Energy Technology Data Exchange (ETDEWEB)

Testoni, P., E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy, Josep Plá n. 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Albanese, R. [Association Euratom/ENEA/CREATE, DIEL, Università Federico II di Napoli, Napoli 80125 (Italy); Lucca, F.; Roccella, M. [L.T. Calcoli S.a.S. Piazza Prinetti, 26/B, Merate, Lecco (Italy); Portone, A. [Fusion for Energy, Josep Plá n. 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Rubinacci, G. [Association Euratom/ENEA/CREATE, DIEL, Università Federico II di Napoli, Napoli 80125 (Italy); Ventre, S.; Villone, F. [Association Euratom/ENEA/CREATE, DAEIMI, Università di Cassino, Cassino 03043 (Italy)

2013-10-15

Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained.

Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

International Nuclear Information System (INIS)

Testoni, P.; Albanese, R.; Lucca, F.; Roccella, M.; Portone, A.; Rubinacci, G.; Ventre, S.; Villone, F.

2013-01-01

Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained

In-vessel dust and tritium control strategy in ITER

Energy Technology Data Exchange (ETDEWEB)

Shimada, M., E-mail: michiya.shimada@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France); Pitts, R.A.; Ciattaglia, S.; Carpentier, S.; Choi, C.H.; Dell Orco, G.; Hirai, T.; Kukushkin, A.; Lisgo, S.; Palmer, J.; Shu, W.; Veshchev, E. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

2013-07-15

A baseline strategy for dust and tritium-inventory control and recovery in ITER has been established and preparations are underway for its implementation. Limits on dust and tritium-inventory are an integral part of the ITER safety case and are fixed at 1 kg for tritium, 1000 kg for mobilisable dust and 11 kg (beryllium)/76 kg (tungsten) for dust on hot surfaces. Maximum average T-retention rates of ∼1 g/shot are estimated for baseline inductive operation at Q{sub DT} = 10, suggesting that the in-vessel T-retention could reach the administrative limit of 640 g in as little as ∼2 months of operation. Baking is expected to remove a significant fraction of the T co-deposited on the divertor targets. Despite large uncertainties, dust quantities are expected to remain well below safety limits over the divertor cassette lifetime. In situ aspiration during divertor cassette exchange is foreseen as the main dust removal technique.

ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

International Nuclear Information System (INIS)

1995-01-01

ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers ampersand Constructors and Chicago Bridge ampersand Iron (Raytheon/CB ampersand I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB ampersand I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB ampersand I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB ampersand I and documented accordingly

Simulation of LLCB TBM in-vessel first wall coolant break into ITER vacuum vessel by using RELAP/MOD3.4

International Nuclear Information System (INIS)

Tony Sandeep, K.; Chaudhari, Vilas; Rajendra Kumar, E.; Dutta, Anu; Singh, R.K.

2013-06-01

To prove Test Blanket Module (TBM) safety in International Thermonuclear Experimental Reactor (ITER), various accident scenarios are postulated . One of the postulated initiating events to be analysed is TBM First wall (FW) coolant leak in ITER Vacuum vessel (VV). This accident has been classified as a reference event for the TBM (probability of occurrence >1 E -06 /a). The postulated accident occurs as a result of small leak of TBM FW helium into ITER vacuum vessel (VV), caused by the TBM weld failure. The ingress of this TBM FW helium into ITER plasma induces intense plasma disruption that deposits 1.8 MJ/m 2 of plasma stored thermal energy onto the TBM FW over a period of 1 sec in duration (assumption). Runaway electrons in this process are lost from plasma current channel and cause multiple TBM and ITER FW cooling tube failures within 10 cm torriodal strip. The size of the break is identified as double ended rupture of all coolant channels within this strip around the reactor. For LLCB TBM this represents failure of 4 FW channels. The size of ITER FW break is 0.02 m 2 . Consequently, a simultaneous blow down of TBM FW helium and ITER FW water occurs, injecting helium and water into VV. This pressurisation causes the activation of VV pressure suppressions system and ingress of water into VV. This pressurisation causes the VV pressure suppressions system (VVPSS) to open in an attempt to contain the pressure below the safety limit of 0.2 MPa. This report is intended to do the above accident analysis and assessment of active components of TBM using RELAP code and hence prove its safety in ITER environment. (author)

ITER in-vessel system design and performance

International Nuclear Information System (INIS)

Parker, R.R.

1999-01-01

This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

ITER in-vessel system design and performance

International Nuclear Information System (INIS)

Parker, R.R.

2001-01-01

This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

Vacuum vessel port structures for ITER-FEAT

International Nuclear Information System (INIS)

Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th.

2001-01-01

The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed

Vacuum vessel port structures for ITER-FEAT

Energy Technology Data Exchange (ETDEWEB)

Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th

2001-11-01

The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed.

Structural analysis of the ITER vacuum vessel from disruption loading with halo asymmetry

International Nuclear Information System (INIS)

Riemer, B.W.; Sayer, R.O.

1996-01-01

Static structural analyses of the ITER vacuum vessel were performed with toroidally asymmetric disruption loads. Asymmetric halo current conditions were assumed to modify symmetric disruption loads which resulted in net lateral loading on the vacuum vessel torus. Structural analyses with the asymmetric loading indicated significantly higher vessel stress and blanket support forces than with symmetric disruption loads. A recent change in the vessel support design which provided toroidal constraints at each mid port was found to be effective in reducing torus lateral movement and vessel stress

Clearance potential of ITER vacuum vessel activated materials

International Nuclear Information System (INIS)

Cepraga, D.G.; Cambi, G.; Frisoni, M.

2002-01-01

To demonstrate fusion's environmental attractiveness over the entire life cycle, a waste analysis is mandatory. The clearance is recommended by IAEA for releasing activated solid materials from regulatory control and for waste management policy. The paper focuses on the approach used to support waste analyses for ITER Generic Site Safety Report. The Material Unconditional Clearance Index of all the materials/zones on the equatorial mid-plane of ITER machine have been evaluated, based on IAEA-TECDOC-855. The Bonami-Nitawl-XSDNRPM sequence of the Scale-4.4a code system (using Vitenea-J library) has been firstly used for radiation transport analyses. Then the Anita-2000 code package is used for the activation calculation. The paper presents also, as an example, an application of the clearance indexes estimation for the ITER vacuum vessel materials. The results of the Anita-2000 have been compared with those obtained using the Fispact-99 activation code. (author)

A New Iterative Method for Equilibrium Problems and Fixed Point Problems

Directory of Open Access Journals (Sweden)

Abdul Latif

2013-01-01

Full Text Available Introducing a new iterative method, we study the existence of a common element of the set of solutions of equilibrium problems for a family of monotone, Lipschitz-type continuous mappings and the sets of fixed points of two nonexpansive semigroups in a real Hilbert space. We establish strong convergence theorems of the new iterative method for the solution of the variational inequality problem which is the optimality condition for the minimization problem. Our results improve and generalize the corresponding recent results of Anh (2012, Cianciaruso et al. (2010, and many others.

Conceptual design finalisation of the ITER In-Vessel Viewing and Metrology System (IVVS)

Energy Technology Data Exchange (ETDEWEB)

Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Damiani, Carlo; Van Uffelen, Marco; Lo Bue, Alessandro; Izquierdo, Jesus; Semeraro, Luigi [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Martins, Jean-Pierre; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components. Periodically or on request, the IVVS probes will be deployed into the Vacuum Vessel from their storage positions (still within the ITER primary confinement) in order to perform both viewing and metrology on plasma facing components (blanket, divertor, heating/diagnostic plugs, test blanket modules) and, more generically, to provide information on the status of the in-vessel components. In 2011, the IO proposed to simplify and strengthen the six IVVS port extensions situated at the divertor level. Among other important consequences, such as the relocation of the Glow Discharge Cleaning (GDC) electrodes at other levels of the machine, this major design change implied the need for a substantial redesign of the IVVS plug, which took part to an on-going effort to bring the integrated IVVS concept – including the scanning probe and its deployment system – to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the various design and R and D activities in progress: plug design integration, probe concept validation under environmental conditions, development of a metrology strategy, the whole supported by a nuclear analysis.

ITER in-vessel system design and performance

Science.gov (United States)

Parker, R. R.

2000-03-01

The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events.

ITER in-vessel system design and performance

International Nuclear Information System (INIS)

Parker, R.R.

2000-01-01

The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events. (author)

ITER diagnostics ex-vessel engineering services

Energy Technology Data Exchange (ETDEWEB)

Arumugam, A.P., E-mail: arun.prakash@iter.org; Walker, C.I.; Andrew, P.; Barnsley, R.; Beltran, D.; Bertalot, L.; Dammann, A.; Direz, M.F.; Drevon, J.M.; Encheva, A.; Giacomin, T.; Hourtoule, J.; Kuehn, I.; Lanza, R.; Levesy, B.; Maquet, P.; Patel, K.M.; Patisson, L.; Pitcher, C.S.; Portales, M.; and others

2013-10-15

Highlights: • This paper describes about the ITER diagnostics ex-vessel engineering services. • It describes various diagnostics systems, its location and its environment. • Diagnostics interfaces with other services such as the buildings, HVAC, electrical services, cooling water, vacuum, liquid and gas distribution. • All the interfaces with these services are identified and defined. • Buildings services for diagnostics, such as penetrations, local shielding, embedment and temperature control are discussed. -- Abstract: Extensive diagnostics systems will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include measurements of temperature, density, impurity concentration, and particle and energy confinement times. ITER diagnostic systems extend from the center of the Tokamak to the various diagnostic areas, where they are controlled and acquired data is processed. This mainly includes the areas such as ports, port cells, gallery, diagnostics enclosures and cubicle areas. The diagnostics port plugs encloses the front end of the diagnostic systems and the diagnostics building houses the diagnostics equipment, instrumentation and control cubicles. There are several systems providing services to diagnostics. These mainly include ITER buildings, electrical power services, cooling water services, Heating Ventilation and Air Conditioning (HVAC), vacuum services, liquid and gas distribution services, cable engineering, de-tritiation systems, control cubicles, etc. Requirements of these service systems have to be defined, even though many of the diagnostics are at an early stage of development. It is a real challenge to define and to design diagnostics systems considering the constraints imposed by these service systems. This paper summarizes the provision of these services to the individual diagnostics and diagnostics areas

ITER diagnostics ex-vessel engineering services

International Nuclear Information System (INIS)

Arumugam, A.P.; Walker, C.I.; Andrew, P.; Barnsley, R.; Beltran, D.; Bertalot, L.; Dammann, A.; Direz, M.F.; Drevon, J.M.; Encheva, A.; Giacomin, T.; Hourtoule, J.; Kuehn, I.; Lanza, R.; Levesy, B.; Maquet, P.; Patel, K.M.; Patisson, L.; Pitcher, C.S.; Portales, M.

2013-01-01

Highlights: • This paper describes about the ITER diagnostics ex-vessel engineering services. • It describes various diagnostics systems, its location and its environment. • Diagnostics interfaces with other services such as the buildings, HVAC, electrical services, cooling water, vacuum, liquid and gas distribution. • All the interfaces with these services are identified and defined. • Buildings services for diagnostics, such as penetrations, local shielding, embedment and temperature control are discussed. -- Abstract: Extensive diagnostics systems will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include measurements of temperature, density, impurity concentration, and particle and energy confinement times. ITER diagnostic systems extend from the center of the Tokamak to the various diagnostic areas, where they are controlled and acquired data is processed. This mainly includes the areas such as ports, port cells, gallery, diagnostics enclosures and cubicle areas. The diagnostics port plugs encloses the front end of the diagnostic systems and the diagnostics building houses the diagnostics equipment, instrumentation and control cubicles. There are several systems providing services to diagnostics. These mainly include ITER buildings, electrical power services, cooling water services, Heating Ventilation and Air Conditioning (HVAC), vacuum services, liquid and gas distribution services, cable engineering, de-tritiation systems, control cubicles, etc. Requirements of these service systems have to be defined, even though many of the diagnostics are at an early stage of development. It is a real challenge to define and to design diagnostics systems considering the constraints imposed by these service systems. This paper summarizes the provision of these services to the individual diagnostics and diagnostics areas

Assembly of the sectors and ports of the ITER vacuum vessel

International Nuclear Information System (INIS)

Corino, S.; Moreno, R.

2014-01-01

The International Thermonuclear Experimental Reactor, ITER is a very complex Project that aims to prove the technical reliability of nuclear fusion. ITER has been Ensa's commitment to the future to strengthen as one of the main manufacturers of big equipment and services internationally in the nuclear field. Ensa started working on the qualification process to be able to bid for the 'Assembly of the ITER vacuum vessel' in June 2010, after two and a half years of pre-qualification, offers, clarifications and long technical meetings, that were followed by commercial meetings Ensa achieved its goal. The 30 of November, Ensa signed what at that time was the biggest of the supplies signed by IO (ITER Organization). A lot of efforts and hard work had been done in order to achieve this goal, but the hardest of all was yet to come, after the signature of the contract, Ensa has 7 years ahead to achieve the final goal, the assembly and welding of the 9 sectors that put together the ITER vacuum vessel and the 54 ports that will allow the assembly of the different auxiliary systems. The scope of the works to be performed can generally be divided into the following areas: - Welding of the sectors and ports; - Non-destructive tests; - Machining; - Dimensional Controls. In order to achieve this goal, the project has been divided into 3 different phases. - Development phase: January 2013 - July 2015; - Pre-production phase: July 2015 - February 2016; - Production phase: February 2016 - February 2020

Design of parallel intersector weld/cut robot for machining processes in ITER vacuum vessel

International Nuclear Information System (INIS)

Wu Huapeng; Handroos, Heikki; Kovanen, Janne; Rouvinen, Asko; Hannukainen, Petri; Saira, Tanja; Jones, Lawrence

2003-01-01

This paper presents a new parallel robot Penta-WH, which has five degrees of freedom driven by hydraulic cylinders. The manipulator has a large, singularity-free workspace and high stiffness and it acts as a transport device for welding, machining and inspection end-effectors inside the ITER vacuum vessel. The presented kinematic structure of a parallel robot is particularly suitable for the ITER environment. Analysis of the machining process for ITER, such as the machining methods and forces are given, and the kinematic analyses, such as workspace and force capacity are discussed

ITER vacuum vessel design (D201 subtask 1.3 and subtask 3). Final report

International Nuclear Information System (INIS)

1996-01-01

ITER Task No. D201, Vacuum Vessel Design (Subtask 1.3 and Subtask 3), was initiated to propose and evaluate local vacuum vessel reinforcement alternatives in proximity to the Neutral Beam, Radial Mid-Plane, Top, and Divertor Ports. These areas were reported to be highly stressed regions based on the results of preliminary stress analyses performed by the USHT (US Home Team) and the ITER Joint Central Team (JCT) at the Garching JWS (Joint Work Site). Initial design activities focused on the divertor port region which was reported to experience the highest stress intensities. Existing stress analysis models and results were reviewed with the USHT stress analysts to obtain an overall understanding of the vessel response to the various applied loads. These reviews indicated that the reported stress intensities in the divertor port region were significantly affected by the loads applied to the vessel in adjacent regions

Hydrogen/hydrocarbon explosions in the ITER vacuum vessel

International Nuclear Information System (INIS)

Goranson, P.L.

1992-01-01

The consequences of H 2 /hydrocarbon detonations in the vacuum vessel (torus) of the International Thermonuclear Experimental Reactor (ITER) have been studied. The most likely scenario for such a detonation involves a water leak into the torus and a vent of the torus to atmosphere, permitting the formation of an explosive fuel-air mixture. The generation of fuel gases and possible sources of air or oxygen are reviewed, and the severity and effects of specific fuel-air mixture explosions are evaluated. Detonation or deflagration of an explosive mixture could result in pressures exceeding the maximum allowable torus pressure. Further studies to examine the design details and develop an event-tree study of events following a gas detonation are recommended

Progress and achievements of R&D activities for the ITER vacuum vessel

Science.gov (United States)

Nakahira, M.; Takahashi, H.; Koizumi, K.; Onozuka, M.; Ioki, K.

2001-04-01

The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R&D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18° toroidal sector, which is composed of two 9° sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of +/-3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R&D activities for the ITER vacuum vessel.

Development of a Remote Handling Robot for the Maintenance of an ITER-Like D-Shaped Vessel

Directory of Open Access Journals (Sweden)

Peihua Chen

2014-01-01

Full Text Available Robotic operation is one of the major challenges in the remote maintenance of ITER vacuum vessel (VV and future fusion reactors as inner operations of Tokamak have to be done by robots due to the internal adverse conditions. This paper introduces a novel remote handling robot (RHR for the maintenance of ITER-like D-shaped vessel. The modular designed RHR, which is an important part of the remote handling system for ITER, consists of three parts: an omnidirectional transfer vehicle (OTV, a planar articulated arm (PAA, and an articulated teleoperated manipulator (ATM. The task of RHR is to carry processing tools, such as the viewing system, leakage detector, and electric screwdriver, to inspect and maintain the components installed inside the D-shaped vessel. The kinematics of the OTV, as well as the kinematic analyses of the PAA and ATM, is studied in this paper. Because of its special length and heavy payload, the dynamics of the PAA is also investigated through a dynamic simulation system based on robot technology middleware (RTM. The results of the path planning, workspace simulations, and dynamic simulation indicate that the RHR has good mobility together with satisfying kinematic and dynamic performances and can well accomplish its maintenance tasks in the ITER-like D-shaped vessel.

Engineering analysis of ITER In-Vessel Viewing System guide tube

Energy Technology Data Exchange (ETDEWEB)

Casal, Natalia, E-mail: natalia.casal@iter.org [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Bates, Philip [Fusion for Energy, Barcelona (Spain); Bede, Ottó [Oxford Technologies Ltd., Abingdon (United Kingdom); Damiani, Carlo; Dubus, Gregory [Fusion for Energy, Barcelona (Spain); Omran, Hassan [Oxford Technologies Ltd., Abingdon (United Kingdom); Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Puiu, Adrian [Fusion for Energy, Barcelona (Spain); Reichle, Roger; Suárez, Alejandro; Walker, Christopher; Walsh, Michael [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France)

2015-10-15

Highlights: • Conceptual design of IVVS Loads action on IVVS Dominant loads. • Seismic and baking conditions. • No active cooling needed for IVVS. • IVVS requires additional support points to avoid excessive deformation. - Abstract: The In Vessel Viewing System (IVVS) will be one of the essential machine diagnostic systems at ITER to provide information about the status of in-vessel and plasma facing components and to evaluate the dust inside the Vacuum Vessel. The current design consists of six scanning probes and their deployment systems, which are placed in dedicated ports at the divertor level. These units are located in resident guiding tubes 10 m long, which allow the IVVS probes to go from their storage location to the scanning position by means of a simple straight translation. Moreover, each resident tube is supported inside the corresponding Vacuum Vessel and Cryostat port extensions, which are part of the primary confinement barrier. As the Vacuum Vessel and the Cryostat will move with respect to each other during operation (especially during baking) and during incidents and accidents (disruptions, vertical displacement events, seismic events), the structural integrity of the resident tube and the surrounding vacuum boundaries would be compromised if the required flexibility and supports are not appropriately assured. This paper focuses on the integration of the present design of the IVVS into the Vacuum Vessel and Cryostat environment. It presents the adopted strategy to withstand all the main interfacing loads without damaging the confinement barriers and the corresponding analysis supporting it.

A superlinear iteration method for calculation of finite length journal bearing's static equilibrium position

Science.gov (United States)

Zhou, Wenjie; Wei, Xuesong; Wang, Leqin; Wu, Guangkuan

2017-05-01

Solving the static equilibrium position is one of the most important parts of dynamic coefficients calculation and further coupled calculation of rotor system. The main contribution of this study is testing the superlinear iteration convergence method-twofold secant method, for the determination of the static equilibrium position of journal bearing with finite length. Essentially, the Reynolds equation for stable motion is solved by the finite difference method and the inner pressure is obtained by the successive over-relaxation iterative method reinforced by the compound Simpson quadrature formula. The accuracy and efficiency of the twofold secant method are higher in comparison with the secant method and dichotomy. The total number of iterative steps required for the twofold secant method are about one-third of the secant method and less than one-eighth of dichotomy for the same equilibrium position. The calculations for equilibrium position and pressure distribution for different bearing length, clearance and rotating speed were done. In the results, the eccentricity presents linear inverse proportional relationship to the attitude angle. The influence of the bearing length, clearance and bearing radius on the load-carrying capacity was also investigated. The results illustrate that larger bearing length, larger radius and smaller clearance are good for the load-carrying capacity of journal bearing. The application of the twofold secant method can greatly reduce the computational time for calculation of the dynamic coefficients and dynamic characteristics of rotor-bearing system with a journal bearing of finite length.

FW/Blanket and vacuum vessel for RTO/RC ITER

International Nuclear Information System (INIS)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M.

2000-01-01

The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste

FW/Blanket and vacuum vessel for RTO/RC ITER

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M

2000-11-01

The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, {approx}50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

Integration test of ITER full-scale vacuum vessel sector

International Nuclear Information System (INIS)

Nakahira, M.; Koizumi, K.; Oka, K.

1999-01-01

The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven ITER R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of - 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

ITER cryostat main chamber and vacuum vessel pressure suppression system design

International Nuclear Information System (INIS)

Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

1999-03-01

Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

ITER cryostat main chamber and vacuum vessel pressure suppression system design

Energy Technology Data Exchange (ETDEWEB)

Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

1999-03-01

Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

Progress and achievements of R and D activities for the ITER vacuum vessel

International Nuclear Information System (INIS)

Nakahira, M.; Takahashi, H.; Koizumi, K.; Onozuka, M.; Ioki, K.

2001-01-01

The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R and D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18 deg. toroidal sector, which is composed of two 9 deg. sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of ±3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R and D activities for the ITER vacuum vessel. (author)

IWR-solution for the ITER vacuum vessel assembly

Energy Technology Data Exchange (ETDEWEB)

Wu, H., E-mail: huapeng@lut.fi [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Handroos, H. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Pela, P. [Tekes (Finland); Wang, Y. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland)

2011-10-15

The assembly of ITER vacuum vessel (VV) is still a very big challenge as the process can only be done from inside the VV. The welding of the VV assembly is carried out using the dedicated robotic systems. The main functions of the robots are: (i) measuring the actual space between every two sectors, (ii) positioning of the 150 kg splice plates between the sector shells, (iii) welding the splice plates to the sector shells, (iv) NDT of the welds, (v) repairing, including machining of the welds, (vi) He-leak tests of the welds, and (vii) the non-planned functions that may turn out. This paper presents a reasonable method to assemble the ITER VV. In this article, one parallel mobile robot, running on the track rail fixed on the wall inside the VV, is designed and tested. The assembling process, carried out by the mobile robot together with the welding robot, is presented.

High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

1999-02-01

In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10 6 R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

Energy Technology Data Exchange (ETDEWEB)

Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

1999-02-01

In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

Progress on the design development and prototype manufacturing of the ITER In-vessel coils

NARCIS (Netherlands)

Encheva, A.; Omran, H.; Devred, A.; Vostner, A.; Mitchell, N.; Mariani, N.; Jun, CH H.; Long, F.; Zhou, C.; Macklin, B.; Marti, H. P.; Sborchia, C.; della Corte, A. Della; Di Zenobio, A.; Anemona, A.; Righetti, R.; Wu, Y.; Jin, H.; Xu, A.; Jin, J.

2017-01-01

ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide a reliable Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements.

From rationality to cooperativeness: The totally mixed Nash equilibrium in Markov strategies in the iterated Prisoner's Dilemma.

Directory of Open Access Journals (Sweden)

Ivan S Menshikov

Full Text Available In this research, the social behavior of the participants in a Prisoner's Dilemma laboratory game is explained on the basis of the quantal response equilibrium concept and the representation of the game in Markov strategies. In previous research, we demonstrated that social interaction during the experiment has a positive influence on cooperation, trust, and gratefulness. This research shows that the quantal response equilibrium concept agrees only with the results of experiments on cooperation in Prisoner's Dilemma prior to social interaction. However, quantal response equilibrium does not explain of participants' behavior after social interaction. As an alternative theoretical approach, an examination was conducted of iterated Prisoner's Dilemma game in Markov strategies. We built a totally mixed Nash equilibrium in this game; the equilibrium agrees with the results of the experiments both before and after social interaction.

Simulation of VDE under intervention of vertical stability control and vertical electromagnetic force on the ITER vacuum vessel

International Nuclear Information System (INIS)

Miyamoto, S.; Sugihara, M.; Shinya, K.; Nakamura, Y.; Toshimitsu, S.; Lukash, V.E.; Khayrutdinov, R.R.; Sugie, T.; Kusama, Y.; Yoshino, R.

2012-01-01

Highlights: ► Taking account of intervention of VS control, VDE simulations were carried out. ► Malfunctioning of VS circuit (positive feedback) enhances the vertical force. ► The worst case was explored for vertical force on the ITER vacuum vessel. ► We confirmed the force is still within the design margin even if the worst case. - Abstract: Vertical displacement events (VDEs) and disruptions usually take place under intervention of vertical stability (VS) control and the vertical electromagnetic force induced on vacuum vessels is potentially influenced. This paper presents assessment of the force that arises from the VS control in ITER VDEs using a numerical simulation code DINA. The focus is on a possible malfunctioning of the ex-vessel VS control circuit: radial magnetic field is unintentionally applied to the direction of enhancing the vertical displacement further. Since this type of failure usually causes the largest forces (or halo currents) observed in the present experiments, this situation must be properly accommodated in the design of the ITER vacuum vessel. DINA analysis shows that although the ex-vessel VS control modifies radial field, it does not affect plasma motion and current quench behavior including halo current generation because the vacuum vessel shields the field created by the ex-vessel coils. Nevertheless, the VS control modifies the force on the vessel by directly acting on the eddy current carried by the conducting structures of the vessel. Although the worst case was explored in a range of plasma inductance and pattern of VS control in combination with the in-vessel VS control circuit, the result confirmed that the force is still within the design margin.

Design developments for the ITER in-Vessel equilibrium pick-up Coils and Halo current Sensors

International Nuclear Information System (INIS)

Chitarin, G; Grando, L.; Pomaro, N.; Peruzzo, S.; Taccon, C.

2006-01-01

The ITER magnetic diagnostics must provide essential information to be used both for plasma diagnostic purposes, and as feedback signals for the machine control loops. Some of the sensors have to be installed in a hostile environment characterized by severe neutron irradiation and plasma heat loads, which can reduce the sensor lifetime (due to mechanical and electrical damage) and also generate undesired DC signals, which might compromise the accuracy of the measurements obtained by time-integration. The paper is focused on the design development and optimization of a typical in-vessel tangential pick-up Coil. The work is aimed to achieve the required measurement precision in spite of Radiation Induced Electromotive Force (RIEMF) and Radiation Induced Thermo-Electric Sensitivity (RITES), which have recently been documented to take place in Mineral Insulated Cables (MIC). To this purpose, a substantial reduction of the thermal gradient and the maximum temperature due to nuclear heating in the pick-up coils is considered necessary. Within the limits of several heavy engineering constraints, a new concept of magnetic pick up coil has been developed. A winding made of a ceramic-coated conductor (instead of a MIC) and '' impregnated '' with ceramic filler is proposed. Different material choices for the coil support structure have been investigated. Similar issues are related to the Halo Sensor design. The possibility of replacing the circular tubes used as support of the Rogowski coils with a ceramic support in order to avoid the non-linear effect of the magnetic material has also been studied. The replacement of the MIC of the winding with a ceramic-coated wire is also investigated in order to increase of the effective area of the sensor. The paper includes also a critical review of each stage of the measurement chain (probes, cabling, conditioning electronics and data acquisition) in order to assess the compliance with the overall system precision that is required for

Integration test of ITER full-scale vacuum vessel sector

International Nuclear Information System (INIS)

Nakahira, M.; Koizumi, K.; Oka, K.

2001-01-01

The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of ± 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

Stress analysis of a double-wall vacuum vessel for ITER

International Nuclear Information System (INIS)

Conner, D.L.; Williamson, D.E.; Nelson, B.E.

1991-01-01

The preliminary structural analyses performed in support of the design of the vacuum vessel for the International Thermonuclear Experimental Reactor (ITER) are described. A thin, double-wall, all-welded structure is the proposed design concept analyzed. The results of the static stress analysis indicate the adequacy of such a structure. The effects of the proposed high-aspect-ratio design configuration on loading and stresses are also discussed. 4 refs., 6 figs., 1 tab

Demonstration tests for manufacturing the ITER vacuum vessel

International Nuclear Information System (INIS)

Shimizu, Katsusuke; Onozuka, Masanori; Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro; Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi

2007-01-01

Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV

Demonstration tests for manufacturing the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: katsusuke_shimizu@mhi.co.jp; Onozuka, Masanori [Mitsubishi Heavy Industries, Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan); Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)

2007-10-15

Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV.

Manufacturing preparations for the European Vacuum Vessel Sector for ITER

International Nuclear Information System (INIS)

Jones, Lawrence; Arbogast, Jean François; Bayon, Angel; Bianchi, Aldo; Caixas, Joan; Facca, Aldo; Fachin, Gianbattista; Fernández, José; Giraud, Benoit; Losasso, Marcello; Löwer, Thorsten; Micó, Gonzalo; Pacheco, Jose Miguel; Paoletti, Roberto; Sanguinetti, Gian Paolo; Stamos, Vassilis; Tacconelli, Massimiliano; Trentea, Alexandru; Utin, Yuri

2012-01-01

The contract for the seven European Sectors of the ITER Vacuum Vessel, which has very tight tolerances and high density of welding, was placed at the end of 2010 with AMW, a consortium of three companies. The start-up of the engineering, including R and D, design and analysis activities of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. The statutory and regulatory requirements of ITER Organization and the French Nuclear Safety regulations have made the design development subject to rigorous controls. AMW was able to make use of the previous extensive R and D and prototype work carried out during the past 9 years, especially in relation to advanced welding and inspection techniques. The paper describes the manufacturing methodology with the focus on controlling distortion with predictions by analysis, avoiding use of welded-on jigs, and making use of low heat input narrow-gap welding with electron beam welding as far as possible and narrow-gap TIG when not. Further R and D and more than ten significant mock-ups are described. All these preparations will help to assure the successful manufacture of this critical path item of ITER.

General and crevice corrosion study of the in-wall shielding materials for ITER vacuum vessel

Science.gov (United States)

Joshi, K. S.; Pathak, H. A.; Dayal, R. K.; Bafna, V. K.; Kimihiro, Ioki; Barabash, V.

2012-11-01

Vacuum vessel In-Wall Shield (IWS) will be inserted between the inner and outer shells of the ITER vacuum vessel. The behaviour of IWS in the vacuum vessel especially concerning the susceptibility to crevice of shielding block assemblies could cause rapid and extensive corrosion attacks. Even galvanic corrosion may be due to different metals in same electrolyte. IWS blocks are not accessible until life of the machine after closing of vacuum vessel. Hence, it is necessary to study the susceptibility of IWS materials to general corrosion and crevice corrosion under operations of ITER vacuum vessel. Corrosion properties of IWS materials were studied by using (i) Immersion technique and (ii) Electro-chemical Polarization techniques. All the sample materials were subjected to a series of examinations before and after immersion test, like Loss/Gain weight measurement, SEM analysis, and Optical stereo microscopy, measurement of surface profile and hardness of materials. After immersion test, SS 304B4 and SS 304B7 showed slight weight gain which indicate oxide layer formation on the surface of coupons. The SS 430 material showed negligible weight loss which indicates mild general corrosion effect. On visual observation with SEM and Metallography, all material showed pitting corrosion attack. All sample materials were subjected to series of measurements like Open Circuit potential, Cyclic polarization, Pitting potential, protection potential, Critical anodic current and SEM examination. All materials show pitting loop in OC2 operating condition. However, its absence in OC1 operating condition clearly indicates the activity of chloride ion to penetrate oxide layer on the sample surface, at higher temperature. The critical pitting temperature of all samples remains between 100° and 200°C.

Computational models for electromagnetic transients in ITER vacuum vessel, cryostat and thermal shield

International Nuclear Information System (INIS)

Alekseev, A.; Arslanova, D.; Belov, A.; Belyakov, V.; Gapionok, E.; Gornikel, I.; Gribov, Y.; Ioki, K.; Kukhtin, V.; Lamzin, E.; Sugihara, M.; Sychevsky, S.; Terasawa, A.; Utin, Y.

2013-01-01

A set of detailed computational models are reviewed that covers integrally the system “vacuum vessel (VV), cryostat, and thermal shields (TS)” to study transient electromagnetics (EMs) in the ITER machine. The models have been developed in the course of activities requested and supervised by the ITER Organization. EM analysis is enabled for all ITER operational scenarios. The input data are derived from results of DINA code simulations. The external EM fields are modeled accurate to the input data description. The known magnetic shell approach can be effectively applied to simulate thin-walled structures of the ITER machine. Using an integral–differential formulation, a single unknown is determined within the shells in terms of the vector electric potential taken only at the nodes of a finite-element (FE) mesh of the conducting structures. As a result, the FE mesh encompasses only the system “VV + Cryostat + TS”. The 3D model requires much higher computational resources as compared to a shell model based on the equivalent approximation. The shell models have been developed for all principal conducting structures in the system “VV + Cryostat + TS” including regular ports and neutral beam ports. The structures are described in details in accordance with the latest design. The models have also been applied for simulations of EM transients in components of diagnostic systems and cryopumps and estimation of the 3D effects of the ITER structures on the plasma performance. The developed models have been elaborated and applied for the last 15 years to support the ITER design activities. The finalization of the ITER VV design enables this set of models to be considered ready to use in plasma-physics computations and the development of ITER simulators

Structural analysis of the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Sannazzaro, G.; Ioki, K.; Johnson, G.; Onozuka, M.; Utin, Y. [ITER Joint Work Site, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States); Swanson, J. [USHT, Raytheon, Princeton (United States)

1998-07-01

The ITER Vacuum Vessel (VV) must withstand a large number of loading conditions including electromagnetic, seismic, operational and upset pressure, thermal and test loads. All of the loading conditions and load combinations have been categorized and classified to permit the allowable stress to be defined in accordance with the recommendations of the ASME code. The most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and the load combination of seismic and electromagnetic loads due to a plasma vertical instability. The areas of high stress are the regions around the VV and the blanket supports, and the attachment of the ports to the main shell. In all of the loading conditions and load combinations the calculated stresses are below the allowable values. (authors)

Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

International Nuclear Information System (INIS)

Sannazzaro, G.; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H.

2001-01-01

In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses

Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Sannazzaro, G. E-mail: sannazg@itereu.de; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H

2001-11-01

In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses.

Ex-vessel break in ITER divertor cooling loop analysis with the ECART code

CERN Document Server

Cambi, G; Parozzi, F; Porfiri, MT

2003-01-01

A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal-hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal-hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed.

Assembly and gap management strategy for the ITER NBI vessel passive magnetic shield

Energy Technology Data Exchange (ETDEWEB)

Ríos, Luis, E-mail: luis.rios@ciemat.es [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Ahedo, Begoña; Alonso, Javier; Barrera, Germán; Cabrera, Santiago; Rincón, Esther; Ramos, Francisco [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); El-Ouazzani, Anass; Graceffa, Joseph; Urbani, Marc; Shah, Darshan [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

2015-10-15

The neutral beam system for ITER consists of two heating and current drive neutral ion beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. The HNB Passive Magnetic Shield (PMS) works in conjunction with the active compensation/correction coils to limit the magnetic field inside the Beam Line Vessel (BLV), Beam Source Vessel (BSV), High Voltage Bushing (HVB) and Transmission Line (TL) elbow to acceptable levels that do not interfere with the operation of the HNB components. This paper describes the current design of the PMS, having had only minor modifications since the preliminary design review (PDR) held in IO in April 2013, and the assembly strategy for the vessel PMS.

Design of In-vessel neutron monitor using micro fission chambers for ITER

International Nuclear Information System (INIS)

Nishitani, Takeo; Kasai, Satoshi

2001-10-01

A neutron monitor using micro fission chambers to be installed inside the vacuum vessel has been designed for compact ITER (ITER-FEAT). We investigated the responses of the micro fission chambers to find the suitable position of micro fission chambers by a neutron Monte Carlo calculation using MCNP version 4b code. It was found that the averaged output of the micro fission chambers behind blankets at upper outboard and lower outboard is insensitive to the changes in the plasma position and the neutron source profile. A set of 235 U micro fission chamber and ''blank'' detector which is a fissile material free detector to identify noise issues such as from γ-rays are installed behind blankets. Employing both pulse counting mode and Campbelling mode in the electronics, the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution can be accomplished. The in-situ calibration has been simulated by MCNP calculation, where a point source of 14 MeV neutrons is moving on the plasma axis. It was found that the direct calibration is possible by using a neutron generator with an intensity of 10 11 n/s. The micro fission chamber system can meet the required 10% accuracy for a fusion power monitor. (author)

Bolted Ribs Analysis for the ITER Vacuum Vessel using Finite Element Submodelling Techniques

Energy Technology Data Exchange (ETDEWEB)

Zarzalejos, José María, E-mail: jose.zarzalejos@ext.f4e.europa.eu [External at F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Fernández, Elena; Caixas, Joan; Bayón, Angel [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Polo, Joaquín [Iberdrola Ingeniería y Construcción, Avenida de Manoteras 20, 28050 Madrid (Spain); Guirao, Julio [Numerical Analysis Technologies, S L., Marqués de San Esteban 52, Entlo, 33209 Gijon (Spain); García Cid, Javier [Iberdrola Ingeniería y Construcción, Avenida de Manoteras 20, 28050 Madrid (Spain); Rodríguez, Eduardo [Mechanical Engineering Department EPSIG, University of Oviedo, Gijon (Spain)

2014-10-15

Highlights: • The ITER Vacuum Vessel Bolted Ribs assemblies are modelled using Finite Elements. • Finite Element submodelling techniques are used. • Stress results are obtained for all the assemblies and a post-processing is performed. • All the elements of the assemblies are compliant with the regulatory provisions. • Submodelling is a time-efficient solution to verify the structural integrity of this type of structures. - Abstract: The ITER Vacuum Vessel (VV) primary function is to enclose the plasmas produced by the ITER Tokamak. Since it acts as the first radiological barrier of the plasma, it is classified as a class 2 welded box structure, according to RCC-MR 2007. The VV is made of an inner and an outer D-shape, 60 mm-thick double shell connected through thick massive bars (housings) and toroidal and poloidal structural stiffening ribs. In order to provide neutronic shielding to the ex-vessel components, the space between shells is filled with borated steel plates, called In-Wall Shielding (IWS) blocks, and water. In general, these blocks are connected to the IWS ribs which are connected to adjacent housings. The development of a Finite Element model of the ITER VV including all its components in detail is unaffordable from the computational point of view due to the large number of degrees of freedom it would require. This limitation can be overcome by using submodelling techniques to simulate the behaviour of the bolted ribs assemblies. Submodelling is a Finite Element technique which allows getting more accurate results in a given region of a coarse model by generating an independent, finer model of the region under study. In this paper, the methodology and several simulations of the VV bolted ribs assemblies using submodelling techniques are presented. A stress assessment has been performed for the elements involved in the assembly considering possible types of failure and including stress classification and categorization techniques to analyse

Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

Science.gov (United States)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R.

1998-10-01

Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

Strong Convergence Iterative Algorithms for Equilibrium Problems and Fixed Point Problems in Banach Spaces

Directory of Open Access Journals (Sweden)

Shenghua Wang

2013-01-01

Full Text Available We first introduce the concept of Bregman asymptotically quasinonexpansive mappings and prove that the fixed point set of this kind of mappings is closed and convex. Then we construct an iterative scheme to find a common element of the set of solutions of an equilibrium problem and the set of common fixed points of a countable family of Bregman asymptotically quasinonexpansive mappings in reflexive Banach spaces and prove strong convergence theorems. Our results extend the recent ones of some others.

Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

Energy Technology Data Exchange (ETDEWEB)

Zaccaria, Pierluigi, E-mail: pierluigi.zaccaria@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Masiello, Antonio [Fusion for Energy F4E, Barcelona (Spain); Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario [Ettore Zanon S.p.A., Schio (VI) (Italy)

2015-10-15

Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

International Nuclear Information System (INIS)

Zaccaria, Pierluigi; Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco; Masiello, Antonio; Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario

2015-01-01

Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

Results from ITER Vacuum Vessel Sector Manufacturing Development in Europe

International Nuclear Information System (INIS)

Jones, L.

2006-01-01

Significant results have been achieved since the previous SOFT conference, when the manufacturing development work required to prepare for the ITER Vacuum Vessel Sector was described. The contract for the manufacture of a full-size, 20 Ton poloidal part of the inboard section, fabricated according to the ITER reference manufacturing route, including bracing fixtures, welding applications, restraint effects, and fit-up aspects is approaching completion. Since the main aim of the work is to establish the practicability of achieving the tight dimensional tolerances, an accompanying SYSWELD analysis programme has been validation by instrumented welding coupons, and then used for predicting the distortion of the actual construction. A local machining tool has been developed to allow the requirement for machining of the cylindrical features at a late stage of manufacture. Experimental and analytical work has also been carried out to establish the possibility of 3-D cold-forming large sections of walls of the VV. A manufacturing programme to validate an alternative method of fabricating parts of the double-walled VV, utilising e-beam welding only and avoiding the quality issues of the one-sided access and inspection of the closing welds is presented. This paper describes the results of the manufacturing development programme and the future activities. (author)

A mobile robot with parallel kinematics to meet the requirements for assembling and machining the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Pessi, Pekka [Lappeenranta University of Technology, Lappeenranta (Finland)], E-mail: pessi@lut.fi; Wu, Huapeng; Handroos, Heikki [Lappeenranta University of Technology, Lappeenranta (Finland); Jones, Lawrence [EFDA Close Support Unit, Boltzmannstrasse 2, Garching D-85748 (Germany)

2007-10-15

The present paper introduces a mobile parallel robot developed for International Thermonuclear Experimental Reactor (ITER). The task of the robot is to carry out welding and machining processes inside the ITER vacuum vessel. The kinematic design of the robot has been optimized for the ITER access. The kinematic analysis is given in the paper. A virtual prototype of the parallel robot is built. A dynamic behavior of the whole robot is studied by the multi-body system simulation (MBS)

A mobile robot with parallel kinematics to meet the requirements for assembling and machining the ITER vacuum vessel

International Nuclear Information System (INIS)

Pessi, Pekka; Wu, Huapeng; Handroos, Heikki; Jones, Lawrence

2007-01-01

The present paper introduces a mobile parallel robot developed for International Thermonuclear Experimental Reactor (ITER). The task of the robot is to carry out welding and machining processes inside the ITER vacuum vessel. The kinematic design of the robot has been optimized for the ITER access. The kinematic analysis is given in the paper. A virtual prototype of the parallel robot is built. A dynamic behavior of the whole robot is studied by the multi-body system simulation (MBS)

Design of ex-vessel neutron monitor for ITER

International Nuclear Information System (INIS)

Nishitani, Takeo; Yamauchi, Michinori; Kasai, Satoshi; Ebisawa, Katsuyuki; Walker, Chris

2002-07-01

A neutron flux monitor has been designed by using 235 U fission chambers to be installed outside the vacuum vessel of ITER. We investigated moderator materials to get flat energy response the responses of 235 U fission chambers. Here we employed graphite and beryllium with a ratio of Be/C=0.25 as moderator, which materials are stable in ITER relevant temperature in a horizontal port. Based on the neutronics calculations, a fission chamber with 200 mg of 235 U is adopted for the neutron flux monitor. Three detectors are mounted in a stainless steel housing with moderation material. Two fission chamber assemblies will be installed in a horizontal port; one is for D-D and calibration operation, and another is for D-T operation. The assembly for the D-D operation and the calibration are installed just outside the port plug in the horizontal port. The assembly for the D-T operation is installed just behind the additional shield in the port. Combining of those assemblies with both pulse counting mode and Campbelling mode in the electronics, a dynamic range of 10 7 can be obtained with 1 ms temporal resolution. Effects of gamma-rays and magnetic fields on the fission chamber are negligible in this arrangement. The neutron flux monitor can meet the required 10% accuracy for a fusion power monitor. (author)

Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R. [ITER JCT, Garching (Germany)

1998-10-01

Design and R and D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R and D results. The resulting design changes are discussed for each system. (orig.) 11 refs.

Gamma irradiation testing of prototype ITER in-vessel magnetic pick-up coils

International Nuclear Information System (INIS)

Vermeeren, Ludo; Leysen, Willem

2013-01-01

Highlights: ► We tested five prototype ITER in-vessel coils up to a gamma dose of 72 MGy. ► Before and after irradiation thermal tests were also performed from 30 °C till 130 °C. ► The continuity resistances and the insulation resistances were continuously monitored. ► The observed behavior of all coils was satisfactory in all conditions. ► For the further design the mechanical robustness should be taken into account. -- Abstract: To fulfill the requirements for ITER in-vessel magnetic diagnostics, several coil prototypes have been developed, aiming at minimizing the disturbing effects of temperature gradients and radiation induced phenomena. As a first step in the radiation resistance testing of these prototypes, an in-situ high dose rate gamma radiation test on a selection of prototypes was performed. The aim of this test was to get a first experimental feedback regarding the behavior of the pick-up coil prototypes under radiation. Five prototypes (a coil wound with glass-insulated copper wire, two LTCC coils and two HTCC coils) were irradiated at a dose rate of 46 kGy/h up to a total dose of 72 MGy and at a temperature of 50 °C. During the irradiation, the continuity resistances and the insulation resistances were continuously measured. Before and after irradiation reference data were recorded as a function of temperature (from 30 °C to 130 °C). This paper includes the results of the temperature and irradiation tests and a discussion of the behavior of the prototype coils in terms of electrical and mechanical properties

Preliminary assessment for dust contamination of ITER In-Vessel Transporter

Energy Technology Data Exchange (ETDEWEB)

Saito, Makiko, E-mail: saito.makiko@jaea.go.jp [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Ueno, Kenichi; Maruyama, Takahito; Murakami, Shin; Takeda, Nobukazu; Kakudate, Satoshi [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Nakahira, Masataka; Tesini, Alessandro [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France)

2014-10-15

Highlights: •To assess the exposure to the maintenance workers, we calculated the effective dose rate. •To reduce the effective dose rate, the IVT was decontaminated and underwent a design change. •The effective dose rate at each maintenance point was also calculated. -- Abstract: After plasma operations of ITER, radioactive dust will have accumulated in the vacuum vessel (VV). The In-Vessel Transporter (IVT) will be introduced into the VV to remove the shield blanket modules for maintenance or replacement and later reinstall them. The IVT itself also needs to undergo regular maintenance in the Hot Cell Facility (HCF). It is assumed that maintenance workers will be exposed to radioactive dust that has adhered to the surfaces of the IVT. In this study, the areas of the IVT that may be contaminated by dust are evaluated to assess the level of exposure to workers during maintenance work in the HCF. Decontamination processes for the IVT, such as a combination of vacuuming and brushing, were investigated and the dose rate after these processes was evaluated. Even though dust was removed from surfaces where decontamination was possible, the dose rate was very high at some assessment points. To decrease the dose rate in accordance with ALARA policy, a decontamination plan and a maintenance plan, which includes the removal of dust, a radiation shield system, and a reduction in working time are proposed.

Assessment of alternative vessel and blanket design on ITER operation

Energy Technology Data Exchange (ETDEWEB)

Cavinato, M., E-mail: mario.cavinato@f4e.europa.e [FUSION FOR ENERGY Joint Undertaking, 08019 Barcelona (Spain); Portone, A.; Saibene, G.; Sartori, R. [FUSION FOR ENERGY Joint Undertaking, 08019 Barcelona (Spain); Albanese, R.; Ambrosino, G.; Ariola, M. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Napoli (Italy); Artaserse, G. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Reggio Calabria (Italy); Mattei, M. [Associazione Euratom-ENEA-CREATE, DIAM, Seconda Universita di Napoli, Via Roma 29, Aversa, CE 81031 Italy (Italy); Pironti, A. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Napoli (Italy); Villone, F. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Cassino (Italy)

2010-12-15

In the framework of the ITER project, an investigation has been conducted on an alternative vessel and blanket design, aimed at reducing cost and production risk. The modifications proposed have a strong impact on plasma control since they affect the main conducting structures surrounding the plasma column, providing passive stabilization but at the same time shielding the field generated by the active coils to control the plasma motion and shape. An extensive analysis was performed to assess the plasma vertical controllability and the modified requirements to the in-vessel vertical stability coils system as well as to the external Poloidal Field coils system. A similar analysis was aimed at assessing the performance of the shape control system in presence of the modified structures. The effect on plasma breakdown was also evaluated in terms of maximum initial loop voltage, quality of magnetic null and the flux loss related to the breakdown delay that was quantified under the same hypothesis employed by ITER for the baseline design. Furthermore, the modified design presents issues for the magnetic diagnostic system, related to the shielding of the probes by the eddy currents, which were analysed with a 3D model. The results of the analyses performed have some general interest in particular regarding the influence on plasma stability of 3D structures with close proximity to the plasma. The present paper aims at giving an overview of the analyses that have been carried out and a summary of the results in terms of impact of the modified design on plasma control and scenario, and in general an evaluation of the role of passive structure in plasma vertical stability and shape control.

ITER EDA Newsletter. V.2, no.6

International Nuclear Information System (INIS)

1993-06-01

This issue of the newsletter on the Engineering Design Activities for the ITER Project includes (i) a status report on these activities describing the development of the design and design parameters, the research and development program, the joint central team, and the joint work sites; (ii) a description of the In-Vessel Ancillaries Division (consisting of the RF Heating and Current Drive Group and the In-Vessel Diagnostics Group) including its organizational chart and priorities of activities; (iii) a report on the Technical Meeting on Design Standards and Remote Handling held at the San Diego Co-Centre from May 24-28, 1993; (iv) and a report on a Technical Committee Meeting on Plasma Equilibrium and Control held at the Naka Co-Centre on April 26-29, 1993

Design progress of the ITER vacuum vessel sectors and port structures

International Nuclear Information System (INIS)

Utin, Yu.; Ioki, K.; Alekseev, A.; Bachmann, Ch.; Cho, S.; Chuyanov, V.; Jones, L.; Kuzmin, E.; Morimoto, M.; Nakahira, M.; Sannazzaro, G.

2007-01-01

Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV Sector (no. 1) has been nearly finalized. Design improvement of other sectors is in progress-in particular, of the VV Sectors no. 2 and no. 3 which interface with tangential ports for the neutral beam (NB) injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB ports was advanced with the focus on the beam-facing components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements

Novel Robot Solutions for Carrying out Field Joint Welding and Machining in the Assembly of the Vacuum Vessel of ITER

International Nuclear Information System (INIS)

Pessi, P.

2009-01-01

It is necessary to use highly specialized robots in ITER (International Thermonuclear Experimental Reactor) both in the manufacturing and maintenance of the reactor due to a demanding environment. The sectors of the ITER vacuum vessel (VV) require more stringent tolerances than normally expected for the size of the structure involved. VV consists of nine sectors that are to be welded together. The vacuum vessel has a toroidal chamber structure. The task of the designed robot is to carry the welding apparatus along a path with a stringent tolerance during the assembly operation. In addition to the initial vacuum vessel assembly, after a limited running period, sectors need to be replaced for repair. Mechanisms with closed-loop kinematic chains are used in the design of robots in this work. One version is a purely parallel manipulator and another is a hybrid manipulator where the parallel and serial structures are combined. Traditional industrial robots that generally have the links actuated in series are inherently not very rigid and have poor dynamic performance in high speed and high dynamic loading conditions. Compared with open chain manipulators, parallel manipulators have high stiffness, high accuracy and a high force/torque capacity in a reduced workspace. Parallel manipulators have a mechanical architecture where all of the links are connected to the base and to the end-effector of the robot. The purpose of this thesis is to develop special parallel robots for the assembly, machining and repairing of the VV of the ITER. The process of the assembly and machining of the vacuum vessel needs a special robot. By studying the structure of the vacuum vessel, two novel parallel robots were designed and built; they have six and ten degrees of freedom driven by hydraulic cylinders and electrical servo motors. Kinematic models for the proposed robots were defined and two prototypes built. Experiments for machine cutting and laser welding with the 6-DOF robot were

Novel Robot Solutions for Carrying out Field Joint Welding and Machining in the Assembly of the Vacuum Vessel of ITER

Energy Technology Data Exchange (ETDEWEB)

Pessi, P.

2009-07-01

It is necessary to use highly specialized robots in ITER (International Thermonuclear Experimental Reactor) both in the manufacturing and maintenance of the reactor due to a demanding environment. The sectors of the ITER vacuum vessel (VV) require more stringent tolerances than normally expected for the size of the structure involved. VV consists of nine sectors that are to be welded together. The vacuum vessel has a toroidal chamber structure. The task of the designed robot is to carry the welding apparatus along a path with a stringent tolerance during the assembly operation. In addition to the initial vacuum vessel assembly, after a limited running period, sectors need to be replaced for repair. Mechanisms with closed-loop kinematic chains are used in the design of robots in this work. One version is a purely parallel manipulator and another is a hybrid manipulator where the parallel and serial structures are combined. Traditional industrial robots that generally have the links actuated in series are inherently not very rigid and have poor dynamic performance in high speed and high dynamic loading conditions. Compared with open chain manipulators, parallel manipulators have high stiffness, high accuracy and a high force/torque capacity in a reduced workspace. Parallel manipulators have a mechanical architecture where all of the links are connected to the base and to the end-effector of the robot. The purpose of this thesis is to develop special parallel robots for the assembly, machining and repairing of the VV of the ITER. The process of the assembly and machining of the vacuum vessel needs a special robot. By studying the structure of the vacuum vessel, two novel parallel robots were designed and built; they have six and ten degrees of freedom driven by hydraulic cylinders and electrical servo motors. Kinematic models for the proposed robots were defined and two prototypes built. Experiments for machine cutting and laser welding with the 6-DOF robot were

Physics design of the in-vessel collection optics for the ITER electron cyclotron emission diagnostic

Energy Technology Data Exchange (ETDEWEB)

Rowan, W. L., E-mail: w.l.rowan@austin.utexas.edu; Houshmandyar, S.; Phillips, P. E.; Austin, M. E. [Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States); Beno, J. H.; Ouroua, A. [Center for Electromechanics, The University of Texas at Austin, Austin, Texas 78712 (United States); Hubbard, A. E. [Plasma Science and Fusion Center, MIT, Cambridge, Massachusetts 02139 (United States); Khodak, A.; Taylor, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

2016-11-15

Measurement of the electron cyclotron emission (ECE) is one of the primary diagnostics for electron temperature in ITER. In-vessel, in-vacuum, and quasi-optical antennas capture sufficient ECE to achieve large signal to noise with microsecond temporal resolution and high spatial resolution while maintaining polarization fidelity. Two similar systems are required. One views the plasma radially. The other is an oblique view. Both views can be used to measure the electron temperature, while the oblique is also sensitive to non-thermal distortion in the bulk electron distribution. The in-vacuum optics for both systems are subject to degradation as they have a direct view of the ITER plasma and will not be accessible for cleaning or replacement for extended periods. Blackbody radiation sources are provided for in situ calibration.

Performance test of micro-fission chambers for in-vessel neutron monitoring of ITER

International Nuclear Information System (INIS)

Yamauchi, Michinori; Nishitani, Takeo; Ochiai, Kentaro; Morimoto, Yuichi; Hori, Jun-ichi; Ebisawa, Katsuyuki; Kasai, Satoshi

2002-03-01

A micro-fission chamber with 12 mg UO 2 and a dummy chamber without uranium were fabricated and the performance was tested. They are designed to be installed inside the vacuum vessel of the compact ITER (ITER-FEAT) for neutron monitoring. The vacuum leak rate of the dummy chamber with MI cable, resistances of chambers between central conductor and outer sheath, and mechanical strength up to 50G acceleration were confirmed to meet the design criteria. The gamma-ray sensitivity was measured for the dummy chamber with the 60 Co gamma-ray irradiation facility at JAERI Takasaki. The output signals for gamma-rays in Campbelling mode were estimated to be less than 0.1% of those by neutrons at the location behind the blanket module in ITER-FEAT. The detector response for 14 MeV neutrons was investigated with the FNS facility. Excellent linearity between count rates, square of Campbelling voltage and neutron fluxes was confirmed in the temperature range from 20degC (room) to 250degC. However, a positive dependence of 14 MeV neutron count rates on temperature was observed, which might be caused by the increase in the pulse height with temperature rise. Effects of a change of surrounding materials were evaluated by the sensitivity measurements of the micro-fission chamber inserted into the shielding blanket mock-up. The sensitivity was enhanced by slow-downed neutrons, which agreed with the calculation result by MCNP-4C code. As a result, it was concluded that the developed micro-fission chamber is applicable for ITER-FEAT. (author)

Iterative algorithms for computing the feedback Nash equilibrium point for positive systems

Science.gov (United States)

Ivanov, I.; Imsland, Lars; Bogdanova, B.

2017-03-01

The paper studies N-player linear quadratic differential games on an infinite time horizon with deterministic feedback information structure. It introduces two iterative methods (the Newton method as well as its accelerated modification) in order to compute the stabilising solution of a set of generalised algebraic Riccati equations. The latter is related to the Nash equilibrium point of the considered game model. Moreover, we derive the sufficient conditions for convergence of the proposed methods. Finally, we discuss two numerical examples so as to illustrate the performance of both of the algorithms.

Study on poloidal field coil optimization and equilibrium control of ITER

International Nuclear Information System (INIS)

Shinya, Kichiro; Sugihara, Masayoshi; Nishio, Satoshi

1989-03-01

The purpose of this report is to present general features of the poloidal field coil optimization for the ITER plasma, flexibility analysis for various plasma options and some other aspect of the equilibrium control which is required for understanding plasma operation in more detail. Double null divertor plasma was selected as a main object of the optimization. Single null divertor plasma was assumed to be an alternative, because single null divertor plasma can be operational within the amounts of the total stored energy and ampere-turns of the double null divertor plasma, if it is shaped appropriately. Plasma parameters used in the present analysis are mainly those employed in the preliminary study by the Basic Device Engineering group of the ITER design team. The most part of the optimization study, however, utilizes the parameters proposed for discussion by the Japan team before starting joint design work at Garching. Plasma shape, and solenoid coil shape and size, which maximize available flux swing with reasonable amounts of the stored energy and ampere-turns, are discussed. Location and minimum number of the poloidal field coils with adequate shaping controllability were also discussed for various plasma options. Some other aspect of the equilibrium control, such as separatrix swing, moving null point operation during plasma heating and possible range of li, were evaluated and the guideline for the engineering design was proposed. Finally, fusion power output was estimated for the different pressure profiles and combinations of the average density and temperature, and the magnetic quantities of the scrape-off region was calculated to be available for the future divertor analysis. (author)

Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

Science.gov (United States)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

2000-12-01

Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ˜50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R&D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R&D is being performed.

ITER ITA newsletter. No. 8, September 2003

International Nuclear Information System (INIS)

2003-10-01

This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities including Robert Aymar's leaving ITER for CERN, ITER related issues at the IAEA General Conference and status and prospects of thermonuclear power and activity during the ITA on materials foe vessel and in-vessel components

ITER vacuum vessel design and electromagnetic analysis on in-vessel components

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.; Iizuka, T.

1995-01-01

Major functional requirements for the vacuum vessel are to provide the first safety barrier and to support electromagnetic loads due to plasma disruptions and vertical displacement events, and to withstand plausible accidents without losing confinement. A double wall structure concept has been developed for the vacuum vessel due to its beneficial characteristics from the viewpoints of structural integrity and electrical continuity. An electromagnetic analysis of the blanket modules and the vacuum vessel has been performed to investigate force distributions on in-vessel components. According to the vertical displacement events (VDE) scenario, which assumes a critical q-value of 1.5, the total downward vertical force, induced by coupling between the eddy current and external fields, is about 110 MN. We have performed a stress analysis for the vacuum vessel using the VDE disruption forces acting on the blankets, and a maximum stress intensity of 112 MPa was obtained in the vicinity of the lower support of the vessel. (orig.)

Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

International Nuclear Information System (INIS)

Onozuka, M.; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H.

2001-01-01

Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region

Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H

2001-11-01

Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region.

A mobile robot with parallel kinematics constructed under requirements for assembling and machining of the ITER vacuum vessel

International Nuclear Information System (INIS)

Pessi, P.; Huapeng Wu; Handroos, H.; Jones, L.

2006-01-01

ITER sectors require more stringent tolerances ± 5 mm than normally expected for the size of structure involved. The walls of ITER sectors are made of 60 mm thick stainless steel and are joined together by high efficiency structural and leak tight welds. In addition to the initial vacuum vessel assembly, sectors may have to be replaced for repair. Since commercially available machines are too heavy for the required machining operations and the lifting of a possible e-beam gun column system, and conventional robots lack the stiffness and accuracy in such machining condition, a new flexible, lightweight and mobile robotic machine is being considered. For the assembly of the ITER vacuum vessel sector, precise positioning of welding end-effectors, at some distance in a confined space from the available supports, will be required, which is not possible using conventional machines or robots. This paper presents a special robot, able to carry out welding and machining processes from inside the ITER vacuum vessel, consisting of a ten-degree-of-freedom parallel robot mounted on a carriage driven by electric motor/gearbox on a track. The robot consists of a Stewart platform based parallel mechanism. Water hydraulic cylinders are used as actuators to reach six degrees of freedom for parallel construction. Two linear and two rotational motions are used for enlargement the workspace of the manipulator. The robot carries both welding gun such as a TIG, hybrid laser or e-beam welding gun to weld the inner and outer walls of the ITER vacuum vessel sectors and machining tools to cut and milling the walls with necessary accuracy, it can also carry other tools and material to a required position inside the vacuum vessel . For assembling an on line six degrees of freedom seam finding algorithm has been developed, which enables the robot to find welding seam automatically in a very complex environment. In the machining multi flexible machining processes carried out automatically by

An Iterative Algorithm to Determine the Dynamic User Equilibrium in a Traffic Simulation Model

Science.gov (United States)

Gawron, C.

An iterative algorithm to determine the dynamic user equilibrium with respect to link costs defined by a traffic simulation model is presented. Each driver's route choice is modeled by a discrete probability distribution which is used to select a route in the simulation. After each simulation run, the probability distribution is adapted to minimize the travel costs. Although the algorithm does not depend on the simulation model, a queuing model is used for performance reasons. The stability of the algorithm is analyzed for a simple example network. As an application example, a dynamic version of Braess's paradox is studied.

Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 1. Comprehending the vacuum vessel structure

International Nuclear Information System (INIS)

Onozuka, Masanori; Nakahira, Masataka

2006-01-01

The functions, conditions and structure of vacuum vessel using tokamak fusion machines are explained. The structural standard and code of vacuum vessel, process of vacuum vessel design, and design of ITER vacuum vessel are described. Production and maintenance of ultra high vacuum, confinement of radioactive materials, support of machines in vessel and electromagnetic force, radiation shield, plasma vertical stability, one-turn electric resistance, high temperature baking heat and remove of nuclear heat, reduce of troidal ripple, structural standard, features of safety of nuclear fusion machines, subjects of structural standard of fusion vacuum vessel, design flow of vacuum vessel, establishment of radial build, selections of materials, baking and cooling method, basic structure, structure of special parts, shield structure, and of support structure, and example of design of structure, ITER, are stated. (S.Y.)

Design standard issues for ITER in-vessel components

International Nuclear Information System (INIS)

Majumdar, S.

1994-01-01

Unique requirements that must be addressed by a structural design code for the ITER have been summarized. Existing codes such as ASME Section III, or the French RCC-MR were developed primarily for fission reactor out-of-core components and are not directly applicable to the ITER. They may be used either as a guide for developing a design code for the ITER or as interim standards. However, new rules will be needed for handling the irradiation-induced embrittlement problems faced by the ITER blanket components. Design standards developed in the past for the design of fission reactor core components in the United States can be used as guides in this area

Development and control towards a parallel water hydraulic weld/cut robot for machining processes in ITER vacuum vessel

International Nuclear Information System (INIS)

Wu Huapeng; Handroos, Heikki; Pessi, Pekka; Kilkki, Juha; Jones, Lawrence

2005-01-01

This paper presents a special robot, able to carry out welding and machining processes from inside the ITER vacuum vessel (VV), consisting of a five degree-of-freedom parallel mechanism, mounted on a carriage driven by two electric motors on a rack. The kinematic design of the robot has been optimised for ITER access and a hydraulically actuated pre-prototype built. A hybrid controller is designed for the robot, including position, speed and pressure feedback loops to achieve high accuracy and high dynamic performances. Finally, the experimental tests are given and discussed

Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

International Nuclear Information System (INIS)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

2000-01-01

Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ∼50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed

Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.deiokik@ipp.mpg.de; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M

2000-12-01

Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve {approx}50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed.

Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

International Nuclear Information System (INIS)

Rivas, J.C.; Dies, J.; Fajarnés, X.

2015-01-01

Highlights: • We have repeated the safety analysis for the hypothesis of passive plasma shutdown for beryllium evaporation during an ex-vessel LOCA of ITER first wall, with AINA code. • We have performed a sensitivity analysis over some key parameters that represents uncertainties in physics and engineering, to identify cliff edge effects. • The obtained results for the 500 MW inductive scenario, with an ex-vessel LOCA affecting a third of first wall surface are similar to those of previous studies and point to the possibility of a passive plasma shutdown during this safety case, before a serious damage is inflicted to the ITER wall. • The sensitivity analysis revealed a new scenario potentially damaging for the first wall if we increase fusion power and time delay for impurity transport, and decrease fraction of affected first wall area and initial beryllium fraction in plasma. • After studying the 700 MW inductive scenario, with an ex-vessel LOCA affecting 10% of first wall surface, with 0.5% of Be in plasma and a time delay twice the energy confinement time, it was found that affected area of first wall would melt before a passive plasma shutdown occurs. - Abstract: In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code. In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge. The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering

Impact of error fields on equilibrium configurations in ITER

Energy Technology Data Exchange (ETDEWEB)

Barbato, Lucio [DIEI, Università di Cassino and Lazio Meridionale, Cassino (Italy); Formisano, Alessandro, E-mail: alessandro.formisano@unina2.it [Department of Industrial and Information Engineering, Seconda Univ. di Napoli, Aversa (Italy); Martone, Raffaele [Department of Industrial and Information Engineering, Seconda Univ. di Napoli, Aversa (Italy); Villone, Fabio [DIEI, Università di Cassino and Lazio Meridionale, Cassino (Italy)

2015-10-15

Highlights: • Error fields (EF) are discrepancies from nominal magnetic field, and may alter plasma behaviour. • They are due to, e.g., coils manufacturing and assembly errors. • EF impact in ITER equilibria is analyzed using numerical simulations. • A high accuracy 3D field computation module and a Grad-Shafranov solver are used. • Deformations size allow using a linearized model, and performing a sensitivity analysis. - Abstract: Discrepancies between design and actual magnetic field maps in tokamaks are unavoidable, and are associated to a number of causes, e.g. manufacturing and assembly tolerances on magnets, presence of feeders and joints, non-symmetric iron parts. Such error fields may drive plasma to loss of stability, and must be carefully controlled using suitable correction coils. Anyway, even when kept below safety threshold, error fields may alter the behavior of plasma. The present paper, using as example the error fields induced by tolerances in toroidal field coils, quantifies their effect on the plasma boundary shape in equilibrium configurations. In particular, a procedure able to compute the shape perturbations due to given deformations of the coils has been set up and used to carry out a thorough statistical analysis of the error field-shape perturbations relationship.

Design and analysis of the vacuum vessel for RTO/RC-ITER

International Nuclear Information System (INIS)

Onozuka, M.; Ioki, K.; Johnson, G.; Kodama, T.; Sannazzaro, G.; Utin, Y.

2000-01-01

Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a 'tight fitting' configuration is required to correctly position the modules' first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible

Design and analysis of the vacuum vessel for RTO/RC-ITER

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Johnson, G.; Kodama, T.; Sannazzaro, G.; Utin, Y

2000-11-01

Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a 'tight fitting' configuration is required to correctly position the modules' first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible.

Progress in the design and R and D of the ITER In-Vessel Viewing and Metrology System (IVVS)

Energy Technology Data Exchange (ETDEWEB)

Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 – Torres Diagonal Litoral – Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Bates, Philip; Damiani, Carlo [Fusion for Energy, c/ Josep Pla, n°2 – Torres Diagonal Litoral – Edificio B3, 08019 Barcelona (Spain); Reichle, Roger; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2014-10-15

The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components, which in turn is related to the amount of mobilised dust present in the Vacuum Vessel. Periodically or on request, the IVVS scanning probes will be deployed into the Vacuum Vessel in order to acquire both visual and metrological data on plasma facing components (blanket, divertor, heating/diagnostic plugs, and test blanket modules). Recent design changes made to the six IVVS port extensions implied the need for a substantial redesign of the IVVS integrated concept – including the scanning probe and its deployment system – in order to bring it to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the concept design for IVVS as well as of the various engineering analyses and R and D activities carried out in support to this design: neutronic, seismic and electromagnetic analyses, probe actuation validation under environmental conditions.

Final Report on ITER Task Agreement 81-08

Energy Technology Data Exchange (ETDEWEB)

Richard L. Moore

2008-03-01

As part of an ITER Implementing Task Agreement (ITA) between the ITER US Participant Team (PT) and the ITER International Team (IT), the INL Fusion Safety Program was tasked to provide the ITER IT with upgrades to the fusion version of the MELCOR 1.8.5 code including a beryllium dust oxidation model. The purpose of this model is to allow the ITER IT to investigate hydrogen production from beryllium dust layers on hot surfaces inside the ITER vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). Also included in the ITER ITA was a task to construct a RELAP5/ATHENA model of the ITER divertor cooling loop to model the draining of the loop during a large ex-vessel pipe break followed by an in-vessel divertor break and compare the results to a simular MELCOR model developed by the ITER IT. This report, which is the final report for this agreement, documents the completion of the work scope under this ITER TA, designated as TA 81-08.

Detailed Design and Fabrication Method of the ITER Vacuum Vessel Ports

International Nuclear Information System (INIS)

Hee-Jae Ahn; Kwon, T.H.; Hong, Y.S.

2006-01-01

The engineering design of the ITER vacuum vessel (VV) has been progressed by the ITER International Team (IT) with the cooperation of several participant teams (PT). The VV and ports are the components allocated to Korea for the construction of the ITER. Hyundai Heavy Industries has been involved in the structural analysis, detailed design and development of the fabrication method of the upper and lower ports within the framework of the ITER transitional arrangements (ITA). The design of the port structures has been investigated to validate and to improve the conceptual designs of the ITER IT and other PT. The special emphasis was laid on the flange joint between the port extension and the in-port plug to develop the design of the upper port. The modified design with a pure friction type flange with forty-eight pieces of bolts instead of the tangential key is recommended. Furthermore, the alternative flange designs developed by the ITER IT have been analyzed in detail to simplify the lip seal maintenance into the port flange. The structural analyses of the lower RH port have been also performed to verify the capacity for supporting the VV. The maximum stress exceeds the allowable value at the reinforcing block and basement. More elaborate local models have been developed to mitigate the stress concentration and to modify the component design. The fabrication method and the sequence of the detailed fabrication for the ports are developed focusing on the cost reduction as well as the simplification. A typical port structure includes a port stub, a stub extension and a port extension with a connecting duct. The fabrication sequence consists of surface treatment, cutting, forming, cleaning, welding, machining, and non-destructive inspection and test. Tolerance study has been performed to avoid the mismatch of each fabricated component and to obtain the suitable tolerances in the assembly at the shop and site. This study is based on the experience in the fabrication of

Chatter suppression methods of a robot machine for ITER vacuum vessel assembly and maintenance

International Nuclear Information System (INIS)

Wu, Huapeng; Wang, Yongbo; Li, Ming; Al-Saedi, Mazin; Handroos, Heikki

2014-01-01

Highlights: •A redundant 10-DOF serial-parallel hybrid robot for ITER assembly and maintains is presented. •A dynamic model of the robot is developed. •A feedback and feedforward controller is presented to suppress machining vibration of the robot. -- Abstract: In the process of assembly and maintenance of ITER vacuum vessel (ITER VV), various machining tasks including threading, milling, welding-defects cutting and flexible hose boring are required to be performed from inside of ITER VV by on-site machining tools. Robot machine is a promising option for these tasks, but great chatter (machine vibration) would happen in the machining process. The chatter vibration will deteriorate the robot accuracy and surface quality, and even cause some damages on the end-effector tools and the robot structure itself. This paper introduces two vibration control methods, one is passive and another is active vibration control. For the passive vibration control, a parallel mechanism is presented to increase the stiffness of robot machine; for the active vibration control, a hybrid control method combining feedforward controller and nonlinear feedback controller is introduced for chatter suppression. A dynamic model and its chatter vibration phenomena of a hybrid robot is demonstrated. Simulation results are given based on the proposed hybrid robot machine which is developed for the ITER VV assembly and maintenance

Chatter suppression methods of a robot machine for ITER vacuum vessel assembly and maintenance

Energy Technology Data Exchange (ETDEWEB)

Wu, Huapeng; Wang, Yongbo, E-mail: yongbo.wang@lut.fi; Li, Ming; Al-Saedi, Mazin; Handroos, Heikki

2014-10-15

Highlights: •A redundant 10-DOF serial-parallel hybrid robot for ITER assembly and maintains is presented. •A dynamic model of the robot is developed. •A feedback and feedforward controller is presented to suppress machining vibration of the robot. -- Abstract: In the process of assembly and maintenance of ITER vacuum vessel (ITER VV), various machining tasks including threading, milling, welding-defects cutting and flexible hose boring are required to be performed from inside of ITER VV by on-site machining tools. Robot machine is a promising option for these tasks, but great chatter (machine vibration) would happen in the machining process. The chatter vibration will deteriorate the robot accuracy and surface quality, and even cause some damages on the end-effector tools and the robot structure itself. This paper introduces two vibration control methods, one is passive and another is active vibration control. For the passive vibration control, a parallel mechanism is presented to increase the stiffness of robot machine; for the active vibration control, a hybrid control method combining feedforward controller and nonlinear feedback controller is introduced for chatter suppression. A dynamic model and its chatter vibration phenomena of a hybrid robot is demonstrated. Simulation results are given based on the proposed hybrid robot machine which is developed for the ITER VV assembly and maintenance.

MHD equilibrium methods for ITER [International Thermonuclear Experimental Reactor] PF [poloidal field] coil design and systems analysis

International Nuclear Information System (INIS)

Strickler, D.J.; Galambos, J.D.; Peng, Y.K.M.

1989-03-01

Two versions of the Fusion Engineering Design Center (FEDC) free-boundary equilibrium code designed to computer the poloidal field (PF) coil current distribution of elongated, magnetically limited tokamak plasmas are demonstrated and applied to the systems analysis of the impact of plasma elongation on the design point of the International Thermonuclear Experimental Reactor (ITER). These notes were presented at the ITER Specialists' Meeting on the PF Coil System and Operational Scenario, held at the Max Planck Institute for Plasma Physics in Garching, Federal Republic of Germany, May 24--27, 1988. 8 refs., 6 figs., 4 tabs

Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

International Nuclear Information System (INIS)

Turner, Andrew; Pampin, Raul; Loughlin, M.J.; Ghani, Zamir; Hurst, Gemma; Lo Bue, Alessandro; Mangham, Samuel; Puiu, Adrian; Zheng, Shanliang

2014-01-01

The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge

Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

Energy Technology Data Exchange (ETDEWEB)

Turner, Andrew, E-mail: andrew.turner@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pampin, Raul [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Loughlin, M.J. [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghani, Zamir; Hurst, Gemma [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lo Bue, Alessandro [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Mangham, Samuel [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Puiu, Adrian [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Zheng, Shanliang [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

2014-10-15

The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge.

Multi-scenario evaluation and specification of electromagnetic loads on ITER vacuum vessel

International Nuclear Information System (INIS)

Rozov, Vladimir; Martinez, J.-M.; Portafaix, C.; Sannazzaro, G.

2014-01-01

Highlights: • We present the results of multi-scenario analysis of EM loads on ITER vacuum vessel (VV). • The differentiation of models provides the economic way to perform big amount of calculations. • Functional approximation is proposed for distributed data/FE/numerical results specification. • Examples of specification of the load profiles by trigonometric polynomials (DHT) are given. • Principles of accounting for toroidal asymmetry at EM interactions in tokamak are considered. - Abstract: The electro-magnetic (EM) transients cause mechanical forces, which represent one of the most critical loads for the ITER vacuum vessel (VV). The paper is focused on the results of multi-scenario analysis and systematization of these EM loads, including specifically addressed pressures on shells and the net vertical force. The proposed mathematical model and computational technology, based on the use of integral parameters and operational analysis methods, enabled qualitative and quantitative analysis of the problem, time-efficient computations and systematic assessment of a large number of scenarios. The obtained estimates, found envelopes and peak values exemplify the principal loads on the VV and provide a database to support engineering load specifications. Special attention is given to the challenge of specification and documenting of the results in a form, suitable for using the data in engineering applications. The practical aspects of specification of distributed data, such as experimental and finite-element (FE) results, by analytical interpolants are discussed. The example of functional approximation of the load profiles by trigonometric polynomials based on discrete Hartley transform (DHT) is given

Multi-scenario evaluation and specification of electromagnetic loads on ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Rozov, Vladimir, E-mail: vladimir.rozov@iter.org; Martinez, J.-M.; Portafaix, C.; Sannazzaro, G.

2014-10-15

Highlights: • We present the results of multi-scenario analysis of EM loads on ITER vacuum vessel (VV). • The differentiation of models provides the economic way to perform big amount of calculations. • Functional approximation is proposed for distributed data/FE/numerical results specification. • Examples of specification of the load profiles by trigonometric polynomials (DHT) are given. • Principles of accounting for toroidal asymmetry at EM interactions in tokamak are considered. - Abstract: The electro-magnetic (EM) transients cause mechanical forces, which represent one of the most critical loads for the ITER vacuum vessel (VV). The paper is focused on the results of multi-scenario analysis and systematization of these EM loads, including specifically addressed pressures on shells and the net vertical force. The proposed mathematical model and computational technology, based on the use of integral parameters and operational analysis methods, enabled qualitative and quantitative analysis of the problem, time-efficient computations and systematic assessment of a large number of scenarios. The obtained estimates, found envelopes and peak values exemplify the principal loads on the VV and provide a database to support engineering load specifications. Special attention is given to the challenge of specification and documenting of the results in a form, suitable for using the data in engineering applications. The practical aspects of specification of distributed data, such as experimental and finite-element (FE) results, by analytical interpolants are discussed. The example of functional approximation of the load profiles by trigonometric polynomials based on discrete Hartley transform (DHT) is given.

Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

International Nuclear Information System (INIS)

Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

1992-09-01

Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

Weld distortion prediction and control of the ITER vacuum vessel manufacturing mock-ups

International Nuclear Information System (INIS)

Ottolini, Marco; Barbensi, Andrea

2014-01-01

The fabrication of the ITER Vacuum Vessel Sectors is an unprecedented challenge, due to their dimensions, the close tolerances, the complex 'D' shape. The technological issues were faced by the production of full scale mock ups to confirm the manufacturing feasibility to achieve very tight tolerances and qualify the main manufacturing processes, by a step by step welding distortion control, by the qualification of not conventional NDT inspection techniques and by innovative 3D dimensional inspections. The Supplier is required to fabricate at least two mock ups, inboard and outboard, related to the manufacturing method of the VV Sectors, to demonstrate the control of the welding distortions to achieve tolerances, optimizing welding sequences and calibrating of welding distortions computer simulations. The stages of this preparatory activity are: prediction of welding distortion for fabrication mock ups representative of selected segments; demonstration that distortion predictions are consistent with experimental results from 3D dimensional inspection; understanding of reasons of possible deviations between numerical and experimental results and definition of action to solve these issues; demonstration that possible calculation simplifications, adopted to speed up the analysis process, do not affect significantly the welding distortion prediction. This paper describes the weld distortion prediction and control on the manufacturing mock-ups of ITER Vacuum Vessel Sectors, with particular emphasis to the lessons learned. (authors)

Structural Analysis for an Upper Port of the ITER Vacuum Vessel

International Nuclear Information System (INIS)

Yun-Seok Hong; Kwon, T. K.; Ahn, H. J.; Kim, Y.K.; Lee, C.D.

2006-01-01

The ITER vacuum vessel (VV) has numerous openings for the port structures including upper, equatorial, and lower ports used for equipment installation, utility feed through, vacuum pumping, and access into the vessel for maintenance. Every upper port, slanted upward slightly, has a trapezoidal/rectangular cross-section and consists of a port stub, a stub extension and a port extension with a connecting duct. To investigate the structural integrity and to increase the structural reliability of the VV and ports, the structural analyses of the upper port structure have been performed. The global structural analysis of the upper port with the in-port components has been carried out. The local analyses of a tangential key, an upper port flange, a connecting duct and a sealing unit have been performed. The design loads are dead weight, normal and abnormal pressure load, electromagnetic load, and seismic load in consideration of the dynamic amplification factors. The stress analyses were performed in a nonlinear elastic approach taking into account the contact surface between port extension flange and port plug flange. Two advanced designs from the ITER international team have been reviewed. To verify the strength of the reinforcing ribs for the connecting duct and of the fastening/sealing units, the local analyses utilizing the sub-modeling technique have been performed. The ASME code and the ITER design criteria were applied for the evaluation of the structural analysis results from the global and local analyses. The clearance between a port and a plug to accommodate the plug deformation has been assessed. The upper port flange based on the original design could withstand design loads, but there could be a gap on the flange surface under the design condition. The modified flange design, which is under the bolt friction only without tangential key was proposed. The deflection of the plug for an advanced design with a removable flange is higher than that for the original

Singular point analysis during rail deployment into vacuum vessel for ITER blanket maintenance

International Nuclear Information System (INIS)

Kakudate, Satoshi; Shibanuma, Kiyoshi

2007-05-01

Remote maintenance of the ITER blanket composed of about 400 modules in the vessel is required by a maintenance robot due to high gamma radiation of ∼500Gy/h in the vessel. A concept of rail-mounted vehicle manipulator system has been developed to apply to the maintenance of the ITER blanket. The most critical issue of the vehicle manipulator system is the feasibility of the deployment of the articulated rail composed of eight rail links into the donut-shaped vessel without any driving mechanism in the rail. To solve this issue, a new driving mechanism and procedure for the rail deployment has been proposed, taking account of a repeated operation of the multi-rail links deployed in the same kinematical manner. The new driving mechanism, which is deferent from those of a usual 'articulated arm' equipped with actuator in the every joint for movement, is composed of three mechanisms. To assess the feasibility of the kinematics of the articulated rail for rail deployment, a kinematical model composed of three rail links related to a cycle of the repeated operation for rail deployment was considered. The determinant det J' of the Jacobian matrix J' was solved so as to estimate the existence of a singular point of the transformation during rail deployment. As a result, it is found that there is a singular point due to det J'=0. To avoid the singular point of the rail links, a new location of the second driving mechanism and the related rail deployment procedure are proposed. As a result of the rail deployment test based on the new proposal using a full-scale vehicle manipulator system, the respective rail links have been successfully deployed within 6 h less than the target of 8 h in the same manner of the repeated operation under a synchronized cooperation among the three driving mechanisms. It is therefore concluded that the feasibility of the rail deployment of the articulated rail composed of simple structures without any driving mechanism has been demonstrated

Manufacturing and maintenance technologies developed for a thick-wall structure of the ITER vacuum vessel

International Nuclear Information System (INIS)

Onozuka, M.; Alfile, J.P.; Aubert, Ph.; Dagenais, J.-F.; Grebennikov, D.; Ioki, K.; Jones, L.; Koizumi, K.; Krylov, V.; Maslakowski, J.; Nakahira, M.; Nelson, B.; Punshon, C.; Roy, O.; Schreck, G.

2001-01-01

Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems have been carried out for on-site manufacturing and maintenance of the thick-wall structure of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV). Conventional techniques, including tungsten inert gas welding, plasma cutting, and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated, including reduced-pressure electron-beam and multi-pass neodymium-doped yttrium aluminum garnet (NdYAG) laser welding, NdYAG laser cutting, and electro-magnetic acoustic transducer inspection, to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible

Manufacturing and maintenance technologies developed for a thick-wall structure of the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. E-mail: onozukm@itereu.de; Alfile, J.P.; Aubert, Ph.; Dagenais, J.-F.; Grebennikov, D.; Ioki, K.; Jones, L.; Koizumi, K.; Krylov, V.; Maslakowski, J.; Nakahira, M.; Nelson, B.; Punshon, C.; Roy, O.; Schreck, G

2001-09-01

Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems have been carried out for on-site manufacturing and maintenance of the thick-wall structure of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV). Conventional techniques, including tungsten inert gas welding, plasma cutting, and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated, including reduced-pressure electron-beam and multi-pass neodymium-doped yttrium aluminum garnet (NdYAG) laser welding, NdYAG laser cutting, and electro-magnetic acoustic transducer inspection, to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible.

Design and R and D for the ITER vacuum vessel

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Iizuka, T.; Parker, R.; Koizumi, K.; Kuzmin, E.; Maisonnier, D.; Nelson, B.

1998-01-01

The current design and key R and D results for the Vacuum Vessel (VV) for the International Thermonuclear Experimental Reactor (ITER) are presented. During the past two years the basic VV design has remained unchanged. Additional details have been defined in key areas and recent R and D results have indicated where further improvements can be made. R and D results have also confirmed the feasibility of important aspects of the design such as limiting weld distortions to acceptable levels and achieving required tolerances with a large welded structure. Recent design progress includes the development of a structural design strategy for the VV, modification of the inboard structure, employment of ferromagnetic material between the VV shells, and confirmation of the cooling characteristics for the VV. This report presents the current design and how it has been affected by R and D results. (authors)

Design and R and D for the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Ioki, K.; Johnson, G.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Iizuka, T.; Parker, R. [ITER Joint Work Site, Garching (Germany); Koizumi, K. [Japan Atomic Energy Research Inst., Naka (Japan); Kuzmin, E. [Efremov Insitute, Saint Petersburg (Russian Federation); Maisonnier, D. [NET Team, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States)

1998-07-01

The current design and key R and D results for the Vacuum Vessel (VV) for the International Thermonuclear Experimental Reactor (ITER) are presented. During the past two years the basic VV design has remained unchanged. Additional details have been defined in key areas and recent R and D results have indicated where further improvements can be made. R and D results have also confirmed the feasibility of important aspects of the design such as limiting weld distortions to acceptable levels and achieving required tolerances with a large welded structure. Recent design progress includes the development of a structural design strategy for the VV, modification of the inboard structure, employment of ferromagnetic material between the VV shells, and confirmation of the cooling characteristics for the VV. This report presents the current design and how it has been affected by R and D results. (authors)

Remote maintenance development for ITER

International Nuclear Information System (INIS)

Tada, Eisuke; Shibanuma, Kiyoshi

1998-01-01

This paper describes the overall ITER remote maintenance design concept developed mainly for in-vessel components such as diverters and blankets, and outlines the ITER R and D program to develop remote handling equipment and radiation hard components. Reactor structures inside the ITER cryostat must be maintained remotely due to DT operation, making remote handling technology basic to reactor design. The overall maintenance scenario and design concepts have been developed, and maintenance design feasibility, including fabrication and testing of full-scale in-vessel remote maintenance handling equipment and tool, is being verified. (author)

Remote maintenance development for ITER

Energy Technology Data Exchange (ETDEWEB)

Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Shibanuma, Kiyoshi

1998-04-01

This paper describes the overall ITER remote maintenance design concept developed mainly for in-vessel components such as diverters and blankets, and outlines the ITER R and D program to develop remote handling equipment and radiation hard components. Reactor structures inside the ITER cryostat must be maintained remotely due to DT operation, making remote handling technology basic to reactor design. The overall maintenance scenario and design concepts have been developed, and maintenance design feasibility, including fabrication and testing of full-scale in-vessel remote maintenance handling equipment and tool, is being verified. (author)

Micro-particles in ITER: A comprehensive review

International Nuclear Information System (INIS)

Grisolia, C.; Rosanvallon, S.; Sharpe, Ph.; Winter, J.

2009-01-01

In a fusion reactor like ITER, in-vessel materials are subjected to interactions with the plasma. One of the main consequences of these plasma-material interactions is the creation of co-deposited layers. Due to internal stresses, part of these layers can crack leading to micro particle creation. The purpose of the following paper is to review the Tokamak operation processes which lead to erosion and layer creation. Then, the proportion of these layers that is converted into micro-particles will be evaluated in the case of Tore Supra experiments and extrapolated for ITER. It is major importance to measure the ITER mobilizable dusts present in the Vacuum Vessel and compare the measured quantity with the safety limits. When approaching these limits, removal systems must be used in order to control the in-vessel dust inventory. In the second part of the paper, diagnostics and removal system under development will be presented.

Follow-up Study of ITER Safety Analysis : Large In-vessel First Wall Pipe Break with Wet Confinement Bypass

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

Previous researches have been analyzed risk assessments of fusion reactors that are dangerous in the severe accidents where the radioactive material released from confinement building to the environment. To simulate the severe accidents in ITER, a number of thermal hydraulics simulation codes were used. Before construction of the fusion reactor, to obtain ITER license about safety issue, MELCOR is chosen as one of the several codes to be used to perform ITER safety analyses. Qualification of the simulation code is to simulate the cooling system in ITER, the transport of radionuclides during design basis accidents (DBAs) including beyond design basis accidents (BDBAs). MELCOR is fully integrated code that models the accidents in Light Water Reactor (LWR). To analyze the accidents in ITER, MELCOR 1.8.2 version is modified. In the nuclear fusion system, the amount of released radioactive material is criteria for safety permission. Tritium (or tritiated water: HTO) and radioactive dust aerosol are the source of radioactive leakage. In the Generic Site Safety Report (GSSR) for the ITER plant, Table I lists the release guidelines for tritium and activation products for normal operation, incidents and accidents. Several accident analyses have been studied to know how much radioactive material could be released from the severe accidents. In the present work, The MELCOR input deck of large First Wall (FW) coolant leak (pipe break) is used to study and radioactive material leakage thorough bypass accident are studied to follow up the ITER safety analysis. In this research, follow-up study of the in-vessel inboard/inboard-outboard FW pipe break was analyzed to investigate the amount of leakage of radioactive aerosol. All of the accident cases released the lower amount of radioactive aerosol compared to the IAEA guide lines. In addition, the OBB pipe break made lower HTO aerosol leakage because of condensation of HTO and adsorption between coolant and aerosol.

Design improvements and R and D achievements for VV and in-vessel components towards ITER construction and implications for the R and D programme

International Nuclear Information System (INIS)

Ioki, K.

2002-01-01

Procurement specifications are now being finalised for ITER components whose construction is lengthy, yet which are needed early, such as the vacuum vessel. Although the basic concept of the vacuum vessel (VV) and in-vessel components of the ITER design has stayed the same as reported at the last conference, there have been several detailed design improvements resulting from efforts to raise reliability, to improve better maintainability and to save money. One of the most important achievements in the VV R and D is demonstration of the necessary assembly tolerances. Further development of advanced methods of cutting, welding and NDT for the VV have been continued in order to refine manufacturing and improve cost and technical performance. With regard to the related FW/blanket and divertor designs, the R and D has resulted in the development of suitable technologies. Prototypes of the FW panel, the blanket shield block and the divertor components have been successfully fabricated. This paper reviews the recent progress in the design as procurement nears. (author)

Assessment of the dynamic behaviours of the ITER Vacuum Vessel

Energy Technology Data Exchange (ETDEWEB)

Blocki, J., E-mail: jacek.blocki@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Combescure, D. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Mazzone, G. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

Highlights: ► The cyclic symmetry structure with special boundary conditions has been analyzed. ► Results based on the FE solid model and on the FE shell model have been compared. ► The effect of the missing mass contained has been checked. -- Abstract: The dynamic behaviour of the ITER Vacuum Vessel (VV) under seismic loads will be assessed by carrying out the modal analysis and then by applying the response spectrum method which describes earthquake motions. The effect of the missing mass is included in this last analysis. Numerical results are based on two different Finite Element (FE) models and on three different methods by which natural frequencies and mode shapes are defined. It means, it is applied the cyclic symmetry method, the component mode synthesis method and the 360° FE model of the VV. Comparisons between obtained results for the different models and methods are presented.

Design and implementation of motion planning of inspection and maintenance robot for ITER-like vessel

Energy Technology Data Exchange (ETDEWEB)

Wang, Hesheng; Lai, Yinping [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Cao, Qixin [Institute of Robotics, Shanghai Jiao Tong University, Shanghai 200240 (China)

2015-12-15

Robot motion planning is a fundamental problem to ensure the robot executing the task without clashes, fast and accurately in a special environment. In this paper, a motion planning of a 12 DOFs remote handling robot used for inspecting the working state of the ITER-like vessel and maintaining key device components is proposed and implemented. Firstly, the forward and inverse kinematics are given by analytic method. The work space and posture space of this manipulator are both considered. Then the motion planning is divided into three stages: coming out of the cassette mover, moving along the in-vessel center line, and inspecting the D-shape section. Lastly, the result of experiments verified the performance of the motion design method. In addition, the task of unscrewing/screwing the screw demonstrated the feasibility of system in function.

Investigation of linearity of the ITER outer vessel steady-state magnetic field sensors at high temperature

Science.gov (United States)

Entler, S.; Duran, I.; Kocan, M.; Vayakis, G.

2017-07-01

Three vacuum vessel sectors in ITER will be instrumented by the outer vessel steady-state magnetic field sensors. Each sensor unit features a pair of metallic Hall sensors with a sensing layer made of bismuth to measure tangential and normal components of the local magnetic field. The influence of temperature and magnetic field on the Hall coefficient was tested for the temperature range from 25 to 250 oC and the magnetic field range from 0 to 0.5 T. A fit of the Hall coefficient normalized temperature function independent of magnetic field was found, and a model of the Hall coefficient functional dependence at a wide range of temperature and magnetic field was built with the purpose to simplify the calibration procedure.

In-vessel tritium retention and removal in ITER

International Nuclear Information System (INIS)

Federici, G.; Anderl, R.A.

1998-01-01

The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world's fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

In-vessel tritium retention and removal in ITER

Energy Technology Data Exchange (ETDEWEB)

Federici, G. [ITER JWS Garching Co-Center (Germany); Anderl, R.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Andrew, P. [JET Joint Undertaking, Abingdon (United Kingdom)] [and others

1998-06-01

The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

Application of remote handling compatibility on ITER plant

International Nuclear Information System (INIS)

Sanders, S.; Rolfe, A.; Mills, S.F.; Tesini, A.

2011-01-01

The ITER plant will require fully remote maintenance during its operational life. For this to be effective, safe and efficient the plant will have to be developed in accordance with remote handling (RH) compatibility requirements. A system for ensuring RH compatibility on plant designed for Tokamaks was successfully developed and applied, inter alia, by the authors when working at the JET project. The experience gained in assuring RH compatibility of plant at JET is now being applied to RH relevant ITER plant. The methodologies required to ensure RH compatibility of plant include the standardization of common plant items, standardization of RH features, availability of common guidance on RH best practice and a protocol for design and interface review and approval. The protocol in use at ITER is covered by the ITER Remote Maintenance Management System (IRMMS) defines the processes and utilization of management controls including Plant Definition Forms (PDF), Task Definition Forms (TDFs) and RH Compatibility Assessment Forms (RHCA) and the ITER RH Code of Practice. This paper will describe specific examples where the authors have applied the methodology proven at JET to ensure remote handling compatibility on ITER plant. Examples studied are: ·ELM coils (to be installed in-vessel behind the Blanket Modules) - handling both in-vessel, in Casks and at the Hot Cell as well as fully remote installation and connection (mechanical and electrical) in-vessel. ·Neutral beam systems (in-vessel and in the NB Cell) - beam sources, cesium oven, beam line components (accessed in the NB Cell) and Duct Liner (remotely replaced from in-vessel). ·Divertor (in-vessel) - cooling pipe work and remotely operated electrical connector. The RH compatibility process can significantly affect plant design. This paper should therefore be of interest to all parties who develop ITER plant designs.

Qualification of phased array ultrasonic examination on T-joint weld of austenitic stainless steel for ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, G.H. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Park, C.K., E-mail: love879@hanmail.net [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Jin, S.W.; Kim, H.S.; Hong, K.H.; Lee, Y.J.; Ahn, H.J.; Chung, W. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Jung, Y.H.; Roh, B.R. [Hyundai Heavy Industries Co. Ltd., Ulsan 682-792 (Korea, Republic of); Sa, J.W.; Choi, C.H. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2016-11-01

Highlights: • PAUT techniques has been developed by Hyundai Heavy Industries Co., LTD (HHI) and Korea Domestic Agency (KODA) to verify and settle down instrument calibration, test procedures, image processing, and so on. As the first step of development for PAUT technique, Several dozens of qualification blocks with artificial defects, which are parallel side drilled hole, embedded lack of fusion, embedded repair weld notch, and so on, have been designed and fabricated to simulate all potential defects during welding process. Real UT qualification group-1 for T-joint weld was successfully conducted in front of ANB inspector. • In this paper, remarkable progresses of UT qualification are presented for ITER vacuum vessel. - Abstract: Full penetration welding and 100% volumetric examination are required for all welds of pressure retaining parts of the ITER Vacuum Vessel (VV) according to RCC-MR Code and French Order of Nuclear Pressure Equipment (ESPN). The NDE requirement is one of important technical issues because radiographic examination (RT) is not applicable to many welding joints. Therefore the ultrasonic examination (UT) has been selected as an alternative method. Generally the UT on the austenitic welds is regarded as a great challenge due to the high attenuation and dispersion of the ultrasonic signal. In this paper, Phased array ultrasonic examination (PAUT) has been introduced on double sided T-shape austenitic welds of the ITER VV as a major NDE method as well as RT. Several dozens of qualification blocks with artificial defects, which are parallel side drilled hole, embedded lack of fusion, embedded repair weld notch, embedded parallel vertical notch, and so on, have been designed and fabricated to simulate all potential defects during welding process. PAUT techniques on the thick austenitic welds have been developed taking into account the acceptance criteria. Test procedure including calibration of equipment is derived and qualified through

Numerical method for partial equilibrium flow

International Nuclear Information System (INIS)

Ramshaw, J.D.; Cloutman, L.D.; Los Alamos, New Mexico 87545)

1981-01-01

A numerical method is presented for chemically reactive fluid flow in which equilibrium and nonequilibrium reactions occur simultaneously. The equilibrium constraints on the species concentrations are established by a quadratic iterative procedure. If the equilibrium reactions are uncoupled and of second or lower order, the procedure converges in a single step. In general, convergence is most rapid when the reactions are weakly coupled. This can frequently be achieved by a judicious choice of the independent reactions. In typical transient calculations, satisfactory accuracy has been achieved with about five iterations per time step

A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

International Nuclear Information System (INIS)

Nakahira, M.; Takeda, N.; Kakudate, S.; Onozuka, M.

2008-01-01

The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV

Structural damages prevention of the ITER vacuum vessel and ports by elasto-plastic analysis with regards to RCC-MR

Energy Technology Data Exchange (ETDEWEB)

Martinez, Jean-Marc, E-mail: jean-marc.martinez@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Jun, Chang Hoon; Portafaix, Christophe; Alekseev, Alexander; Sborchia, Carlo; Choi, Chang-Ho [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Albin, Vincent [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France); Borrelly, Stephane [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Cambazar, Magali [Assystem EOS, 117 rue Jacquard, 84120 Pertuis (France); Gaucher, Thomas [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France); Sfarni, Samir; Tailhardat, Olivier [Assystem EOS, 117 rue Jacquard, 84120 Pertuis (France)

2015-10-15

Highlights: • ITER vacuum vessel (VV) is a part of the first barrier to confine the plasma. • ITER VV as NPE necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, and conformance testing and quality assurance, i.e. ANB. • Several types of damages have to be prevented in order to guarantee the structural integrity with regards to RCC-MR. • It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. • Several structural analyses were performed with many different global and local models of the whole ITER VV. - Abstract: Several types of damages have to be prevented in order to guarantee the structural integrity of a structure with regards to RCC-MR; the P-type damages which can result from the application to a structure of a steadily and regularly increasing loading or a constant loading and the S-type damages during operational loading conditions which can only result from repeated application of loadings associated to the progressive deformations and fatigue. Following RCC-MR, the S-type damages prevention has to be started only when the structural integrity is guaranteed against P-type damages. The verification of the last one on the ITER vacuum vessel and ports has been performed by limit analysis with elasto-(perfectly)plastic material behavior. It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. Some elasto-plastic analyses have been performed considering several cyclic loadings to evaluate also more realistic structural margins of the against S-type damages.

A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

Energy Technology Data Exchange (ETDEWEB)

Nakahira, M. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)], E-mail: nakahira.masataka@jaea.go.jp; Takeda, N.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Onozuka, M. [Mitsubishi Nuclear Energy Systems, Inc., 1700K Street NW, Suite 440, Washington, DC 20006 (United States)

2008-12-15

The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV.

Development of radiation hardness components for ITER remote maintenance

Energy Technology Data Exchange (ETDEWEB)

Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yagi, Toshiaki; Morita, Yousuke

1998-04-01

In the ITER, in-vessel remote handling is required to assemble and maintain in-vessel components in DT operations. Since in-vessel remote handling systems must operate under intense gamma ray radiation exceeding 30 kGy/h, their components must have sufficiently high radiation hardness to allow maintenance long enough in ITER in-vessel environments. Thus, extensive radiation tests and quality improvement, including optimization of material compositions, have been conducted through the ITER R and D program to develop radiation hardness components that meet radiation doses from 10 to 100 MGy at 10 kGy/h. This paper presents the latest on radiation hardness component development conducted by the Japan Home Team as a contribution to the ITER. The remote handling components tested are categorized for use in robotic or viewing systems, or as common components. Radiation tests have been conducted on commercially available products for screening, on modified products, and on new products to improve the radiation hardness. (author)

Development of radiation hardness components for ITER remote maintenance

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; Ito, Akira; Yagi, Toshiaki; Morita, Yousuke

1998-01-01

In the ITER, in-vessel remote handling is required to assemble and maintain in-vessel components in DT operations. Since in-vessel remote handling systems must operate under intense gamma ray radiation exceeding 30 kGy/h, their components must have sufficiently high radiation hardness to allow maintenance long enough in ITER in-vessel environments. Thus, extensive radiation tests and quality improvement, including optimization of material compositions, have been conducted through the ITER R and D program to develop radiation hardness components that meet radiation doses from 10 to 100 MGy at 10 kGy/h. This paper presents the latest on radiation hardness component development conducted by the Japan Home Team as a contribution to the ITER. The remote handling components tested are categorized for use in robotic or viewing systems, or as common components. Radiation tests have been conducted on commercially available products for screening, on modified products, and on new products to improve the radiation hardness. (author)

Design and development of ITER high-frequency magnetic sensor

Energy Technology Data Exchange (ETDEWEB)

Ma, Y., E-mail: Yunxing.Ma@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Fircroft Engineering, Lingley House, 120 Birchwood Point, Birchwood Boulevard, Warrington, WA3 7QH (United Kingdom); Vayakis, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Begrambekov, L.B. [National Research Nuclear University (MEPhI), 115409, Moscow, Kashirskoe shosse 31 (Russian Federation); Cooper, J.-J. [Culham Centre for Fusion Energy (CCFE), Abingdon, Oxfordshire OX14 3DB (United Kingdom); Duran, I. [IPP Prague, Za Slovankou 1782/3, 182 00 Prague 8 (Czech Republic); Hirsch, M.; Laqua, H.P. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Moreau, Ph. [CEA Cadarache, 13108 Saint Paul lez Durance Cedex (France); Oosterbeek, J.W. [Eindhoven University of Technology (TU/e), PO Box 513, 5600 MB Eindhoven (Netherlands); Spuig, P. [CEA Cadarache, 13108 Saint Paul lez Durance Cedex (France); Stange, T. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Walsh, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2016-11-15

Highlights: • ITER high-frequency magnetic sensor system has been designed. • Prototypes have been successfully manufactured. • Manufactured prototypes have been tested in various labs. • Test results experimentally validated the design. - Abstract: High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in a harsh environment with considerable neutron/nuclear radiation and high thermal load. Essential components of the HF sensor system, including inductive coil, electron cyclotron heating (ECH) shield, electrical cabling and termination load, have been designed to meet ITER measurement requirements. System performance (e.g. frequency response, thermal conduction) has been assessed. A prototyping campaign was initiated to demonstrate the manufacturability of the designed components. Prototypes have been produced according to the specifications. A series of lab tests have been performed to examine assembly issues and validate electrical and thermo-mechanical aspects of the design. In-situ microwave radiation test has been conducted in the MISTRAL test facility at IPP-Greifswald to experimentally examine the microwave shielding efficiency and structural integrity of the ECH shield. Low-power microwave attenuation measurement and scanning electron microscopic inspection were conducted to probe and examine the quality of the metal coating on the ECH shield.

Alternatives of ITER vacuum vessel support system

International Nuclear Information System (INIS)

Ohmori, Junji; Kitamura, Kazunori; Araki, Masanori; Ohno, Isamu; Shoji, Teruaki

2002-07-01

Optional designs of vacuum vessel (VV) support have been performed for the International Thermonuclear Experimental Reactor (ITER) to reduce stresses and buckling concern of the flexible plate structure in ITER-FDR. One of the optional designs is hanging type VV support concept that consists of top hanging supports at the top of VV and middle radial stoppers in the middle of outboard VV. The hanging supports are located at the toroidal field (TF) coil inboard top region (R∼5400 mm) using the narrow window space surrounded by a poloidal field coil (PF1) and TF coil. The radial stoppers are located inside TF coil cases in the TF coil outboard middle region (R∼9300 mm). The upper flange connection of the radial stoppers should slide in vertical direction to eliminate thermal stress produced by relative thermal displacement between VV wall and TF coil case. Both supports consist of flexible plates and are mounted on 18 locations in toroidal direction. The radial and toroidal reaction forces are shared with the hanging supports and the radial stoppers. However, the vertical force is sustained by only the hanging supports. The others are compressive type support concept that consists of nine VV supports located in alternate divertor port regions in toroidal direction. Two designs have been performed for the VV support concept. One is mounted on TF inter-coil structures (OIS) and the other is on cryostat ring. The compressive support on TF coil OIS is dependent on TF coil movement but that on cryostat is independent. In the optional designs, the bending stress due to the relative thermal displacement between TF coil and VV is classified to primary stress according to ASME Sec. III NF. The stress due to TF coil displacement is also considered as primary stress. The stress due to non-uniform temperature distribution of the flexible plate is classified to secondary stress. The preliminary structural assessments for the optional designs have been performed for all load

Structural analysis of support structure for ITER vacuum vessel

International Nuclear Information System (INIS)

Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka

2004-12-01

ITER vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed. This independent concept has two advantages: (1) thermal load due to the temperature deference between VV and the lower temperature components such as TF coil becomes lower and (2) the other components such as TF coil is categorized as a non-safety component because of its independence from VV. Stress analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support. (author)

Feasibility studies on plasma vertical position control by ex-vessel coils in ITER-like tokamak fusion reactors

International Nuclear Information System (INIS)

Nishio, Satoshi; Sugihara, Masayoshi; Shimomura, Yasuo

1993-01-01

Feasibility of the plasma vertical position control by control coils installed outside the vacuum vessel (ex-vessel) in a tokamak fusion reactor is examined for an ITER-like device. When a pair of ex-vessel control coils is made of normal conductor material and located near the outmost superconducting (SC) poloidal field (PF) coils, the applied voltage of several hundred volts on the control coils is the maximum allowable value which is limited by the maximum allowable induced voltage and eddy current heating on the SC PF coils, under the conditions that the SC PF coils are connected in series and a partitioning connection is employed for each of these PF coils. A proportional and derivative (PD) controller with and without voltage limitation has been employed to examine the feasibility. Indices of settling time and overshoot are introduced to measure the controllability of the control system. Based on these control schemes and indices, higher elongation (κ=2) and moderate elongation (κ=1.6) plasmas are examined for normal and deteriorated (low beta value and peaked current profile) plasma conditions within the restriction of applied voltage and current of control coils. The effect of the time constant of the passive stabilizer is also examined. The major results are: (1) A plasma with an elongation of 2.0 inevitably requires a passive stabilizer close to the plasma surface, (2) in case of a higher elongation than κ=2, even the ex-vessel control coil system is marginally controllable under normal plasma conditions, while it is difficult to control the deteriorated plasma conditions, (3) the time constant of the passive stabilizer is not an essential parameter for the controllability, (4) when the elongation is reduced down to 1.6, the ex-vessel control coil system can control the plasma even under deteriorated plasma conditions. (orig.)

ITER EDA newsletter. V. 5, no. 9

International Nuclear Information System (INIS)

1996-09-01

This issue of the Newsletter on the Engineering Design Activities (EDA) for the ITER project contains an overview of one of the seven large ITER Research and Development Projects identified by the ITER Director, namely the Vacuum Vessel Sector, as well as an account of computer animation created for ITER

Design of the ITER Tokamak Assembly Tools

International Nuclear Information System (INIS)

Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

2006-01-01

ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

Eliminating Islands in High-pressure Free-boundary Stellarator Magnetohydrodynamic Equilibrium Solutions

International Nuclear Information System (INIS)

Hudson, S.R.; Monticello, D.A.; Reiman, A.H.; Boozer, A.H.; Strickler, D.J.; Hirshman, S.P.; Zarnstorff, M.C.

2002-01-01

Magnetic islands in free-boundary stellarator equilibria are suppressed using a procedure that iterates the plasma equilibrium equations and, at each iteration, adjusts the coil geometry to cancel resonant fields produced by the plasma. The coils are constrained to satisfy certain measures of engineering acceptability and the plasma is constrained to ensure kink stability. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible. The method is applied with success to a candidate plasma and coil design for the National Compact Stellarator eXperiment [Physics of Plasma, 7 (2000) 1911

Disruptions, loads, and dynamic response of ITER

International Nuclear Information System (INIS)

Nelson, B.; Riemer, B.; Sayer, R.; Strickler, D.; Barabaschi, P.; Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

1995-01-01

Plasma disruptions and the resulting electromagnetic loads are critical to the design of the vacuum vessel and in-vessel components of the International Thermonuclear Experimental Reactor (ITER). This paper describes the status of plasma disruption simulations and related analysis, including the dynamic response of the vacuum vessel and in-vessel components, stresses and deflections in the vacuum vessel, and reaction loads in the support structures

Radwaste management aspects of the test blanket systems in ITER

Energy Technology Data Exchange (ETDEWEB)

Laan, J.G. van der, E-mail: JaapG.vanderLaan@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Canas, D. [CEA, DEN/DADN, centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Chaudhari, V. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Iseli, M. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Kawamura, Y. [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Petit, P. [European Commission, DG ENER, Brussels (Belgium); Pitcher, C.S.; Torcy, D. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Ugolini, D. [Fusion for Energy, Barcelona (Spain); Zhang, H. [China Nuclear Energy Industry Corporation, Beijing 100032 (China)

2016-11-01

Highlights: • Test Blanket Systems are operated in ITER to test tritium breeding technologies. • The in-vessel parts of TBS become radio-active during the ITER nuclear phase. • For each TBM campaign the TBM, its shield and the Pipe Forests are removed. • High tritium contents and novel materials are specific TBS radwaste features. • A preliminary assessment confirmed RW routing, provided its proper conditioning. - Abstract: Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment.

Remote maintenance development for ITER

International Nuclear Information System (INIS)

Tada, Eisuke; Shibanuma, Kiyoshi

1997-01-01

This paper both describes the overall design concept of the ITER remote maintenance system, which has been developed mainly for use with in-vessel components such as divertor and blanket, and outlines of the ITER R and D program, which has been established to develop remote handling equipment/tools and radiation hard components. In ITER, the reactor structures inside cryostat have to be maintained remotely because of activation due to DT operation. Therefore, remote-handling technology is fundamental, and the reactor-structure design must be made consistent with remote maintainability. The overall maintenance scenario and design concepts of the required remote handling equipment/tools have been developed according to their maintenance classification. Technologies are also being developed to verify the feasibility of the maintenance design and include fabrication and testing of a fullscale remote-handling equipment/tools for in-vessel maintenance. (author)

Applicability assessment of plug weld to ITER vacuum vessel by crack propagation analysis

International Nuclear Information System (INIS)

Ohmori, Junji; Nakahira, Masataka; Takeda, Nobukazu; Shibanuma, Kiyoshi; Sago, Hiromi; Onozuka, Masanori

2006-03-01

In order to improve the fabricability of the vacuum vessel (VV) of International Thermonuclear Experimental Reactor (ITER), applicability of plug weld between VV outer shell and stiffening ribs/blanket support housings has been assessed using crack propagation analysis for the plug weld. The ITER VV is a double-wall structure of inner and outer shells with ribs and housings between the shells. For the fabrication of VV, ribs and housings are welded to outer shell after welding to inner shell. A lot of weld grooves should be adjusted for welding outer shell. The plug weld is that outer shells with slit at the weld region are set on ribs/housings then outer shells are welded to them by filling the slits with weld metal. The plug weld can allow larger tolerance of weld groove gap than ordinary butt weld. However, un-welded lengths parallel to outer sell surface remain in the plug weld region. It is necessary to evaluate the allowable un-welded length to apply the plug weld to ITER VV fabrication. For the assessment, the allowable un-welded lengths have been calculated by crack propagation analyses for load conditions, conservatively assuming the un-welded region is a crack. In the analyses, firstly allowable crack lengths are calculated from the stresses of the weld region. Then assuming initial crack length, crack propagation is calculated during operation period. Allowable initial crack lengths are determined on the condition that the propagated cracks should not exceed the allowable crack lengths. The analyses have been carried out for typical inboard straight region and inboard upper curved region with the maximum housing stress. The allowable initial cracks of ribs are estimated to be 8.8mm and 38mm for the rib and the housing, respectively, considering inspection error of 4.4mm. Plug weld between outer shell and ribs/housings could be applicable. (author)

The ITER remote maintenance system

International Nuclear Information System (INIS)

Tesini, A.; Palmer, J.

2007-01-01

ITER is a joint international research and development project that aims to demonstrate the scientific and technological feasibility of fusion power. As soon as the plasma operation begins using tritium, the replacement of the vacuum vessel internal components will need to be done with remote handling techniques. To accomplish these operations ITER has equipped itself with a Remote Maintenance System; this includes the Remote Handling equipment set and the Hot Cell facility. Both need to work in a cooperative way, with the aim of minimizing the machine shutdown periods and to maximize the machine availability. The ITER Remote Handling equipment set is required to be available, robust, reliable and retrievable. The machine components, to be remotely handle-able, are required to be designed simply so as to ease their maintenance. The baseline ITER Remote Handling equipment is described. The ITER Hot Cell Facility is required to provide a controlled and shielded area for the execution of repair operations (carried out using dedicated remote handling equipment) on those activated components which need to be returned to service, inside the vacuum vessel. The Hot Cell provides also the equipment and space for the processing and temporary storage of the operational and decommissioning radwaste. A conceptual ITER Hot Cell Facility is described. (orig.)

Development of radiation-hard electric connector with ball bearing for in-vessel remote maintenance equipment of ITER

International Nuclear Information System (INIS)

Ito, Akira; Obara, Kenjiro; Tada, Eisuke; Morita, Yousuke; Yagi, Toshiaki; Iida, Kazuhisa; Sato, Masaru.

1997-12-01

Development of radiation-hard electric connector with ball bearing for in-vessel remote maintenance equipment of ITER (International Thermonuclear Experimental Reactor) has been conducted. Since the in-vessel remote maintenance equipment is operated under the condition of 10 6 R/h gamma ray dose rate, the electric connector has to be radiation hard for an accumulation dose of 10 10 R. In addition, the simple attachment/removal mechanism is essential for remote operation. For this, the alumina (Al203) ceramics and a ball bearing were adopted to electric insulator and plug (male) of connector, respectively. The handling tests on attachment/removal of the connector were conducted by using master slave manipulator and general purpose robot with handling tool, and as a result, the validity of the attachment/removal mechanism was verified. In the gamma ray irradiation tests, which are under way, no degradation in break down voltage (1000V 1min.) up to 10 10 R was confirmed. However insulation resistance and contact resistance between contact pin and contact socket were deteriorated in proportion to the accumulation dose. Increase of contact resistance is considered due to an erosion of contact pin. (author)

Structural analysis of the ITER Divertor toroidal rails

Energy Technology Data Exchange (ETDEWEB)

Viganò, F., E-mail: Fabio.Vigano@LTCalcoli.it [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Escourbiac, F.; Gicquel, S.; Komarov, V. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Lucca, F. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ngnitewe, R. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy)

2013-10-15

The Divertor is one of the most technically challenging components of the ITER machine, which has the main function of extracting the power conducted in the scrape-off layer while maintaining the plasma purity. There are 54 Divertor cassettes installed in the vacuum vessel (VV). Each cassette body (CB) is fastened on the inner and outer concentric Divertor toroidal rails. The comprehensive assessment (in accordance with the Structural Design Criteria for ITER In-vessel Components: ITER SDC-IC) of the Divertor toroidal rails has been performed during design activity based on performing of thermal and stress analyses at operating conditions of neutron stage of ITER operation. This paper outlines the engineering aspects of the ITER Divertor toroidal rails and focuses on some critical regions of the present design highlighted by the performed structural assessment. The structural assessment has been performed with help of using Finite Element (FE) Abaqus code and based on criteria given by ITER SDC-IC.

A General Iterative Method of Fixed Points for Mixed Equilibrium Problems and Variational Inclusion Problems

Directory of Open Access Journals (Sweden)

Phayap Katchang

2010-01-01

Full Text Available The purpose of this paper is to investigate the problem of finding a common element of the set of solutions for mixed equilibrium problems, the set of solutions of the variational inclusions with set-valued maximal monotone mappings and inverse-strongly monotone mappings, and the set of fixed points of a family of finitely nonexpansive mappings in the setting of Hilbert spaces. We propose a new iterative scheme for finding the common element of the above three sets. Our results improve and extend the corresponding results of the works by Zhang et al. (2008, Peng et al. (2008, Peng and Yao (2009, as well as Plubtieng and Sriprad (2009 and some well-known results in the literature.

Iterative Schemes for Convex Minimization Problems with Constraints

Directory of Open Access Journals (Sweden)

Lu-Chuan Ceng

2014-01-01

Full Text Available We first introduce and analyze one implicit iterative algorithm for finding a solution of the minimization problem for a convex and continuously Fréchet differentiable functional, with constraints of several problems: the generalized mixed equilibrium problem, the system of generalized equilibrium problems, and finitely many variational inclusions in a real Hilbert space. We prove strong convergence theorem for the iterative algorithm under suitable conditions. On the other hand, we also propose another implicit iterative algorithm for finding a fixed point of infinitely many nonexpansive mappings with the same constraints, and derive its strong convergence under mild assumptions.

Challenging issues in the design and manufacturing of the European sectors of the ITER vacuum vessel

International Nuclear Information System (INIS)

Dans, Andres; Jucker, P.; Bayon, A.; Arbogast, J.-F.; Caixas, J.; Fernández, J.; Micó, G.; Pacheco, J.; Trentea, A.; Stamos, V.

2014-01-01

Highlights: • ITER Vacuum Vessel was described with its features and particularities. • Engineering and CAD design of Sector 5 is finish; the work of sectors 3 and 4 is ongoing. • Fabrication Mock Ups almost finished with an important know-how acquired. • Procurement of raw material (plates and forgings) started. • Qualification of welding, NDT and forming close to be finished. - Abstract: Fusion for Energy (F4E), the European Domestic Agency for the ITER project, has to supply seven sectors as part of the European contribution to the project. F4E signed the Procurement Agreement with ITER Organization (IO) in 2009. After a call for tender in 2010, the contract for the manufacturing of seven sectors was placed in October 2010 to a consortium of three Italian companies, Ansaldo, Mangiarotti and Walter Tosto (AMW). The first sector in the manufacturing route is Sector 5 (later will come 4, 3, 2, 9, 8, 7). This paper will cover: the status of the engineering activities, design, procurement and preparation to begin the manufacturing in 2013. Also will be presented the statutory and regulatory requirements of the French Nuclear Safety regulator and the status of the relevant R and D mock-ups to demonstrate manufacturing feasibility control of distortions (using predictions with analysis and algorithms to change in real time the manufacturing route in order to correct such distortions, inspectability and metrology). Another important aspect at this stage of the manufacturing is qualification of activities like welding, Non-destructive Examination and Hot Forming. This paper describes the status of the activities currently in process in order to meet with the challenging design, schedule and high quality requirements of the project

Challenging issues in the design and manufacturing of the European sectors of the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Dans, Andres, E-mail: andresdans@gmail.com; Jucker, P.; Bayon, A.; Arbogast, J.-F.; Caixas, J.; Fernández, J.; Micó, G.; Pacheco, J.; Trentea, A.; Stamos, V.

2014-10-15

Highlights: • ITER Vacuum Vessel was described with its features and particularities. • Engineering and CAD design of Sector 5 is finish; the work of sectors 3 and 4 is ongoing. • Fabrication Mock Ups almost finished with an important know-how acquired. • Procurement of raw material (plates and forgings) started. • Qualification of welding, NDT and forming close to be finished. - Abstract: Fusion for Energy (F4E), the European Domestic Agency for the ITER project, has to supply seven sectors as part of the European contribution to the project. F4E signed the Procurement Agreement with ITER Organization (IO) in 2009. After a call for tender in 2010, the contract for the manufacturing of seven sectors was placed in October 2010 to a consortium of three Italian companies, Ansaldo, Mangiarotti and Walter Tosto (AMW). The first sector in the manufacturing route is Sector 5 (later will come 4, 3, 2, 9, 8, 7). This paper will cover: the status of the engineering activities, design, procurement and preparation to begin the manufacturing in 2013. Also will be presented the statutory and regulatory requirements of the French Nuclear Safety regulator and the status of the relevant R and D mock-ups to demonstrate manufacturing feasibility control of distortions (using predictions with analysis and algorithms to change in real time the manufacturing route in order to correct such distortions, inspectability and metrology). Another important aspect at this stage of the manufacturing is qualification of activities like welding, Non-destructive Examination and Hot Forming. This paper describes the status of the activities currently in process in order to meet with the challenging design, schedule and high quality requirements of the project.

Tokamak equilibrium reconstruction code LIUQE and its real time implementation

International Nuclear Information System (INIS)

Moret, J.-M.; Duval, B.P.; Le, H.B.; Coda, S.; Felici, F.; Reimerdes, H.

2015-01-01

Highlights: • Algorithm vertical stabilisation using a linear parametrisation of the current density. • Experimentally derived model of the vacuum vessel to account for vessel currents. • Real-time contouring algorithm for flux surface averaged 1.5 D transport equations. • Full real time implementation coded in SIMULINK runs in less than 200 μs. • Applications: shape control, safety factor profile control, coupling with RAPTOR. - Abstract: Equilibrium reconstruction consists in identifying, from experimental measurements, a distribution of the plasma current density that satisfies the pressure balance constraint. The LIUQE code adopts a computationally efficient method to solve this problem, based on an iterative solution of the Poisson equation coupled with a linear parametrisation of the plasma current density. This algorithm is unstable against vertical gross motion of the plasma column for elongated shapes and its application to highly shaped plasmas on TCV requires a particular treatment of this instability. TCV's continuous vacuum vessel has a low resistance designed to enhance passive stabilisation of the vertical position. The eddy currents in the vacuum vessel have a sizeable influence on the equilibrium reconstruction and must be taken into account. A real time version of LIUQE has been implemented on TCV's distributed digital control system with a cycle time shorter than 200 μs for a full spatial grid of 28 by 65, using all 133 experimental measurements and including the flux surface average of quantities necessary for the real time solution of 1.5 D transport equations. This performance was achieved through a thoughtful choice of numerical methods and code optimisation techniques at every step of the algorithm, and was coded in MATLAB and SIMULINK for the off-line and real time version respectively

Diagnostic integration solutions in the ITER first wall

International Nuclear Information System (INIS)

Martínez, Gonzalo; Martin, Alex; Watts, Christopher; Veshchev, Evgeny; Reichle, Roger; Shigin, Pavel; Sabourin, Flavien; Gicquel, Stefan; Mitteau, Raphael; González, Jorge

2015-01-01

Highlights: • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets. • An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW. - Abstract: ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems.

Diagnostic integration solutions in the ITER first wall

Energy Technology Data Exchange (ETDEWEB)

Martínez, Gonzalo, E-mail: gonzalo.martinez@iter.org [Technical University of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Martin, Alex; Watts, Christopher; Veshchev, Evgeny; Reichle, Roger [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Shigin, Pavel [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); National Research Nuclear University (MEPhI), Kashirskoe shosse, 115409 Moscow (Russian Federation); Sabourin, Flavien [ABMI-Groupe, Parc du Relais BatD 201 Route de SEDS, 13127 Vitrolles (France); Gicquel, Stefan; Mitteau, Raphael [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); González, Jorge [RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat (Spain)

2015-10-15

Highlights: • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets. • An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW. - Abstract: ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems.

Structural analysis of ITER multi-purpose deployer

International Nuclear Information System (INIS)

Manuelraj, Manoah Stephen; Dutta, Pramit; Gotewal, Krishan Kumar; Rastogi, Naveen; Tesini, Alessandro; Choi, Chang-Hwan

2016-01-01

Highlights: • System modelling for structural analysis of the Multi-Purpose Deployer (MPD). • Finite element modeling of the Multi-Purpose Deployer (MPD). • Static, modal and seismic response analysis of the Multi-Purpose Deployer (MPD). • Iterative structural analysis and design update to satisfy the structural criteria. • Modal analysis for various kinematic configurations. • Reaction force calculations on the interfacing systems. - Abstract: The Multi-Purpose Deployer (MPD) is a general purpose ITER in-vessel remote handling (RH) system. The main handling equipment, known as the MPD Transporter, consists of a series of linked bodies, which provide anchoring to the vacuum vessel port and an articulated multi-degree of freedom motion to perform various in-vessel maintenance tasks. During the in-vessel operations, the structural integrity of the system should be guaranteed against various operational and seismic loads. This paper presents the structural analysis results of the concept design of the MPD Transporter considering the seismic events. Static structural, modal and frequency response spectrum analyses have been performed to verify the structural integrity of the system, and to provide reaction forces to the interfacing systems such as vacuum vessel and cask. Iterative analyses and design updates are carried out based on the reference design of the system to improve the structural behavior of the system. The frequency responses of the system in various kinematics and payloads are assessed.

A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

Energy Technology Data Exchange (ETDEWEB)

Nakahira, Masataka; Takeda, Nobukazu; Onozuka, Masanori [Japan Atomic Energy Agency (Japan); Kakudate, Satoshi [Mitsubishi Heavy Industries, Ltd. (Japan)

2007-07-01

The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. In the mock-up fabrication, it is planned to conduct the following items: 1. Feasibility of the Japanese proposed VV structure and fabrication methods and the applicability to the ITER are to be confirmed; 2. Assembly procedure and inspection procedure are to be confirmed; 3. Manufacturing tolerances are to be assessed; 4. Manufacturing schedule is to be assessed. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. General scope of the mock-up fabrication and the detailed dimensions are also shown. In the VV fabrication, several types of weld joint configuration will be used. This report shows the joint configurations proposed by Japan to be used for the inner shell connection, the rib-to-shell connection and outer shell connection, and the housing-to-shell connection, respectively. Non-destructive testing considered to be applied to each joint configuration is also presented. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of nondestructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. (orig.)

Status and issues of the European contribution to ITER

International Nuclear Information System (INIS)

Bindslev, H.

2015-01-01

Highlights: • We describe the technical status of F4E's contributions to the ITER International Fusion Energy Project. • The foundations of the ITER Tokamak Complex have been completed. • We describe the production of the Toroidal Field coils and the achieved accuracy. • The first stage of ITER's pre-qualification programme for the ITER first wall panels was completed. • Technical developments for several other ITER components are described. - Abstract: Fusion for Energy (F4E), on behalf of Europe, is responsible for the procurement of most of the high-technology items for the ITER device. This paper provides an overview of the technical status of Europe's contributions to ITER and the related challenges. In particular, we report on progress in the construction of the buildings at the Cadarache site, the fabrication of the superconducting magnets and the vacuum vessel and the testing and qualification of the in-vessel components (first wall and divertor). The status of the design and development of the additional heating systems and the test blanket modules will also be described.

ITER vacuum vessel dynamic stress analysis of a disruption

International Nuclear Information System (INIS)

Riemer, B.W.; Conner, D.L.; Strickler, D.J.; Williamson, D.E.

1994-01-01

Dynamic stress analysis of the International Thermonuclear Experimental Reactor vacuum vessel loaded by disruption forces was performed. The deformation and stress results showed strong inertial effects when compared to static analyses. Maximum stress predicted dynamically was 300 MPa, but stress shown by static analysis from loads at the same point in time reached only 80 MPa. The analysis also provided a reaction load history in the vessel's supports which is essential in evaluating support design. The disruption forces were estimated by assuming a 25-MA plasma current decaying at 1 MA/ms while moving vertically. In addition to forces developed within the vessel, vertical loadings from the first wall/strong back assemblies and the divertor were applied to the vessel at their attachment points. The first 50 natural modes were also determined. The first mode's frequency was 6.0 Hz, and its shape is characterized by vertical displacement of the vessel inner leg. The predicted deformation of the vessel appeared similar to its first mode shape combined with radial contraction. Kinetic energy history from the analysis also correlated with the first mode frequency

Generalized multivalued equilibrium-like problems: auxiliary principle technique and predictor-corrector methods

Directory of Open Access Journals (Sweden)

Vahid Dadashi

2016-02-01

Full Text Available Abstract This paper is dedicated to the introduction a new class of equilibrium problems named generalized multivalued equilibrium-like problems which includes the classes of hemiequilibrium problems, equilibrium-like problems, equilibrium problems, hemivariational inequalities, and variational inequalities as special cases. By utilizing the auxiliary principle technique, some new predictor-corrector iterative algorithms for solving them are suggested and analyzed. The convergence analysis of the proposed iterative methods requires either partially relaxed monotonicity or jointly pseudomonotonicity of the bifunctions involved in generalized multivalued equilibrium-like problem. Results obtained in this paper include several new and known results as special cases.

Rigorous approximation of stationary measures and convergence to equilibrium for iterated function systems

International Nuclear Information System (INIS)

Galatolo, Stefano; Monge, Maurizio; Nisoli, Isaia

2016-01-01

We study the problem of the rigorous computation of the stationary measure and of the rate of convergence to equilibrium of an iterated function system described by a stochastic mixture of two or more dynamical systems that are either all uniformly expanding on the interval, either all contracting. In the expanding case, the associated transfer operators satisfy a Lasota–Yorke inequality, we show how to compute a rigorous approximations of the stationary measure in the L "1 norm and an estimate for the rate of convergence. The rigorous computation requires a computer-aided proof of the contraction of the transfer operators for the maps, and we show that this property propagates to the transfer operators of the IFS. In the contracting case we perform a rigorous approximation of the stationary measure in the Wasserstein–Kantorovich distance and rate of convergence, using the same functional analytic approach. We show that a finite computation can produce a realistic computation of all contraction rates for the whole parameter space. We conclude with a description of the implementation and numerical experiments. (paper)

Feasibility of Batch Reactive Distillation with Equilibrium-Limited Consecutive Reactions in Rectifier, Stripper, or Middle-Vessel Column

Directory of Open Access Journals (Sweden)

T. Lukács

2011-01-01

Full Text Available A general overall feasibility methodology of batch reactive distillation of multireaction systems is developed to study all the possible configurations of batch reactive distillation. The general model equations are derived for multireaction system with any number of chemical equilibrium-limited reactions and for any number of components. The present methodology is demonstrated with the detailed study of the transesterification of dimethyl carbonate in two reversible cascade reactions in batch reactive distillation process. Pure methanol is produced as distillate, and pure diethyl carbonate is produced at the bottom simultaneously in middle-vessel column; in each section, continuous feeding of ethanol is necessary. The results of feasibility study are successfully validated by rigorous simulations.

The optimal monochromatic spectral computed tomographic imaging plus adaptive statistical iterative reconstruction algorithm can improve the superior mesenteric vessel image quality

Energy Technology Data Exchange (ETDEWEB)

Yin, Xiao-Ping; Zuo, Zi-Wei; Xu, Ying-Jin; Wang, Jia-Ning [CT/MRI room, Affiliated Hospital of Hebei University, Baoding, Hebei, 071000 (China); Liu, Huai-Jun, E-mail: hebeiliu@outlook.com [Department of Medical Imaging, The Second Hospital of Hebei Medical University, Shijiazhuang, Hebei, 050000 (China); Liang, Guang-Lu [CT/MRI room, Affiliated Hospital of Hebei University, Baoding, Hebei, 071000 (China); Gao, Bu-Lang, E-mail: browngao@163.com [Department of Medical Research, Shijiazhuang First Hospital, Shijiazhuang, Hebei, 050011 (China)

2017-04-15

Objective: To investigate the effect of the optimal monochromatic spectral computed tomography (CT) plus adaptive statistical iterative reconstruction on the improvement of the image quality of the superior mesenteric artery and vein. Materials and methods: The gemstone spectral CT angiographic data of 25 patients were reconstructed in the following three groups: 70 KeV, the optimal monochromatic imaging, and the optimal monochromatic plus 40%iterative reconstruction mode. The CT value, image noises (IN), background CT value and noises, contrast-to-noise ratio (CNR), signal-to-noise ratio (SNR) and image scores of the vessels and surrounding tissues were analyzed. Results: In the 70 KeV, the optimal monochromatic and the optimal monochromatic images plus 40% iterative reconstruction group, the mean scores of image quality were 3.86, 4.24 and 4.25 for the superior mesenteric artery and 3.46, 3.78 and 3.81 for the superior mesenteric vein, respectively. The image quality scores for the optimal monochromatic and the optimal monochromatic plus 40% iterative reconstruction groups were significantly greater than for the 70 KeV group (P < 0.05). The vascular CT value, image noise, background noise, CNR and SNR were significantly (P < 0.001) greater in the optimal monochromatic and the optimal monochromatic images plus 40% iterative reconstruction group than in the 70 KeV group. The optimal monochromatic plus 40% iterative reconstruction group had significantly (P < 0.05) lower image and background noise but higher CNR and SNR than the other two groups. Conclusion: The optimal monochromatic imaging combined with 40% iterative reconstruction using low-contrast agent dosage and low injection rate can significantly improve the image quality of the superior mesenteric artery and vein.

The ITER Thomson scattering core LIDAR diagnostic

NARCIS (Netherlands)

Naylor, G.A.; Scannell, R.; Beurskens, M.; Walsh, M.J.; Pastor, I.; Donné, A.J.H.; Snijders, B.; Biel, W.; Meszaros, B.; Giudicotti, L.; Pasqualotto, R.; Marot, L.

2012-01-01

The central electron temperature and density of the ITER plasma may be determined by Thomson scattering. A LIDAR topology is proposed in order to minimize the port access required of the ITER vacuum vessel. By using a LIDAR technique, a profile of the electron temperature and density can be

Overview and status of ITER Cryostat manufacturing

Energy Technology Data Exchange (ETDEWEB)

Bhardwaj, Anil K., E-mail: anil.bhardwaj@iter-india.org [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Gupta, Girish; Prajapati, Rajnikant; Joshi, Vaibhav; Patel, Mitul; Bhavsar, Jagrut; More, Vipul; Jindal, Mukesh; Bhattacharya, Avik; Jogi, Gourav; Palaliya, Amit; Jha, Saroj; Pandey, Manish; Shukla, Dileep [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Iyer, Ganesh; Jadhav, Pandurang; Goyal, Dipesh; Desai, Anish [Larsen & Toubro Limited, Heavy Engineering, Hazira Manufacturing Complex, Gujarat (India); Sekachev, I.; Vitupier, Guillaume [ITER Organization, Route de Vinon sur Verdon – CS 90046, 13067 Saint Paul Lez Durance Cedex (France); and others

2016-11-01

Highlights: • Manufacturing status of one of the largest and the heaviest fully welded stainless steel vacuum chambers in the world (ITER Cryostat). • Overview of manufacturing stages and its segmentation. • Overview of manufacturing procedures and assembly and installation. - Abstract: One of ITER-India's commitments to the ITER Organization is procurement of the ITER Cryostat. It is a large vacuum vessel (∼29 m dia. and ∼29 m height), which is made up of 304/304 L dual marked stainless steel and has a total mass over 3500 t. The thickness of the vessel wall varies from 50 mm to 190 mm. It is one of the largest and the heaviest fully welded stainless steel vacuum chambers in the world which provides vacuum thermal insulation for the superconducting magnets operating at 4.5 K and for the thermal shield operating at 80 K. It also mechanically supports the magnet system along with the vacuum vessel (VV). The cryostat is designed and constructed according to ASME Section-VIII Division-2 with additional ITER Vacuum Handbook requirements and it is classified as protection important component (PIC-2). Manufacturing of cryostat segments is ongoing in India; sub-assembly of four major sections of the cryostat from the segments will be done at the ITER site in a temporary workshop building and the final assembly will be done in the pit of the tokamak building, the final location. The cryostat manufacturing contract has been awarded to Larsen and Toubro Limited in August 2012 after completion of design [4] and signing of Procurement Arrangement [1] with ITER Organization. Manufacturing of the cryostat was started in January 2014 after approval of the manufacturing drawings and procedures. The temporary workshop of 44 m × 110 m × 26 m in height has been completed in November 2014 at the ITER site with a 200 t crane installed. This paper gives an overview and the status of the cryostat manufacturing.

Discharge of Non-Reactive Fluids from Vessels

Directory of Open Access Journals (Sweden)

M. Castier

Full Text Available Abstract This paper presents simulations of discharges from pressure vessels that consistently account for non-ideal fluid behavior in all the required thermodynamic properties and individually considers all the chemical components present. The underlying assumption is that phase equilibrium occurs instantaneously inside the vessel and, thus, the dynamics of the fluid in the vessel comprises a sequence of equilibrium states. The formulation leads to a system of differential-algebraic equations in which the component mass balances and the energy balance are ordinary differential equations. The algebraic equations account for the phase equilibrium conditions inside the vessel and at the discharge point, and for sound speed calculations. The simulator allows detailed predictions of the condition inside the vessel and at the discharge point as a function of time, including the flow rate and composition of the discharge. The paper presents conceptual applications of the simulator to predict the effect of leaks from vessels containing mixtures of light gases and/or hydrocarbons and comparisons to experimental data.

Hybrid Iterative Scheme for Triple Hierarchical Variational Inequalities with Mixed Equilibrium, Variational Inclusion, and Minimization Constraints

Directory of Open Access Journals (Sweden)

Lu-Chuan Ceng

2014-01-01

Full Text Available We introduce and analyze a hybrid iterative algorithm by combining Korpelevich's extragradient method, the hybrid steepest-descent method, and the averaged mapping approach to the gradient-projection algorithm. It is proven that, under appropriate assumptions, the proposed algorithm converges strongly to a common element of the fixed point set of finitely many nonexpansive mappings, the solution set of a generalized mixed equilibrium problem (GMEP, the solution set of finitely many variational inclusions, and the solution set of a convex minimization problem (CMP, which is also a unique solution of a triple hierarchical variational inequality (THVI in a real Hilbert space. In addition, we also consider the application of the proposed algorithm to solving a hierarchical variational inequality problem with constraints of the GMEP, the CMP, and finitely many variational inclusions.

ITER EDA newsletter. V. 3, no. 2

International Nuclear Information System (INIS)

1994-02-01

This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the Fifth ITER Council Meeting held in Garching, Germany, 27-28 January 1994, a visit (28 January 1994) of an international group of Harvard Fellows to the San Diego Joint Work Site, the Inauguration Ceremony of the EC-hosted ITER joint work site in Garching (28 January 1994), on an ITER Technical Meeting on Assembly and Maintenance held in Garching, Germany, January 19-26, 1994, and a report on a Technical Committee Meeting on radiation effects on in-vessel components held in Garching, Germany, November 15-19, 1993, as well as an ITER Status Report

Validation of the inspections with ultrasound of the welds of the reactor of ITER vacuum vessel; Validacion de las inspecciones con ultrasonidos de las soldaduras de la Vasija de Vacio del reactor del ITER

Energy Technology Data Exchange (ETDEWEB)

Garcia, A.; Fernandez, F.; Perez, C.; Sillero, J. A.

2013-07-01

The ITER fusion reactor vacuum vessel has thousands of welding austenitic with shapes and different manufacturing processes. The RCC-MR code, which is that applied to the manufacture of the fusion reactor, requires a volumetric test all of them. This test should be mainly by x-rays and welds where it was not possible to use this method, ultrasonic.09-06.

The design of the ITER first wall panels

Energy Technology Data Exchange (ETDEWEB)

Mitteau, R., E-mail: raphael.mitteau@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Calcagno, B.; Chappuis, P.; Eaton, R.; Gicquel, S. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Chen, J. [Southwestern Institute of Physics, Huangjing Road, Chengdu 610225 (China); Labusov, A. [Efremov Research Institute, 189631 St. Petersburg (Russian Federation); Martin, A.; Merola, M.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ulrickson, M. [Sandia National Laboratory, Albuquerque, NM (United States); Zacchia, F. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

2013-10-15

Highlights: • The ITER blanket is in the final stage of design completion. • Issues raised about the blanket heat loads and remote handling strategy are addressed, while integrating the in-vessel coils. • Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R and D activities. -- Abstract: The ITER blanket is in the final stage of design completion. The issues raised during the 2007 ITER design review about the first wall (FW) heat loads and remote handling strategy have been addressed, while integrating the recently confirmed in-vessel coils. This paper focuses on the FW design, which is nearing completion. Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R and D activities.

Optimization and Control for Sharing of the ITER Vacuum Vessel Support Force

International Nuclear Information System (INIS)

Rozov, V.

2006-01-01

The ITER Vacuum Vessel (VV) is a complex body supported in 9 points below lower ports by restraints in the radial, toroidal and vertical directions. The applied load produces a combination of reaction forces, which must be consistent with the design of the supported object. A reasonable sharing of the load among the supports is important for overall performance of the structure and helps to avoid excessive stress at the joints between the VV and lower ports. Optimization has been performed of the sharing of the total horizontal load applied to the ITER VV between radial and toroidal restraints. An effective method of finding simple parametric relationships between the design parameters of supports and the balance of the reaction forces has been developed. This allows purely analytical prediction of the sharing of the reaction forces for any desired stiffness of the applied restraints with no need for finite element structural analysis, and also allows control of the sharing by a proper selection of parameters of the supports. The method is based on the use of elementary mono-directional schemes - equivalent oscillators built for the main global modes, in static problems. The types of schemes and parameters of their members, related to the a-priori unknown stiffness of the VV structure under the supports, are found from consideration of the free vibration problem for the object using a 3D model of the VV with mass simulators - a series of simple eigenvalue analyses with variation of stiffness of the external restraints, that demands quite moderate computational resources. The equivalent schemes for the main modes not only enable simple one-line analytical calculation of the natural frequencies at any desired stiffness of the supports, but also indicate the contributions and balance of stiffness, to be considered in the static problem. The results of assessments of the reaction forces by direct static structural analyses for several cases are in agreement with values

Manufacture of EAST VS In-Vessel Coil

International Nuclear Information System (INIS)

Long, Feng; Wu, Yu; Du, Shijun; Jin, Huan; Yu, Min; Han, Qiyang; Wan, Jiansheng; Liu, Bin; Qiao, Jingchun; Liu, Xiaochuan; Li, Chang; Cai, Denggang; Tong, Yunhua

2013-01-01

Highlights: • ITER like Stainless Steel Mineral Insulation Conductor (SSMIC) used for EAST Tokamak VS In-Vessel Coil manufacture first time. • Research on SSMIC fabrication was introduced in detail. • Two sets totally four single-turn VS coils were manufactured and installed in place symmetrically above and below the mid-plane in the vacuum vessel of EAST. • The manufacture and inspection of the EAST VS coil especially the joint for the SSMIC connection was described in detail. • The insulation resistances of all the VS coils have no significant reduction after endurance test. -- Abstract: In the ongoing latest update round of EAST (Experimental Advanced Superconducting Tokamak), two sets of two single-turn Vertical Stabilization (VS) coils were manufactured and installed symmetrically above and below the mid-plane in the vacuum vessel of EAST. The Stainless Steel Mineral Insulated Conductor (SSMIC) developed for ITER In-Vessel Coils (IVCs) in Institute of Plasma Physics, Chinese Academy of Science (ASIPP) was used for the EAST VS coils manufacture. Each turn poloidal field VS coil includes three internal joints in the vacuum vessel. The middle joint connects two pieces of conductor which together form an R2.3 m arc segment inside the vacuum vessel. The other two joints connect the arc segment with the two feeders near the port along the toroidal direction to bear lower electromagnetic loads during operation. Main processes and tests include material performances checking, conductor fabrication, joint connection and testing, coil forming, insulation performances measurement were described herein

Weld distortion prediction of the ITER Vacuum Vessel using Finite Element simulations

Energy Technology Data Exchange (ETDEWEB)

Caixas, Joan, E-mail: joan.caixas@f4e.europa.eu [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Guirao, Julio [Numerical Analysis Technologies, S. L., Marqués de San Esteban 52, Entlo, 33209 Gijon (Spain); Bayon, Angel; Jones, Lawrence; Arbogast, Jean François [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Barbensi, Andrea [Ansaldo Nucleare, Corso F.M. Perrone, 25, I-16152 Genoa (Italy); Dans, Andres [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Facca, Aldo [Mangiarotti, Pannellia di Sedegliano, I-33039 Sedegliano (UD) (Italy); Fernandez, Elena; Fernández, José [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Iglesias, Silvia [Numerical Analysis Technologies, S. L., Marqués de San Esteban 52, Entlo, 33209 Gijon (Spain); Jimenez, Marc; Jucker, Philippe; Micó, Gonzalo [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Ordieres, Javier [Numerical Analysis Technologies, S. L., Marqués de San Esteban 52, Entlo, 33209 Gijon (Spain); Pacheco, Jose Miguel [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Paoletti, Roberto [Walter Tosto, Via Erasmo Piaggio, 72, I-66100 Chieti Scalo (Italy); Sanguinetti, Gian Paolo [Ansaldo Nucleare, Corso F.M. Perrone, 25, I-16152 Genoa (Italy); Stamos, Vassilis [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Tacconelli, Massimiliano [Walter Tosto, Via Erasmo Piaggio, 72, I-66100 Chieti Scalo (Italy)

2013-10-15

Highlights: ► Computational simulations of the weld processes can rapidly assess different sequences. ► Prediction of welding distortion to optimize the manufacturing sequence. ► Accurate shape prediction after each manufacture phase allows to generate modified procedures and pre-compensate distortions. ► The simulation methodology is improved using condensed computation techniques with ANSYS in order to reduce computation resources. ► For each welding process, the models are calibrated with the results of coupons and mock-ups. -- Abstract: The as-welded surfaces of the ITER Vacuum Vessel sectors need to be within a very tight tolerance, without a full-scale prototype. In order to predict welding distortion and optimize the manufacturing sequence, the industrial contract includes extensive computational simulations of the weld processes which can rapidly assess different sequences. The accurate shape prediction, after each manufacturing phase, enables actual distortions to be compared with the welding simulations to generate modified procedures and pre-compensate distortions. While previous mock-ups used heavy welded-on jigs to try to restrain the distortions, this method allows the use of lightweight jigs and yields important cost and rework savings. In order to enable the optimization of different alternative welding sequences, the simulation methodology is improved using condensed computation techniques with ANSYS in order to reduce computational resources. For each welding process, the models are calibrated with the results of coupons and mock-ups. The calibration is used to construct representative models of each segment and sector. This paper describes the application to the construction of the Vacuum Vessel sector of the enhanced simulation methodology with condensed Finite Element computation techniques and results of the calibration on several test pieces for different types of welds.

Weld distortion prediction of the ITER Vacuum Vessel using Finite Element simulations

International Nuclear Information System (INIS)

Caixas, Joan; Guirao, Julio; Bayon, Angel; Jones, Lawrence; Arbogast, Jean François; Barbensi, Andrea; Dans, Andres; Facca, Aldo; Fernandez, Elena; Fernández, José; Iglesias, Silvia; Jimenez, Marc; Jucker, Philippe; Micó, Gonzalo; Ordieres, Javier; Pacheco, Jose Miguel; Paoletti, Roberto; Sanguinetti, Gian Paolo; Stamos, Vassilis; Tacconelli, Massimiliano

2013-01-01

Highlights: ► Computational simulations of the weld processes can rapidly assess different sequences. ► Prediction of welding distortion to optimize the manufacturing sequence. ► Accurate shape prediction after each manufacture phase allows to generate modified procedures and pre-compensate distortions. ► The simulation methodology is improved using condensed computation techniques with ANSYS in order to reduce computation resources. ► For each welding process, the models are calibrated with the results of coupons and mock-ups. -- Abstract: The as-welded surfaces of the ITER Vacuum Vessel sectors need to be within a very tight tolerance, without a full-scale prototype. In order to predict welding distortion and optimize the manufacturing sequence, the industrial contract includes extensive computational simulations of the weld processes which can rapidly assess different sequences. The accurate shape prediction, after each manufacturing phase, enables actual distortions to be compared with the welding simulations to generate modified procedures and pre-compensate distortions. While previous mock-ups used heavy welded-on jigs to try to restrain the distortions, this method allows the use of lightweight jigs and yields important cost and rework savings. In order to enable the optimization of different alternative welding sequences, the simulation methodology is improved using condensed computation techniques with ANSYS in order to reduce computational resources. For each welding process, the models are calibrated with the results of coupons and mock-ups. The calibration is used to construct representative models of each segment and sector. This paper describes the application to the construction of the Vacuum Vessel sector of the enhanced simulation methodology with condensed Finite Element computation techniques and results of the calibration on several test pieces for different types of welds

F4E R and D programme and results on in-vessel dust and tritium

International Nuclear Information System (INIS)

Le Guern, F.; Gulden, W.; Ciattaglia, S.; Counsell, G.; Bengaouer, A.; Brinster, J.; Dabbene, F.; Denkevitz, A.; Jordan, T.; Kuznetsov, M.; Porfiri, M.T.; Redlinger, R.; Roblin, Ph.; Roth, J.; Segre, J.; Sugiyama, K.; Tkatschenko, I.; Xu, Z.

2011-01-01

In a Tokamak vacuum vessel, plasma-wall interactions can result in the production of radioactive dust and H isotopes (including tritium) can be trapped both in in-vessel material and in dust. The vacuum vessel represents the most important confinement barrier to this radioactive material. In the event of an accident involving ingress of steam to the vacuum vessel, hydrogen could be produced by chemical reactions with hot metal and dust. Hydrogen isotopes could also be desorbed from in-vessel components, e.g. cryopumps. In events where an ingress of air to the vacuum vessel occurs, reaction of the air with hydrogen and/or dust therefore cannot be completely excluded. Due to the radiological risks highlighted by the safety evaluation studies for ITER in normal conditions (e.g. in-vessel maintenance chronic release) and accidental ones (e.g. challenge of vacuum vessel tightness in the event of a hydrogen/dust explosion with air), limitations on the accumulation of dust and tritium in the vacuum vessel are imposed as well as controls over the maximum extent of the quantity of accidental air ingress. ITER IO has defined a strategy for the control of in-vessel dust and tritium inventories below the safety limits based primarily on the measurement and removal of dust and tritium. In this context, this paper will report on the efforts under F4E responsibility to develop a number of the new ITER baseline systems. In particular this paper, after a review of safety constraints and ITER strategy, provides the status of: (1) tasks being launched on diagnostics for in-vessel dust inventory measurement, (2) experiments to enrich the data about the effectiveness of desorption of tritium from Be at 350 o C (divertor baking aiming to release significant amount of tritium trapped in Be co-deposit), (3) on-going R and D programme (experimental and numerical simulation) at FZK, CEA and ENEA on in-vacuum vessel H2 dust explosion.

Equilibrium vertical field in the TBR Tokamak

International Nuclear Information System (INIS)

Ueta, A.Y.

1985-01-01

An experimental study on the influence of the vertical magnetic field of the TBR tokamak on the stability and equilibrium of plasma column, was done. Magnetic pick-up coils were built to measure plasma current and position, together with active networks, necessary fo the electronic processing of signals. Some measurements were on the space configuration of the vertical field, and on the influence due to the toroidal vessel. From the data obtained it was possible to discuss the influence of the currents induced on the vessel surface, on plasma equilibrium. Theoretical and experimental results of the vertica field, as a function of plasma current were compared, and allowed an evaluation of the plasma kinetic pressure and temperature. (Author) [pt

Remote operational trials with the ITER FDR divertor handling equipment

International Nuclear Information System (INIS)

Irving, M.; Baldi, L.; Benamati, G.; Galbiati, L.; Giacomelli, S.; Lorenzelli, L.; Micciche, G.; Muro, L.; Polverari, A.; Palmer, J.; Martin, E.

2003-01-01

The ITER divertor test platform (DTP) located at ENEA's Research Centre in Brasimone, Italy is a full-scale mock-up of a 72 deg. arc of the ITER 1998 vessel divertor region--the result of a major initiative over the period 1996-2000. Since the implementation of this facility, the design of the ITER vessel--and therefore much of the remote maintenance equipment--has changed substantially. However, the nature and principles of the remote handling equipment are still very similar, and hence many valuable lessons can yet be learned from the existing equipment for the future. In particular, true remote handling tests of the major maintenance subsystems were seen as an important step in determining their suitability for ITER. This paper describes and documents a series of three, discrete, remote-handling trials carried out using most of the major DTP subsystems, and presents an overview of the conclusions and suggestions for future development of ITER cassette remote handling equipment

The radiation analyses of ITER lower ports

International Nuclear Information System (INIS)

Petrizzi, L.; Brolatti, G.; Martin, A.; Loughlin, M.; Moro, F.; Villari, R.

2010-01-01

The ITER Vacuum Vessel has upper, equatorial, and lower ports used for equipment installation, diagnostics, heating and current drive systems, cryo-vacuum pumping, and access inside the vessel for maintenance. At the level of the divertor, the nine lower ports for remote handling, cryo-vacuum pumping and diagnostic are inclined downwards and toroidally located each every 40 o . The cryopump port has additionally a branch to allocate a second cryopump. The ports, as openings in the Vacuum Vessel, permit radiation streaming out of the vessel which affects the heating in the components in the outer regions of the machine inside and outside the ports. Safety concerns are also raised with respect to the dose after shutdown at the cryostat behind the ports: in such zones the radiation dose level must be kept below the regulatory limit to allow personnel access for maintenance purposes. Neutronic analyses have been required to qualify the ITER project related to the lower ports. A 3-D model was used to take into account full details of the ports and the lower machine surroundings. MCNP version 5 1.40 has been used with the FENDL 2.1 nuclear data library. The ITER 40 o model distributed by the ITER Organization was developed in the lower part to include the relevant details. The results of a first analysis, focused on cryopump system only, were recently published. In this paper more complete data on the cryopump port and analysis for the remote handling port and the diagnostic rack are presented; the results of both analyses give a complete map of the radiation loads in the outer divertor ports. Nuclear heating, dpa, tritium production, and dose rates after shutdown are provided and the implications for the design are discussed.

Is Carbon a Realistic Choice for ITER's Divertor?

International Nuclear Information System (INIS)

Skinner, C.H.; Federici, G.

2005-01-01

Tritium retention by co-deposition with carbon on the divertor target plate is predicted to limit ITER's DT burning plasma operations (e.g. to about 100 pulses for the worst conditions) before the in-vessel tritium inventory limit, currently set at 350 g, is reached. At this point, ITER will only be able to continue its burning plasma program if technology is available that is capable of rapidly removing large quantities of tritium from the vessel with over 90% efficiency. The removal rate required is four orders of magnitude faster than that demonstrated in current tokamaks. Eighteen years after the observation of co-deposition on JET and TFTR, such technology is nowhere in sight. The inexorable conclusion is that either a major initiative in tritium removal should be funded or that research priorities for ITER should focus on metal alternatives

Fabrication of full-size mock-up for 10° section of ITER vacuum vessel thermal shield

Energy Technology Data Exchange (ETDEWEB)

Kang, Dong Kwon [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo, E-mail: kwnam@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kang, Kyoung-O; Noh, Chang Hyun; Chung, Wooho [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Lim, Kisuk; Kang, Youngkil [SFA Engineering Corp., Asan-si, Chungcheongnam-do 336-873 (Korea, Republic of); Hamlyn-Harris, Craig; Her, Namil; Robby, Hicks [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2015-10-15

In this paper, a full-scale prototype fabrication for vacuum vessel thermal shield (VVTS) of ITER tokamak is described and test results are reported. All the manufacturing processes except for silver coating were performed in the fabrication of 10° section of VVTS. Pre-qualification test was conducted to compare the vertical and the horizontal welding positions. After shell welding, shell distortion was measured and adjusted. Shell thickness change was also measured after buffing process. Specially, VVTS ports need large bending and complex welding of shell and flange. Bending method for the complex and long cooling tube layout especially for the VVTS ports was developed in detail. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner and the scanning data was analyzed.

Status of Preliminary Design on the Assembly Tools for ITER Tokamak Machine

International Nuclear Information System (INIS)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin; Moon, Jae Hwan; Kim, Byung Seok; Lee, Jae Hyuk; Shaw, Robert

2012-01-01

The ITER Tokamak device is principally composed of nine 40 .deg. sectors. Each 40 .deg. sector is made up of one 40 .deg. vacuum vessel (VV), two 20 .deg. toroidal filed coils (TFC) and associated vacuum vessel thermal shield (VVTS) segments which consist of one inboard and two outboard vacuum vessel thermal shields. Based on the design description document and final report prepared by the ITER organization (IO) and conceptual design, Korea has carried out the preliminary design of these assembly tools. The assembly strategy and relevant tools for the 40 .deg. sector sub-assembly and sector assembly at in-pit should be developed to satisfy the basic assembly requirements of the ITER Tokamak machine. Assembly strategy, preliminary design of the sector sub-assembly and assembly tools are described in this paper

Design and structural analysis of support structure for ITER vacuum vessel

International Nuclear Information System (INIS)

Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka; Shibanuma, Kiyoshi

2004-01-01

The International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed as a new concept, which is deferent from the current design, i.e., the VV support is directly connected to the toroidal coils (TF coils). This independent concept has two advantages comparing to the current one: (1) thermal load due to the temperature deference between VV and TF coils becomes lower and (2) the TF coils are categorized as non-safety components because of its independence from VV. Stress Analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coils is found to be 15 mm, much less than the current design clearance of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support. (author)

ITER assembly and maintenance

International Nuclear Information System (INIS)

Honda, T.; Davis, F.; Lousteau, D.

1991-01-01

This document is intended to describe the work conducted by the ITER Assembly and Maintenance (A and M) Design Unit and the supporting home teams during the ITER Conceptual Design Activities, carried out from 1988 through 1990. Its content consists of two main sections, i.e., Chapter III, which describes the identified tasks to be performed by the A and M system and a general description of the required equipment; and Chapter IV, which provides a more detailed description of the equipment proposed to perform the assigned tasks. A two-stage R and D program is now planned, i.e., (1) a prototype equipment functional tests using full scale mock-ups and (2) a full scale integration demonstration test facility with real components (vacuum vessel with ports, blanket modules, divertor modules, armor tiles, etc.). Crucial in-vessel and ex-vessel operations and the associated remote handling equipment, including handling of divertor plates and blanket modules will be demonstrated in the first phase, whereby the database needed to proceed with the engineering phase will be acquired. The second phase will demonstrate the ability of the overall system to execute the required maintenance procedures and evaluate the performance of the prototype equipment

ITER primary cryopump test facility

International Nuclear Information System (INIS)

Petersohn, N.; Mack, A.; Boissin, J.C.; Murdoc, D.

1998-01-01

A cryopump as ITER primary vacuum pump is being developed at FZK under the European fusion technology programme. The ITER vacuum system comprises of 16 cryopumps operating in a cyclic mode which fulfills the vacuum requirements in all ITER operation modes. Prior to the construction of a prototype cryopump, the concept is tested on a reduced scale model pump. To test the model pump, the TIMO facility is being built at FZK in which the model pump operation under ITER environmental conditions, except for tritium exposure, neutron irradiation and magnetic fields, can be simulated. The TIMO facility mainly consists of a test vessel for ITER divertor duct simulation, a 600 W refrigerator system supplying helium in the 5 K stage and a 30 kW helium supply system for the 80 K stage. The model pump test programme will be performed with regard to the pumping performance and cryogenic operation of the pump. The results of the model pump testing will lead to the design of the full scale ITER cryopump. (orig.)

Nuclear aspects of diagnostics in RTO/RC ITER

International Nuclear Information System (INIS)

Walker, C.I.; Yamamoto, S.; Costley, A.; Kock, L. de; Ebisawa, K.; Janeschitz, G.; Khripunov, V.; Martin, E.; Vayakis, G.

2000-01-01

ITER (international thermonuclear experimental reactor) will be the first fusion device where the design of the plasma diagnostic systems will make extensive use of the materials and techniques developed in the nuclear technology field. The designs have to satisfy stringent requirements for tritium confinement, nuclear shielding and vacuum integrity. This paper introduces the requirements for diagnostics in the ITER long pulse, burning plasma environment, and addresses the impact of the reactor environment on the diagnostics and ancillary equipment. These systems necessarily require access to the plasma or first wall, which generally conflicts with the requirements of the basic machine. Holes are required through the first wall, primary shielding, containment boundaries and biological shielding. Components have a limited life and require maintenance. This paper describes the effect of the radiation environment on diagnostic design at different locations. Ex-vessel and in-vessel remote handling, hot cell refurbishment and tritium confinement are also described

Rational Verification in Iterated Electric Boolean Games

Directory of Open Access Journals (Sweden)

Youssouf Oualhadj

2016-07-01

Full Text Available Electric boolean games are compact representations of games where the players have qualitative objectives described by LTL formulae and have limited resources. We study the complexity of several decision problems related to the analysis of rationality in electric boolean games with LTL objectives. In particular, we report that the problem of deciding whether a profile is a Nash equilibrium in an iterated electric boolean game is no harder than in iterated boolean games without resource bounds. We show that it is a PSPACE-complete problem. As a corollary, we obtain that both rational elimination and rational construction of Nash equilibria by a supervising authority are PSPACE-complete problems.

Articulated inspection arm for ITER, a demonstration in the Tore Supra tokamak

Energy Technology Data Exchange (ETDEWEB)

Cordier, J.J.; Gargiulo, L.; Grisolia, C.; Samaille, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Friconneau, J.P.; Perrot, Y. [CEA Fontenay-aux-Roses, LIST Robotics and Interactive Systems Unit, 92 (France); Palmer, J.D. [Max-Planck-Institut fuer Plasmaphysik Boltzmannstr.2, Garching (Germany)

2003-07-01

The aim of this program is to demonstrate for ITER the feasibility of an in-vessel remote handling inspection using a long reach, limited payload carrier (1 to 10 kg) for penetration of the ITER chamber through the openings. This device is dedicated to close inspection of the Plasma Facing Components (PFC). An articulated demonstrator called articulated inspection arm (AIA) has been manufactured. A feasibility study of a full AIA operation in Tore Supra was performed, taking into account ITER reference requirements. A scale one demonstration of the AIA under ITER relevant condition is feasible on Tore Supra and would give significant improvement in research results for ITER remote Handling equipment. The test of the AIA demonstrator behaviour is foreseen in 2005 in real Tokamak conditions. The paper presents the full robot concept, the results of the first test campaign, the AIA new design and its integration on Tore Supra. Several potential uses of the AIA for the in vessel components inspection are being studied such as PFC visual inspection, water loop leak testing, laser ablation for wall detritiation and carbon dust and flakes removal are foreseen as utilities to be placed at the AIA head. These various systems are described in the paper.

Articulated inspection arm for ITER, a demonstration in the Tore Supra tokamak

International Nuclear Information System (INIS)

Cordier, J.J.; Gargiulo, L.; Grisolia, C.; Samaille, F.; Palmer, J.D.

2003-01-01

The aim of this program is to demonstrate for ITER the feasibility of an in-vessel remote handling inspection using a long reach, limited payload carrier (1 to 10 kg) for penetration of the ITER chamber through the openings. This device is dedicated to close inspection of the Plasma Facing Components (PFC). An articulated demonstrator called articulated inspection arm (AIA) has been manufactured. A feasibility study of a full AIA operation in Tore Supra was performed, taking into account ITER reference requirements. A scale one demonstration of the AIA under ITER relevant condition is feasible on Tore Supra and would give significant improvement in research results for ITER remote Handling equipment. The test of the AIA demonstrator behaviour is foreseen in 2005 in real Tokamak conditions. The paper presents the full robot concept, the results of the first test campaign, the AIA new design and its integration on Tore Supra. Several potential uses of the AIA for the in vessel components inspection are being studied such as PFC visual inspection, water loop leak testing, laser ablation for wall detritiation and carbon dust and flakes removal are foreseen as utilities to be placed at the AIA head. These various systems are described in the paper

46 CFR 42.20-12 - Conditions of equilibrium.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Conditions of equilibrium. 42.20-12 Section 42.20-12... BY SEA Freeboards § 42.20-12 Conditions of equilibrium. The following conditions of equilibrium are... stability. Through an angle of 20 degrees beyond its position of equilibrium, the vessel must meet the...

Modeling of ELM Dynamics in ITER

International Nuclear Information System (INIS)

Pankin, A.Y.; Bateman, G.; Kritz, A.H.; Brennan, D.P.; Snyder, P.B.; Kruger, S.

2007-01-01

Edge localized modes (ELMs) are large scale instabilities that alter the H-mode pedestal, reduce the total plasma stored energy, and can result in heat pulses to the divertor plates. These modes can be triggered by pressure driven ballooning modes or by current driven peeling instabilities. In this study, stability analyses are carried out for a series of ITER equilibria that are generated with the TEQ and TOQ equilibrium codes. The H-mode pedestal pressure and parallel component of plasma current density are varied in a systematic way in order to include the relevant parameter space for a specific ITER discharge. Ideal MHD stability codes, DCON, ELITE, and BALOO code, are employed to determine whether or not each ITER equilibrium profile is unstable to peeling or ballooning modes in the pedestal region. Several equilibria that are close to the marginal stability boundary for peeling and ballooning modes are tested with the NIMROD non-ideal MHD code. The effects of finite resistivity are studied in a series of linear NIMROD computations. It is found that the peeling-ballooning stability threshold is very sensitive to the resistivity and viscosity profiles, which vary dramatically over a wide range near the separatrix. Due to the effects of finite resistivity and viscosity, the peeling-ballooning stability threshold is shifted compared to the ideal threshold. A fundamental question in the integrated modeling of ELMy H-mode discharges concerning how much plasma and current density is removed during each ELM crash can be addressed with nonlinear non-ideal MHD simulations. In this study, the NIMROD computer simulations are continued into the nonlinear stage for several ITER equilibria that are marginally unstable to peeling or ballooning modes. The role of two-fluid and finite Larmor radius effects on the ELM dynamics in ITER geometry is examined. The formation of ELM filament structures, which are observed in many existing tokamak experiments, is demonstrated for ITER

Improvements for real-time magnetic equilibrium reconstruction on ASDEX Upgrade

International Nuclear Information System (INIS)

Giannone, L.; Fischer, R.; McCarthy, P.J.; Odstrcil, T.; Zammuto, I.; Bock, A.; Conway, G.; Fuchs, J.C.; Gude, A.; Igochine, V.; Kallenbach, A.; Lackner, K.; Maraschek, M.; Rapson, C.; Ruan, Q.; Schuhbeck, K.H.; Suttrop, W.; Wenzel, L.

2015-01-01

Highlights: • Spline basis current functions with second-order linear regularisation. • Perturbations of magnetic probe measurements due to ferromagnetic tiles on the inner wall and from oscillations in the fast position coil current are corrected. • A constraint of the safety factor on the magnetic axis is introduced. Soft X-ray tomography is used to assess the quality of the real-time magnetic equilibrium reconstruction. • External loop voltage measurements and magnetic probe pairs inside and outside the vessel wall were used to measure the vacuum vessel wall resistivity. - Abstract: Real-time magnetic equilibria are needed for NTM stabilization and disruption avoidance experiments on ASDEX Upgrade. Five improvements to real-time magnetic equilibrium reconstruction on ASDEX Upgrade have been investigated. The aim is to include as many features of the offline magnetic equilibrium reconstruction code in the real-time equilibrium reconstruction code. Firstly, spline current density basis functions with regularization are used in the offline equilibrium reconstruction code, CLISTE [1]. It is now possible to have the same number of spline basis functions in the real-time code. Secondly, in the presence of edge localized modes, (ELM's), it is found to be necessary to include the low pass filter effect of the vacuum vessel on the fast position control coil currents to correctly compensate the magnetic probes for current oscillations in these coils. Thirdly, the introduction of ferromagnetic tiles in ASDEX Upgrade means that a real-time algorithm for including the perturbations of the magnetic equilibrium generated by these tiles is required. A methodology based on tile surface currents is described. Fourthly, during current ramps it was seen that the difference between fitted and measured magnetic measurements in the equilibrium reconstruction were larger than in the constant current phase. External loop voltage measurements and magnetic probe pairs inside and

Improvements for real-time magnetic equilibrium reconstruction on ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Giannone, L.; Fischer, R. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); McCarthy, P.J. [Department of Physics, University College Cork, Cork (Ireland); Odstrcil, T.; Zammuto, I.; Bock, A.; Conway, G.; Fuchs, J.C.; Gude, A.; Igochine, V.; Kallenbach, A.; Lackner, K.; Maraschek, M.; Rapson, C. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Ruan, Q. [National Instruments, Austin, TX 78759-3504 (United States); Schuhbeck, K.H.; Suttrop, W. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Wenzel, L. [National Instruments, Austin, TX 78759-3504 (United States)

2015-11-15

Highlights: • Spline basis current functions with second-order linear regularisation. • Perturbations of magnetic probe measurements due to ferromagnetic tiles on the inner wall and from oscillations in the fast position coil current are corrected. • A constraint of the safety factor on the magnetic axis is introduced. Soft X-ray tomography is used to assess the quality of the real-time magnetic equilibrium reconstruction. • External loop voltage measurements and magnetic probe pairs inside and outside the vessel wall were used to measure the vacuum vessel wall resistivity. - Abstract: Real-time magnetic equilibria are needed for NTM stabilization and disruption avoidance experiments on ASDEX Upgrade. Five improvements to real-time magnetic equilibrium reconstruction on ASDEX Upgrade have been investigated. The aim is to include as many features of the offline magnetic equilibrium reconstruction code in the real-time equilibrium reconstruction code. Firstly, spline current density basis functions with regularization are used in the offline equilibrium reconstruction code, CLISTE [1]. It is now possible to have the same number of spline basis functions in the real-time code. Secondly, in the presence of edge localized modes, (ELM's), it is found to be necessary to include the low pass filter effect of the vacuum vessel on the fast position control coil currents to correctly compensate the magnetic probes for current oscillations in these coils. Thirdly, the introduction of ferromagnetic tiles in ASDEX Upgrade means that a real-time algorithm for including the perturbations of the magnetic equilibrium generated by these tiles is required. A methodology based on tile surface currents is described. Fourthly, during current ramps it was seen that the difference between fitted and measured magnetic measurements in the equilibrium reconstruction were larger than in the constant current phase. External loop voltage measurements and magnetic probe pairs inside

Equilibrium shoreface profiles

DEFF Research Database (Denmark)

Aagaard, Troels; Hughes, Michael G

2017-01-01

Large-scale coastal behaviour models use the shoreface profile of equilibrium as a fundamental morphological unit that is translated in space to simulate coastal response to, for example, sea level oscillations and variability in sediment supply. Despite a longstanding focus on the shoreface...... profile and its relevance to predicting coastal response to changing environmental conditions, the processes and dynamics involved in shoreface equilibrium are still not fully understood. Here, we apply a process-based empirical sediment transport model, combined with morphodynamic principles to provide......; there is no tuning or calibration and computation times are short. It is therefore easily implemented with repeated iterations to manage uncertainty....

Status of experimental data related to Be in ITER materials R and D data bank

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Shigeru [ITER Joint Central Team, Muenchen (Germany)

1998-01-01

To keep traceability of many valuable raw data that were experimentally obtained in the ITER Technology R and D Tasks related to materials for In-Vessel components (divertor, first wall, blanket, vacuum vessel, etc.) and to easily make the best use of these data in the ITER design activities, the `ITER Materials R and D Data Bank` has been built up, with the use of Excel{sup TM} spread sheets. The paper describes status of experimental data collected in this data bank on thermo-mechanical properties of unirradiated and neutron irradiated Be, on plasma-material interactions of Be, on mechanical properties of various kinds of Be/Cu joints (including plasma sprayed Be), and on thermal fatigue tests of Be/Cu mock-ups. (author)

Modeling and analysis of alternative concept of ITER vacuum vessel primary heat transfer system

International Nuclear Information System (INIS)

Carbajo, Juan; Yoder, Graydon; Dell'Orco, G.; Curd, Warren; Kim, Seokho

2010-01-01

A RELAP5-3D model of the ITER (Latin for 'the way') vacuum vessel (VV) primary heat transfer system has been developed to evaluate a proposed design change that relocates the heat exchangers (HXs) from the exterior of the tokamak building to the interior. This alternative design protects the HXs from external hazards such as wind, tornado, and aircraft crash. The proposed design integrates the VV HXs into a VV pressure suppression system (VVPSS) tank that contains water to condense vapour in case of a leak into the plasma chamber. The proposal is to also use this water as the ultimate sink when removing decay heat from the VV system. The RELAP5-3D model has been run under normal operating and abnormal (decay heat) conditions. Results indicate that this alternative design is feasible, with no effects on the VVPSS tank under normal operation and with tank temperature and pressure increasing under decay heat conditions resulting in a requirement to remove steam generated if the VVPSS tank low pressure must be maintained.

Structural analysis of ITER sub-assembly tools

International Nuclear Information System (INIS)

Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Lee, J.H.; Kim, K.K.; Im, K.; Shaw, R.

2011-01-01

The ITER Tokamak assembly tools are purpose-built assembly tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. The sector sub-assembly tools descried in this paper are main assembly tools to assemble vacuum vessel, thermal shield and toroidal filed coils into a complete 40 o sector. The 40 o sector sub-assembly tools are composed of sector sub-assembly tool, including radial beam, vacuum vessel supports and mid-plane brace tools. These tools shall have sufficient strength to transport and handle heavy weight of the ITER Tokamak machine reached several hundred tons. Therefore these tools should be designed and analyzed to confirm both the strength and structural stability even in the case of conservative assumptions. To verify structural stabilities of the sector sub-assembly tools in terms of strength and deflection, ANSYS code was used for linear static analysis. The results of the analysis show that these tools are designed with sufficient strength and stiffness. The conceptual designs of these tools are briefly described in this paper also.

Disruptions in ITER and strategies for their control and mitigation

Energy Technology Data Exchange (ETDEWEB)

Lehnen, M., E-mail: michael.lehnen@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Aleynikova, K.; Aleynikov, P.B.; Campbell, D.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Drewelow, P. [Max-Planck-Institut für Plasmaphysik, Greifswald branch, EURATOM Ass., D-17491 Greifswald (Germany); Eidietis, N.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Gasparyan, Yu. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoe sh. 31, Moscow 115409 (Russian Federation); Granetz, R.S. [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Gribov, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Hartmann, N. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research—Plasma Physics, Association EURATOM-FZJ, Trilateral Euregio Cluster, 52425 Jülich (Germany); Hollmann, E.M. [University of California-San Diego, La Jolla, CA 92093 (United States); Izzo, V.A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Jachmich, S. [Laboratory for Plasma Physics, ERM/KMS, Association EURATOM – Belgian State, B-1000 Brussels (Belgium); Kim, S.-H.; Kočan, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Koslowski, H.R. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research—Plasma Physics, Association EURATOM-FZJ, Trilateral Euregio Cluster, 52425 Jülich (Germany); Kovalenko, D. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Kruezi, U. [CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); and others

2015-08-15

The thermal and electromagnetic loads related to disruptions in ITER are substantial and require careful design of tokamak components to ensure they reach the projected lifetime and to ensure that safety relevant components fulfil their function for the worst foreseen scenarios. The disruption load specifications are the basis for the design process of components like the full-W divertor, the blanket modules and the vacuum vessel and will set the boundary conditions for ITER operations. This paper will give a brief overview on the disruption loads and mitigation strategies for ITER and will discuss the physics basis which is continuously refined through the current disruption R&D programs.

Structural analysis of ITER TBM Frame and Dummy TBM

International Nuclear Information System (INIS)

Marin, Anna; Kim, Byoung Yoon; Bertolini, Claudio; Lucca, Flavio; Komarov, Victor; Merola, Mario; Giancarli, Luciano; Gicquel, Stefan

2013-01-01

One of the main engineering performance goals of ITER is to test and validate design concepts of tritium breeding blankets. To accomplish these goals, three ITER equatorial ports are dedicated to the test of Test Blanket Modules (TBMs) that are mock-ups of tritium breeding blankets. These TBMs, associated with appropriate shield blocks, will also provide the same thermal and nuclear shielding as the main blanket. The main function of TBM Port Plug (PP) is to accommodate TBMs and provide a standardized interface with the vacuum vessel (VV)/port structure. The feasibility of the design concept of the Frame including two Dummy TBMs has been investigated by proposing design improvements of the reference design through an extensive set of thermal, electromagnetic (EM) and stress analyses. As well, the related static strength was evaluated in accordance with the structural design criteria for ITER in-vessel components (SDC-IC). This paper outlines the engineering aspects of the ITER TBM Frame and Dummy TBM and focuses on the feasibility of the present design by structural assessment

The ITER poloidal field system

Energy Technology Data Exchange (ETDEWEB)

Wesley, J [General Atomics, San Diego, CA (USA); Beljakov, V; Kavin, A; Korshakov, V; Kostenko, A; Roshal, A; Zakharov, L [Kurchatov Inst. of Atomic Energy, Moscow (USSR); Bulmer, R; Kaiser, T; Miller, J R; Pearlstein, L D [Lawrence Livermore National Lab., CA (USA); Hogan, J [Oak Ridge National Lab., TN (USA); Kurihara, K; Shimomura, Y; Sugihara, M; Yoshino, R [Japan Atomic Energy Resea

1990-12-15

The ITER poloidal field (PF) system uses superconducting coils to provide the plasma equilibrium fields, slow equilibrium control and plasma flux linkage (V-s) needed for the ITER Operations and Research Program. Double-null (DN) divertor plasmas and operation scenarios for 22 MA Physics (high-Q/ignition) and 15 MA Technology (high-fluence testing) phases are provided. For 22 MA plasmas, total PF flux swing is 333 V-s. This provides inductive current drive (CD) for start-up with 66 V-s of resistive loss and 440-s (330-s minimum) sustained burn. The PF system also allows plasma start-up and shutdown scenarios, and can maintain the plasma configuration during burn over a range of current and pressure profiles. Other capabilities include increased plasma current (25 MA with inductive CD; 28 MA with non-inductive CD assist), divertor separatrix sweeping, and semi-DN and single-null plasmas.

New applications of Equinox code for real-time plasma equilibrium and profile reconstruction for tokamaks

International Nuclear Information System (INIS)

Bosak, K.; Blum, J.; Joffrin, E.

2004-01-01

Recent development of real-time equilibrium code Equinox using a fixed-point algorithm allow major plasma magnetic parameters to be identified in real-time, using rigorous analytical method. The code relies on the boundary flux code providing magnetic flux values on the first wall of vacuum vessel. By means of least-square minimization of differences between magnetic field obtained from previous solution and the next measurements the code identifies the source term of the non-linear Grad-Shafranov equation. The strict use of analytical equations together with a flexible algorithm offers an opportunity to include new measurements into stable magnetic equilibrium code and compare the results directly between several tokamaks while maintaining the same physical model (i.e. no iron model is necessary inside the equilibrium code). The successful implementation of this equilibrium code for JET and Tore Supra has already been published. In this paper, we show the preliminary results of predictive runs of the Equinox code using the ITER geometry. Because the real-time control experiments of plasma profile at JET using the code has been shown unstable when using magnetic and polarimetric measurements (that could be indirectly translated into accuracy vs robustness tradeoff), we plan an outline of the algorithm that will allow us to further constrain the plasma current profile using the central value of pressure of the plasma in real-time in order to better define the poloidal beta (this constraint is not necessary with purely magnetic equilibrium). (authors)

New applications of Equinox code for real-time plasma equilibrium and profile reconstruction for tokamaks

Energy Technology Data Exchange (ETDEWEB)

Bosak, K.; Blum, J. [Universite de Nice-Sophia-Antipolis, Lab. J. A. Dieudonne, 06 - Nice (France); Joffrin, E. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

2004-07-01

Recent development of real-time equilibrium code Equinox using a fixed-point algorithm allow major plasma magnetic parameters to be identified in real-time, using rigorous analytical method. The code relies on the boundary flux code providing magnetic flux values on the first wall of vacuum vessel. By means of least-square minimization of differences between magnetic field obtained from previous solution and the next measurements the code identifies the source term of the non-linear Grad-Shafranov equation. The strict use of analytical equations together with a flexible algorithm offers an opportunity to include new measurements into stable magnetic equilibrium code and compare the results directly between several tokamaks while maintaining the same physical model (i.e. no iron model is necessary inside the equilibrium code). The successful implementation of this equilibrium code for JET and Tore Supra has already been published. In this paper, we show the preliminary results of predictive runs of the Equinox code using the ITER geometry. Because the real-time control experiments of plasma profile at JET using the code has been shown unstable when using magnetic and polarimetric measurements (that could be indirectly translated into accuracy vs robustness tradeoff), we plan an outline of the algorithm that will allow us to further constrain the plasma current profile using the central value of pressure of the plasma in real-time in order to better define the poloidal beta (this constraint is not necessary with purely magnetic equilibrium). (authors)

A wide angle view imaging diagnostic with all reflective, in-vessel optics at JET

Energy Technology Data Exchange (ETDEWEB)

Clever, M. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, 52425 Jülich (Germany); Arnoux, G.; Balshaw, N. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Garcia-Sanchez, P. [Laboratorio Nacional de Fusion, Asociacion EURATOM-CIEMAT, Madrid (Spain); Patel, K. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Sergienko, G. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, 52425 Jülich (Germany); Soler, D. [Winlight System, 135 rue Benjamin Franklin, ZA Saint Martin, F-84120 Pertuis (France); Stamp, M.F.; Williams, J.; Zastrow, K.-D. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

2013-10-15

Highlights: ► A new wide angle view camera system has been installed at JET. ► The system helps to protect the ITER-like wall plasma facing components from damage. ► The coverage of the vessel by camera observation systems was increased. ► The system comprises an in-vessel part with parabolic and flat mirrors. ► The required image quality for plasma monitoring and wall protection was delivered. -- Abstract: A new wide angle view camera system has been installed at JET in preparation for the ITER-like wall campaigns. It considerably increases the coverage of the vessel by camera observation systems and thereby helps to protect the – compared to carbon – more fragile plasma facing components from damage. The system comprises an in-vessel part with parabolic and flat mirrors and an ex-vessel part with beam splitters, lenses and cameras. The system delivered the image quality required for plasma monitoring and wall protection.

The ITER remote maintenance system

International Nuclear Information System (INIS)

Tesini, A.; Palmer, J.

2008-01-01

The aim of this paper is to summarize the ITER approach to machine components maintenance. A major objective of the ITER project is to demonstrate that a future power producing fusion device can be maintained effectively and offer practical levels of plant availability. During its operational lifetime, many systems of the ITER machine will require maintenance and modification; this can be achieved using remote handling methods. The need for timely, safe and effective remote operations on a machine as complex as ITER and within one of the world's most hostile remote handling environments represents a major challenge at every level of the ITER Project organization, engineering and technology. The basic principles of fusion reactor maintenance are presented. An updated description of the ITER remote maintenance system is provided. This includes the maintenance equipment used inside the vacuum vessel, inside the hot cell and the hot cell itself. The correlation between the functions of the remote handling equipment, of the hot cell and of the radwaste processing system is also described. The paper concludes that ITER has equipped itself with a good platform to tackle the challenges presented by its own maintenance and upgrade needs

ITER tokamak device

International Nuclear Information System (INIS)

Doggett, J.; Salpietro, E.; Shatalov, G.

1991-01-01

The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

The ITER tritium systems

International Nuclear Information System (INIS)

Glugla, M.; Antipenkov, A.; Beloglazov, S.; Caldwell-Nichols, C.; Cristescu, I.R.; Cristescu, I.; Day, C.; Doerr, L.; Girard, J.-P.; Tada, E.

2007-01-01

ITER is the first fusion machine fully designed for operation with equimolar deuterium-tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium-tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium. The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium-tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept. Not only the interfaces between the primary fuel cycle systems - being procured through different Participant Teams - but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping - again procured through different Participant Teams - and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task

Mechanical design of the ITER ion cyclotron heating launcher based on in-vessel tuning system

Energy Technology Data Exchange (ETDEWEB)

Vulliez, K. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 St Paul Lez Durance (France)], E-mail: karl.vulliez@cea.fr; Bosia, G. [Dipartimento di Fisica Generale, Universita di Torino (Italy); Agarici, G.; Beaumont, B.; Argouarch, A.; Mollard, P. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 St Paul Lez Durance (France); Testoni, P. [Electrical and Electronics Engineering Department, University of Cagliari (Italy); Maggiora, R.; Milanesio, D. [Dipartimento di Elettronica Politecnico di Torino (Italy)

2007-10-15

Since the release of the ITER ICRH system reference design report [ITER Final Design Report: DDD 5.1 -Ion Cyclotron and Current Drive System, July 2001], further design studies have been conducted. If the base of the reference design [Final Report on EFDA contract 04/1129, ITER ICRF antenna and Matching system design (Internalmatching), April 2005] is kept unchanged, several significant modifications have been proposed for a better efficiency and reliability. The increase of the poloidal order of the array and strong modifications of the matching system concept are the main changes. Technical aspects insufficiently covered in previous studies are also now worked out in detail, like the integration on a mid-plane port satisfying the constraints of the ITER environment.

Progress in physics basis and its impact on ITER

International Nuclear Information System (INIS)

Shimada, M.; Campbell, D.; Stambaugh, R.; Ide, S.; Kamada, Y.; Leonard, A.; Polevoi, A.; Mukhovatov, V.; Costley, A.E.; Gribov, Y.; Oikawa, T.; Sugihara, M.; Asakura, N.; Donne, A.J.H.; Doyle, E.J.; Federici, G.; Kukushkin, A.S.; Gormezano, C.; Gruber, O.; Houlberg, W.; Lipschultz, B.; Medvedev, S.

2005-01-01

This paper summarises recent progress in the physics basis and its impact on the expected performance of ITER. Significant progress has been made in many outstanding issues and in the development of hybrid and steady state operation scenarios, leading to increased confidence of achieving ITER's goals. Experiments show that tailoring the current profile can improve confinement over the standard H-mode and allow an increase in beta up to the no-wall limit at safety factors ∼ 4. Extrapolation to ITER suggests that at the reduced plasma current of ∼ 12MA, high Q > 10 and long pulse (>1000 s) operation is possible with benign ELMs. Analysis of disruption scenarios has been performed based on guidelines on current quench rates and halo currents, derived from the experimental database. With conservative assumptions, estimated electromagnetic forces on the in-vessel components are below the design target values, confirming the robustness of the ITER design against disruption forces. (author)

Beryllium application in ITER plasma facing components

International Nuclear Information System (INIS)

Raffray, A.R.; Federici, G.; Barabash, V.; Cardella, A.; Jakeman, R.; Ioki, K.; Janeschitz, G.; Parker, R.; Tivey, R.; Pacher, H.D.; Wu, C.H.; Bartels, H.W.

1997-01-01

Beryllium is a candidate armour material for the in-vessel components of the International Thermonuclear Experimental Reactor (ITER), namely the primary first wall, the limiter, the baffle and the divertor. However, a number of issues arising from the performance requirements of the ITER plasma facing components (PFCs) must be addressed to better assess the attractiveness of Be as armour for these different components. These issues include heat loading limits arising from temperature and stress constraints under steady state conditions, armour lifetime including the effects of sputtering erosion as well as vaporisation and loss of melt during disruption events, tritium retention and permeation, and chemical hazards, in particular with respect to potential Be/steam reaction. Other issues such as fabrication and the possibility of in-situ repair are not performance-dependent but have an important impact on the overall assessment of Be as PFC armour. This paper describes the present view on Be application for ITER PFCs. The key issues are discussed including an assessment of the current level of understanding based on analysis and experimental data; and on-going activities as part of the ITER EDA R and D program are highlighted. (orig.)

Design and technical status of the EU contribution to ITER

International Nuclear Information System (INIS)

Gasparotto, Maurizio; Federici, Gianfranco; Casci, Federico Riccardo

2009-01-01

Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R and D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost. This paper provides an update of the design and technical status of EU contributions to ITER.

Progress in Development of the ITER Plasma Control System Simulation Platform

Science.gov (United States)

Walker, Michael; Humphreys, David; Sammuli, Brian; Ambrosino, Giuseppe; de Tommasi, Gianmaria; Mattei, Massimiliano; Raupp, Gerhard; Treutterer, Wolfgang; Winter, Axel

2017-10-01

We report on progress made and expected uses of the Plasma Control System Simulation Platform (PCSSP), the primary test environment for development of the ITER Plasma Control System (PCS). PCSSP will be used for verification and validation of the ITER PCS Final Design for First Plasma, to be completed in 2020. We discuss the objectives of PCSSP, its overall structure, selected features, application to existing devices, and expected evolution over the lifetime of the ITER PCS. We describe an archiving solution for simulation results, methods for incorporating physics models of the plasma and physical plant (tokamak, actuator, and diagnostic systems) into PCSSP, and defining characteristics of models suitable for a plasma control development environment such as PCSSP. Applications of PCSSP simulation models including resistive plasma equilibrium evolution are demonstrated. PCSSP development supported by ITER Organization under ITER/CTS/6000000037. Resistive evolution code developed under General Atomics' Internal funding. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

The remote handling systems for ITER

Energy Technology Data Exchange (ETDEWEB)

Ribeiro, Isabel, E-mail: mir@isr.ist.utl.pt [Institute for Systems and Robotics/Instituto Superior Tecnico, Lisboa (Portugal); Damiani, Carlo [Fusion for Energy, Barcelona (Spain); Tesini, Alessandro [ITER Organization, Cadarache (France); Kakudate, Satoshi [ITER Tokamak Device Group, Japan Atomic Energy Agency, Ibaraki (Japan); Siuko, Mikko [VTT Systems Engineering, Tampere (Finland); Neri, Carlo [Associazione EURATOM ENEA, Frascati (Italy)

2011-10-15

The ITER remote handling (RH) maintenance system is a key component in ITER operation both for scheduled maintenance and for unexpected situations. It is a complex collection and integration of numerous systems, each one at its turn being the integration of diverse technologies into a coherent, space constrained, nuclearised design. This paper presents an integrated view and recent results related to the Blanket RH System, the Divertor RH System, the Transfer Cask System (TCS), the In-Vessel Viewing System, the Neutral Beam Cell RH System, the Hot Cell RH and the Multi-Purpose Deployment System.

Fusion Power measurement at ITER

Energy Technology Data Exchange (ETDEWEB)

Bertalot, L.; Barnsley, R.; Krasilnikov, V.; Stott, P.; Suarez, A.; Vayakis, G.; Walsh, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

2015-07-01

Nuclear fusion research aims to provide energy for the future in a sustainable way and the ITER project scope is to demonstrate the feasibility of nuclear fusion energy. ITER is a nuclear experimental reactor based on a large scale fusion plasma (tokamak type) device generating Deuterium - Tritium (DT) fusion reactions with emission of 14 MeV neutrons producing up to 700 MW fusion power. The measurement of fusion power, i.e. total neutron emissivity, will play an important role for achieving ITER goals, in particular the fusion gain factor Q related to the reactor performance. Particular attention is given also to the development of the neutron calibration strategy whose main scope is to achieve the required accuracy of 10% for the measurement of fusion power. Neutron Flux Monitors located in diagnostic ports and inside the vacuum vessel will measure ITER total neutron emissivity, expected to range from 1014 n/s in Deuterium - Deuterium (DD) plasmas up to almost 10{sup 21} n/s in DT plasmas. The neutron detection systems as well all other ITER diagnostics have to withstand high nuclear radiation and electromagnetic fields as well ultrahigh vacuum and thermal loads. (authors)

Enhanced nonlinear iterative techniques applied to a non-equilibrium plasma flow

Energy Technology Data Exchange (ETDEWEB)

Knoll, D.A.; McHugh, P.R. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1996-12-31

We study the application of enhanced nonlinear iterative methods to the steady-state solution of a system of two-dimensional convection-diffusion-reaction partial differential equations that describe the partially-ionized plasma flow in the boundary layer of a tokamak fusion reactor. This system of equations is characterized by multiple time and spatial scales, and contains highly anisotropic transport coefficients due to a strong imposed magnetic field. We use Newton`s method to linearize the nonlinear system of equations resulting from an implicit, finite volume discretization of the governing partial differential equations, on a staggered Cartesian mesh. The resulting linear systems are neither symmetric nor positive definite, and are poorly conditioned. Preconditioned Krylov iterative techniques are employed to solve these linear systems. We investigate both a modified and a matrix-free Newton-Krylov implementation, with the goal of reducing CPU cost associated with the numerical formation of the Jacobian. A combination of a damped iteration, one-way multigrid and a pseudo-transient continuation technique are used to enhance global nonlinear convergence and CPU efficiency. GMRES is employed as the Krylov method with Incomplete Lower-Upper(ILU) factorization preconditioning. The goal is to construct a combination of nonlinear and linear iterative techniques for this complex physical problem that optimizes trade-offs between robustness, CPU time, memory requirements, and code complexity. It is shown that a one-way multigrid implementation provides significant CPU savings for fine grid calculations. Performance comparisons of the modified Newton-Krylov and matrix-free Newton-Krylov algorithms will be presented.

Wall conditioning for ITER: Current experimental and modeling activities

Energy Technology Data Exchange (ETDEWEB)

Douai, D., E-mail: david.douai@cea.fr [CEA, IRFM, Association Euratom-CEA, 13108 St. Paul lez Durance (France); Kogut, D. [CEA, IRFM, Association Euratom-CEA, 13108 St. Paul lez Durance (France); Wauters, T. [LPP-ERM/KMS, Association Belgian State, 1000 Brussels (Belgium); Brezinsek, S. [FZJ, Institut für Energie- und Klimaforschung Plasmaphysik, 52441 Jülich (Germany); Hagelaar, G.J.M. [Laboratoire Plasma et Conversion d’Energie, UMR5213, Toulouse (France); Hong, S.H. [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Lomas, P.J. [CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Lyssoivan, A. [LPP-ERM/KMS, Association Belgian State, 1000 Brussels (Belgium); Nunes, I. [Associação EURATOM-IST, Instituto de Plasmas e Fusão Nuclear, 1049-001 Lisboa (Portugal); Pitts, R.A. [ITER International Organization, F-13067 St. Paul lez Durance (France); Rohde, V. [Max-Planck-Institut für Plasmaphysik, 85748 Garching (Germany); Vries, P.C. de [ITER International Organization, F-13067 St. Paul lez Durance (France)

2015-08-15

Wall conditioning will be required in ITER to control fuel and impurity recycling, as well as tritium (T) inventory. Analysis of conditioning cycle on the JET, with its ITER-Like Wall is presented, evidencing reduced need for wall cleaning in ITER compared to JET–CFC. Using a novel 2D multi-fluid model, current density during Glow Discharge Conditioning (GDC) on the in-vessel plasma-facing components (PFC) of ITER is predicted to approach the simple expectation of total anode current divided by wall surface area. Baking of the divertor to 350 °C should desorb the majority of the co-deposited T. ITER foresees the use of low temperature plasma based techniques compatible with the permanent toroidal magnetic field, such as Ion (ICWC) or Electron Cyclotron Wall Conditioning (ECWC), for tritium removal between ITER plasma pulses. Extrapolation of JET ICWC results to ITER indicates removal comparable to estimated T-retention in nominal ITER D:T shots, whereas GDC may be unattractive for that purpose.

ITER status, design and material objectives

International Nuclear Information System (INIS)

Aymar, R.

2002-01-01

During the ITER Engineering Design Activities (EDA), completed in July 2001, the Joint Central Team and Home Teams developed a robust design of ITER, summarised in this paper, with parameters which fully meet the required scientific and technological objectives, construction costs and safety requirements, with appropriate margins. The design is backed by R and D to qualify the technology, including materials R and D. Materials for ITER components have been selected largely because of their availability and well-established manufacturing technologies, taking account of the low fluence experienced during neutron irradiation, and the experimental nature of the device. Nevertheless, for specific needs relevant to a future fusion reactor, improved materials, in particular for magnet structures, in-vessel components, and joints between the different materials needed for plasma facing components, have been successfully developed. Now, with the technical readiness to decide on ITER construction, negotiations, supported by coordinated technical activities of an international team and teams from participant countries, are underway on joint construction of ITER with a view to the signature/ratification of an agreement in 2003

Absorbed decay-photon dose analysis of the IVVS/GDC plug in ITER

Energy Technology Data Exchange (ETDEWEB)

Leichtle, D.; Serikov, A.; Fischer, U. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (DE). Inst. for Neutron Physics and Reactor Technology (INR)

2011-07-01

The In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit share a common port at the equatorial level of the ITER tokamak. The plug consists mainly of the IWS probe, capable of performing the laser-based in-vessel viewing and metrology, the GDC electrode, capable of producing glow discharge in the vacuum vessel during intermediate maintenance and wall conditioning periods, and their respective deployment systems to move the electrodes. The plug extends over a length of about 11 m from the GDC tip to the rear end at the bioshield level. At the present stage of the conceptual design a neutronics analysis has been requested to provide valuable input to the design strategy. To this end, a first assessment has been performed focusing on operational loads on the GDC electrode head in the so-called shielding position and on absorbed decay-photon dose rate levels in the structural components of the entire system. In this contribution we are reporting on the absorbed dose rates after the ITER life time irradiation at several cooling times. Gamma sources from activated materials of the IVVS/GDC and surrounding structures, like blanket, vacuum vessel, toroidal and poloidal field coils, have been taken into account. (orig.)

Fabrication of divertor cassette for ITER

International Nuclear Information System (INIS)

Sanguinetti, G.P.

2008-01-01

The Divertor is the component located on the bottom of the ITER vacuum vessel, whose main function is to adsorb the high thermal flux generated by the plasma whilst keeping the plasma impurity at a reasonable low level. The divertor consist of 54 units, each comprising outer components, facing the plasma and a component supporting the plasma facing components (PFC) and providing coolant distribution to them (divertor cassette). The divertor cassette is a box structure, butt welded and machined, made from plates and forgins of austenitic stainless steels. The cassette fabrication, which is in detail described, includes manufacturing of the attachments of the PFC to the cassette, the coolant distribution channels, and the cassette to vacuum vessel locking system. The divertor cassette is a pressure component (the cooling water runs at 40 bar) and therefore divertor cassette design, fabrication and service shall comply with the European PED and the applicable French law for the ITER. (orig.)

Exploring Chemical and Thermal Non-equilibrium in Nitrogen Arcs

International Nuclear Information System (INIS)

Ghorui, S; Das, A K

2012-01-01

Plasma torches operating with nitrogen are of special importance as they can operate with usual tungsten based refractory electrodes and offer radical rich non-oxidizing high temperature environment for plasma chemistry. Strong gradients in temperature as well as species densities and huge convective fluxes lead to varying degrees of chemical non-equilibrium in associated regions. An axi-symmetric two-temperature chemical non-equilibrium model of a nitrogen plasma torch has been developed to understand the effects of thermal and chemical non-equilibrium in arcs. A 2-D finite volume CFD code in association with a non-equilibrium property routine enabled extraction of steady state self-consistent distributions of various plasma quantities inside the torch under various thermal and chemical non-equilibrium conditions. Chemical non-equilibrium has been incorporated through computation of diffusive and convective fluxes in each finite volume cell in every iteration and associating corresponding thermodynamic and transport properties through the scheme of 'chemical non-equilibrium parameter' introduced by Ghorui et. al. Recombination coefficient data from Nahar et. al. and radiation data from Krey and Morris have been used in the simulation. Results are presented for distributions of temperature, pressure, velocity, current density, electric potential, species densities and chemical non-equilibrium effects. Obtained results are compared with similar results under LTE.

One-group constant libraries for nuclear equilibrium state

Energy Technology Data Exchange (ETDEWEB)

Mizutani, Akihiko; Sekimoto, Hiroshi [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

1997-03-01

One-group constant libraries for the nuclear equilibrium state were generated for both liquid sodium cooled MOX fuel type fast reactor and PWR type thermal reactor with Equilibrium Cell Iterative Calculation System (ECICS) using JENDL-3.2, -3, -2 and ENDF/B-VI nuclear data libraries. ECICS produced one-group constant sets for 129 heavy metal nuclides and 1238 fission products. (author)

Constructing integrable full-pressure full-current free-boundary stellarator magnetohydrodynamic equilibrium solutions

International Nuclear Information System (INIS)

Hudson, S.R.

2002-01-01

For stellarators to be feasible candidates for fusion power stations it is essential that the magnetic field lines lie on nested flux surfaces; however, the lack of a continuous symmetry implies that magnetic islands, caused by Pfirsch-Schlueter currents, diamagnetic currents and resonant coil fields, are guaranteed to exist. The challenge is to design the plasma and coils such that these effects cancel. Magnetic islands in free-boundary full-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the PIES code [Comp. Phys. Comm., 43:157, 1986] which iterates the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to lie in the nullspace of certain measures of engineering acceptability and kink stability. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible. The method is applied to a candidate plasma and coil design for NCSX [Phys. Plas., 7:1911, 2000]. (author)

Micro fission chamber for the ITER neutron monitor

International Nuclear Information System (INIS)

Yamauchi, Michinori; Nishitani, Takeo; Ochiai, Kentaro; Ebisawa, Katsuyuki

2004-01-01

This paper describes the design and the fabrication of a prototype micro-fission chamber and test results under ITER relevant conditions including wide neutron spectrum and intense gamma-rays, and the performance as a ITER power monitor is discussed. A micro-fission chamber with 12 mg UO 2 and a dummy chamber without uranium were designed and fabricated for the in-vessel neutron flux monitoring of ITER. The measurement ability was tested with the FNS facility for 14 MeV neutrons and the 60 Co gamma-ray irradiation facility at JAERI-Takasaki. Employing the Campbelling mode in the electronics, the ITER requirement for the temporal resolution was satisfied. The excellent linearity of the detector output versus the neutron flux was confirmed in the temperature range from 20degC to 250degC. As a result, it was concluded that the developed micro-fission chamber is applicable for ITER. (author)

Design of ITER neutron monitor using micro fission chambers

International Nuclear Information System (INIS)

Nishitani, Takeo; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi; Johnson, L.C.; Walker, C.

1998-08-01

We are designing micro fission chambers, which are pencil size gas counters with fissile material inside, to be installed in the vacuum vessel as neutron flux monitors for ITER. We found that the 238 U micro fission chambers are not suitable because the detection efficiency will increase up to 50% in the ITER life time by breading 239 Pu. We propose to install 235 U micro fission chambers on the front side of the back plate in the gap between adjacent blanket modules and behind the blankets at 10 poloidal locations. One chamber will be installed in the divertor cassette just under the dome. Employing both pulse counting mode and Campbelling mode in the electronics, we can accomplish the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution, and eliminate the effect of gamma-rays. We demonstrate by neutron Monte Carlo calculation with three-dimensional modeling that we avoid those detection efficiency changes by installing micro fission chambers at several poloidal locations inside the vacuum vessel. (author)

Software design of the hybrid robot machine for ITER vacuum vessel assembly and maintenance

Energy Technology Data Exchange (ETDEWEB)

Li, Ming, E-mail: Ming.Li@lut.fi [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Wu, Huapeng; Handroos, Heikki [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Yang, Guangyou [School of Mechanical Engineering, Hubei University of Technology, Wuhan (China)

2013-10-15

A specific software design is elaborated in this paper for the hybrid robot machine used for the ITER vacuum vessel (VV) assembly and maintenance. In order to provide the multi-machining-function as well as the complicated, flexible and customizable GUI designing satisfying the non-standardized VV assembly process in one hand, and in another hand guarantee the stringent machining precision in the real-time motion control of robot machine, a client–server-control software architecture is proposed, which separates the user interaction, data communication and robot control implementation into different software layers. Correspondingly, three particular application protocols upon the TCP/IP are designed to transmit the data, command and status between the client and the server so as to deal with the abundant data streaming in the software. In order not to be affected by the graphic user interface (GUI) modification process in the future experiment in VV assembly working field, the real-time control system is realized as a stand-alone module in the architecture to guarantee the controlling performance of the robot machine. After completing the software development, a milling operation is tested on the robot machine, and the result demonstrates that both the specific GUI operability and the real-time motion control performance could be guaranteed adequately in the software design.

Software design of the hybrid robot machine for ITER vacuum vessel assembly and maintenance

International Nuclear Information System (INIS)

Li, Ming; Wu, Huapeng; Handroos, Heikki; Yang, Guangyou

2013-01-01

A specific software design is elaborated in this paper for the hybrid robot machine used for the ITER vacuum vessel (VV) assembly and maintenance. In order to provide the multi-machining-function as well as the complicated, flexible and customizable GUI designing satisfying the non-standardized VV assembly process in one hand, and in another hand guarantee the stringent machining precision in the real-time motion control of robot machine, a client–server-control software architecture is proposed, which separates the user interaction, data communication and robot control implementation into different software layers. Correspondingly, three particular application protocols upon the TCP/IP are designed to transmit the data, command and status between the client and the server so as to deal with the abundant data streaming in the software. In order not to be affected by the graphic user interface (GUI) modification process in the future experiment in VV assembly working field, the real-time control system is realized as a stand-alone module in the architecture to guarantee the controlling performance of the robot machine. After completing the software development, a milling operation is tested on the robot machine, and the result demonstrates that both the specific GUI operability and the real-time motion control performance could be guaranteed adequately in the software design

Nash equilibrium in differential games and the construction of the programmed iteration method

International Nuclear Information System (INIS)

Averboukh, Yurii V

2011-01-01

This work is devoted to the study of nonzero-sum differential games. The set of payoffs in a situation of Nash equilibrium is examined. It is shown that the set of payoffs in a situation of Nash equilibrium coincides with the set of values of consistent functions which are fixed points of the program absorption operator. A condition for functions to be consistent is given in terms of the weak invariance of the graph of the functions under a certain differential inclusion. Bibliography: 18 titles.

Intelligent tit-for-tat in the iterated prisoner's dilemma game

Science.gov (United States)

Baek, Seung Ki; Kim, Beom Jun

2008-07-01

We seek a route to the equilibrium where all the agents cooperate in the iterated prisoner’s dilemma game on a two-dimensional plane, focusing on the role of tit-for-tat strategy. When a time horizon, within which a strategy can recall the past, is one time step, an equilibrium can be achieved as cooperating strategies dominate the whole population via proliferation of tit-for-tat. Extending the time horizon, we filter out poor strategies by simplified replicator dynamics and observe a similar evolutionary pattern to reach the cooperating equilibrium. In particular, the rise of a modified tit-for-tat strategy plays a central role, which implies how a robust strategy is adopted when provided with an enhanced memory capacity.

Seismic analysis of ITER multi-purpose deployer

International Nuclear Information System (INIS)

Manuelraj, Manoah Stephen; Gotewal, Krishan Kumar; Dutta, Pramit; Rastogi, Naveen; Choi, Chang-Hwan; Tesini, Alessandro

2015-01-01

The Multi-Purpose Deployer (MPD) is a general purpose ITER in-vessel remote handling (RH) system. The MPD will perform various in-vessel maintenance tasks such as dust and tritium inventory control, in-service inspection, leak localization and in-vessel diagnostics maintenance. The main handling equipment, called as the MPD Transporter, consists of a series of linked bodies, which provide anchoring to the vacuum vessel port and an articulated multi-degree of freedom motion to perform the aforementioned tasks. The target payload for the MPD Transporter is 2 tons. The total length is 16.6 m and 18.1 m for short and long configuration respectively, while the total weight of the system is about 25.5 tons including the payload. During the in-vessel operations, the structural integrity of the system should be guaranteed against various operational and seismic loads. This paper presents the seismic structural analysis results of the concept design of the MPD Transporter. Static structural, modal and frequency response spectrum analyses have been performed to verify the structural integrity of the MPD itself, and to provide reaction loads to the interfacing systems such as vacuum vessel and cask. The analyses are carried out by using the ANSYS. The first analysis iteration was carried out for the reference design of the MPD Transporter, which showed stresses higher than the permissible limit. Structural optimizations and reinforcements were performed for individual bodies referring the stress levels in each body, and a reinforced design was proposed. The reinforced design satisfies the required structural criteria in terms of general global stresses. Though local stress concentrations were observed, it can be solved in the next design phase by further local reinforcements or proper material choice. (author)

Status of ITER

International Nuclear Information System (INIS)

Aymar, R.

2002-01-01

At the end of engineering design activities (EDA) in July 2001, all the essential elements became available to make a decision on construction of ITER. A sufficiently detailed and integrated engineering design now exists for a generic site, has been assessed for feasibility, and costed, and essential physics and technology R and D has been carried out to underpin the design choices. Formal negotiations have now begun between the current participants--Canada, Euratom, Japan, and the Russian Federation--on a Joint Implementation Agreement for ITER which also establishes the legal entity to run ITER. These negotiations are supported on technical aspects by Coordinated Technical Activities (CTA), which maintain the integrity of the project, for the good of all participants, and concentrate on preparing for procurement by industry of the longest lead items, and for formal application for a construction license with the host country. This paper highlights the main features of the ITER design. With cryogenically-cooled magnets close to neutron-generating plasma, the design of shielding with adequate access via port plugs for auxiliaries such as heating and diagnostics, and of remote replacement and refurbishing systems for in-vessel components, are particularly interesting nuclear technology challenges. Making a safety case for ITER to satisfy potential regulators and to demonstrate, as far as possible at this stage, the environmental attractiveness of fusion as an energy source, is also important. The paper gives illustrative details on this work, and an update on the progress of technical preparations for construction, as well as the status of the above negotiations

Constructing Integrable High-pressure Full-current Free-boundary Stellarator Magnetohydrodynamic Equilibrium Solutions

International Nuclear Information System (INIS)

Hudson, S.R.; Monticello, D.A.; Reiman, A.H.; Strickler, D.J.; Hirshman, S.P.; Ku, L-P; Lazarus, E.; Brooks, A.; Zarnstorff, M.C.; Boozer, A.H.; Fu, G-Y.; Neilson, G.H.

2003-01-01

For the (non-axisymmetric) stellarator class of plasma confinement devices to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux surfaces; however, the inherent lack of a continuous symmetry implies that magnetic islands responsible for breaking the smooth topology of the flux surfaces are guaranteed to exist. Thus, the suppression of magnetic islands is a critical issue for stellarator design, particularly for small aspect ratio devices. Pfirsch-Schluter currents, diamagnetic currents, and resonant coil fields contribute to the formation of magnetic islands, and the challenge is to design the plasma and coils such that these effects cancel. Magnetic islands in free-boundary high-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the Princeton Iterative Equilibrium Solver [Reiman and Greenside, Comp. Phys. Comm. 43 (1986) 157] which iterate s the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to preserve certain measures of engineering acceptability and to preserve the stability of ideal kink modes. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible, the plasma is stable to ideal kink modes, and the coils satisfy engineering constraints. The method is applied to a candidate plasma and coil design for the National Compact Stellarator Experiment [Reiman, et al., Phys. Plasmas 8 (May 2001) 2083

Constructing integrable high-pressure full-current free-boundary stellarator magnetohydrodynamic equilibrium solutions

International Nuclear Information System (INIS)

Hudson, S.R.; Monticello, D.A.; Reiman, A.H.

2003-01-01

For the (non-axisymmetric) stellarator class of plasma confinement devices to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux surfaces; however, the inherent lack of a continuous symmetry implies that magnetic islands responsible for breaking the smooth topology of the flux surfaces are guaranteed to exist. Thus, the suppression of magnetic islands is a critical issue for stellarator design, particularly for small aspect ratio devices. Pfirsch-Schlueter currents, diamagnetic currents and resonant coil fields contribute to the formation of magnetic islands, and the challenge is to design the plasma and coils such that these effects cancel. Magnetic islands in free-boundary high-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the Princeton Iterative Equilibrium Solver (Reiman and Greenside 1986 Comput. Phys. Commun. 43 157) which iterates the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to preserve certain measures of engineering acceptability and to preserve the stability of ideal kink modes. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible, the plasma is stable to ideal kink modes, and the coils satisfy engineering constraints. The method is applied to a candidate plasma and coil design for the National Compact Stellarator eXperiment (Reiman et al 2001 Phys. Plasma 8 2083). (author)

Design of ITER shielding blanket

International Nuclear Information System (INIS)

Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

1997-05-01

A mechanical configuration of ITER integrated primary first wall/shield blanket module were developed focusing on the welded attachment of its support leg to the back plate. A 100 mm x 150 mm space between the legs of adjacent modules was incorporated for the working space of welding/cutting tools. A concept of coolant branch pipe connection to accommodate deformation due to the leg welding and differential displacement of the module and the manifold/back plate during operation was introduced. Two-dimensional FEM analyses showed that thermal stresses in Cu-alloy (first wall) and stainless steel (first wall coolant tube and shield block) satisfied the stress criteria following ASME code for ITER BPP operation. On the other hand, three-dimensional FEM analyses for overall in-vessel structures exhibited excessive primary stresses in the back plate and its support structure to the vacuum vessel under VDE disruption load and marginal stresses in the support leg of module No.4. Fabrication procedure of the integrated primary first wall/shield blanket module was developed based on single step solid HIP for the joining of Cu-alloy/Cu-alloy, Cu-alloy/stainless steel, and stainless steel/stainless steel. (author)

A Conceptual Design and Structural Analysis for ITER Mid-plane Brace Tools

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Kihak; Robert, Shaw [ITER Organization, St Paul lez Durance Cedex (France)

2010-10-15

The ITER, International Thermonuclear Experimental Reactor, Tokamak machine is mainly composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The ITER Tokamak assembly tools are purpose-built tools to assemble the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the IO (ITER international organization), Korea has carried out the conceptual design of assembly tools with IO cooperation. The 40 .deg. sector assemblies attached mid-plane brace tools sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. The sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and mid-plane brace tools. The mid-plane brace tool is assembled to inner surface of VV and TFCs in phase of sector sub-assembly after completion of all sector components. VV, TFC and VVTS are separated fully before completion of 9 sectors at Tokamak in-pit. In this paper the mid-plane brace tools is introduced about function, structure and status of research and development are also described

Progress in the integration of Test Blanket Systems in ITER equatorial port cells and in the interfaces definition

Energy Technology Data Exchange (ETDEWEB)

Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Beloglazov, S.; Bonagiri, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Commin, L. [CEA, IRFM, Cadarache (France); Cortes, P.; Giancarli, L.M.; Gliss, C.; Iseli, M.; Lanza, R.; Levesy, B.; Martins, J.-P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Neviere, J.-C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L.; Plutino, D.; Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Swami, H.L. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

2012-08-15

Highlights: Black-Right-Pointing-Pointer The design integration of two test blanket systems in ITER port cell is addressed. Black-Right-Pointing-Pointer Definition of interfaces of TBSs with building and other ITER systems is done. Black-Right-Pointing-Pointer Designs of pipe forest, bioshield plug and ancillary equipment unit are described. Black-Right-Pointing-Pointer The maintenance of the two test blanket systems in ITER port cell is considered. Black-Right-Pointing-Pointer The management of the heat and tritium releases in the TBM port cell is described. - Abstract: In the framework of the TBM Program, three ITER vacuum vessel equatorial ports (no. 16, no. 18 and no. 02) have been allocated for the testing of up to six mock-ups of six different DEMO tritium breeding blankets. Each one is called a Test Blanket System (TBS). A TBS consists mainly of the Test Blanket Module (TBM), the in-vessel component facing the plasma, and several ancillary systems, in particular the cooling system and the tritium extraction system. Each port accommodates two TBMs and therefore the two TBSs have to share the corresponding port cell. This paper deals with the design integration aspects of the two TBSs in each port cell performed at ITER Organization (IO) with the corresponding definition of interfaces with other ITER systems. The performed activities have raised several issues that are discussed in the paper and for which design solutions are proposed.

Implementation of GPU parallel equilibrium reconstruction for plasma control in EAST

Energy Technology Data Exchange (ETDEWEB)

Huang, Yao, E-mail: yaohuang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); School of Nuclear Science & Technology, University of Science & Technology of China (China); Luo, Z.P.; Yuan, Q.P.; Pei, X.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yue, X.N. [School of Nuclear Science & Technology, University of Science & Technology of China (China)

2016-11-15

Highlights: • We described parallel equilibrium reconstruction code P-EFIT running on GPU was integrated with EAST plasma control system. • Compared with RT-EFIT used in EAST, P-EFIT has better spatial resolution and full algorithm of EFIT per iteration. • With the data interface through RFM, 65 × 65 spatial grids P-EFIT can satisfy the accuracy and time feasibility requirements for plasma control. • Successful control using ISOFLUX/P-EFIT was established in the dedicated experiment during the EAST 2014 campaign. • This work is a stepping-stone towards versatile ISOFLUX/P-EFIT control, such as real-time equilibrium reconstruction with more diagnostics. - Abstract: Implementation of P-EFIT code for plasma control in EAST is described. P-EFIT is based on the EFIT framework, but built with the CUDA™ architecture to take advantage of massively parallel Graphical Processing Unit (GPU) cores to significantly accelerate the computation. 65 × 65 grid size P-EFIT can complete one reconstruction iteration in 300 μs, with one iteration strategy, it can satisfy the needs of real-time plasma shape control. Data interface between P-EFIT and PCS is realized and developed by transferring data through RFM. First application of P-EFIT to discharge control in EAST is described.

Development of parallellized higher-order generalized depletion perturbation theory for application in equilibrium cycle optimization

Energy Technology Data Exchange (ETDEWEB)

Geemert, R. van E-mail: rene.vangeemert@psi.ch; Hoogenboom, J.E. E-mail: j.e.hoogenboom@iri.tudelft.nl

2001-09-01

As nuclear fuel economy is basically a multi-cycle issue, a fair way of evaluating reload patterns is to consider their performance in the case of an equilibrium cycle. The equilibrium cycle associated with a reload pattern is defined as the limit fuel cycle that eventually emerges after multiple successive periodic refueling, each time implementing the same reload scheme. Since the equilibrium cycle is the solution of a reload operation invariance equation, it can in principle be found with sufficient accuracy only by applying an iterative procedure, simulating the emergence of the limit cycle. For a design purpose such as the optimization of reload patterns, in which many different equilibrium cycle perturbations (resulting from many different limited changes in the reload operator) must be evaluated, this requires far too much computational effort. However, for very fast calculation of these many different equilibrium cycle perturbations it is also possible to set up a generalized variational approach. This approach results in an iterative scheme that yields the exact perturbation in the equilibrium cycle solution as well, in an accelerated way. Furthermore, both the solution of the adjoint equations occurring in the perturbation theory formalism and the implementation of the optimization algorithm have been parallellized and executed on a massively parallel machine. The combination of parallellism and generalized perturbation theory offers the opportunity to perform very exhaustive, fast and accurate sampling of the solution space for the equilibrium cycle reload pattern optimization problem.

Full Tokamak discharge simulation and kinetic plasma profile control for ITER

International Nuclear Information System (INIS)

Hee Kim, S.

2009-10-01

Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode

Efficient approach to simulate EM loads on massive structures in ITER machine

Energy Technology Data Exchange (ETDEWEB)

Alekseev, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Andreeva, Z.; Belov, A.; Belyakov, V.; Filatov, O. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Gribov, Yu.; Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Kukhtin, V.; Labusov, A.; Lamzin, E.; Lyublin, B.; Malkov, A.; Mazul, I. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Rozov, V.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Sychevsky, S., E-mail: sytch@sintez.niiefa.spb.su [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation)

2013-10-15

Highlights: ► A modelling technique to predict EM loads in ITER conducting structures is presented. ► The technique provides low computational cost and parallel computations. ► Detailed models were built for the system “vacuum vessel, cryostat, thermal shields”. ► EM loads on massive in-vessel structures were simulated with the use of local models. ► A flexible combination of models enables desired accuracy of load distributions. -- Abstract: Operation of the ITER machine is associated with high electromagnetic (EM) loads. An essential contributor to EM loads is eddy currents induced in passive conductive structures. Reasoning from the ITER construction, a modelling technique has been developed and applied in computations to efficiently predict anticipated loads. The technique allows us to avoid building a global 3D finite-element (FE) model that requires meshing of the conducting structures and their vacuum environment into 3D solid elements that leads to high computational cost. The key features of the proposed technique are: (i) the use of an existing shell model for the system “vacuum vessel (VV), cryostat, and thermal shields (TS)” implementing the magnetic shell approach. A solution is obtained in terms of a single-component, in this case, vector electric potential taken within the conducting shells of the “VV + cryostat + TS” system. (ii) EM loads on in-vessel conducting structures are simulated with the use of local FE models. The local models use either the 3D solid body or shell approximations. Reasoning from the simulation efficiency, the local boundary conditions are put with respect to the total field or an external field. The use of an integral-differential formulation and special procedures ensures smooth and accurate simulated distributions of fields from current sources of any geometry. The local FE models have been developed and applied for EM analyses of a variety of the ITER components including the diagnostic systems

Control of dust production in ITER

International Nuclear Information System (INIS)

Rodriguez-Rodrigo, L.; Ciattaglia, S.; Elbez-Uzan, J.

2006-01-01

In the last years dust has been observed in a number of fusion devices and is being studied more in detail for understanding in particular the physical phenomena related to its formation, its composition, physical and chemical characteristics, and the amount of produced dust. The extrapolation of dust formation to ITER predicts (with large error bars), a large mass of dust production with a scattered size distribution. To evaluate the impact of dust on safety, assumptions have also been made on radionuclide inventory, and mobility in off-normal events, as well as any postulated contributions the dust may make to effluents or accidental releases. Solid activation products in structures are generally not readily mobilisable in incidental and accidental situations, so that activated dust, tritium and activated corrosions products are the important in-vessel source terms in postulated scenarios that assume a mobilisation and release of some fraction of this inventory. Such a release would require the simultaneous leak or bypass of several robust confinement barriers. Further concerns for dust may be the potential for chemical reactions between dust and coolant in the event of an in-vessel leak, and the theoretical possibility of a dust explosion, either of which could in principle cause a pressure rise that challenges one or more of the confinement barriers. Although these hazards can - and will - be controlled by other measures in the ITER design, application of the principle of Defence in Depth dictates that the dust inventory should also be minimised and controlled to prevent the potential hazard. A well-coordinated R-and-D programme is required to support this dust production control. This document provides from the safety point of view, an overview of existing data given in '' Dossier d'Options de Surete '', the first safety report presented in 2001 to the French Safety Authorities, and ITER documents; it also gathers information on status of studies on activated

MHD equilibrium identification on ASDEX-Upgrade

International Nuclear Information System (INIS)

McCarthy, P.J.; Schneider, W.; Lakner, K.; Zehrfeld, H.P.; Buechl, K.; Gernhardt, J.; Gruber, O.; Kallenbach, A.; Lieder, G.; Wunderlich, R.

1992-01-01

A central activity accompanying the ASDEX-Upgrade experiment is the analysis of MHD equilibria. There are two different numerical methods available, both using magnetic measurements which reflect equilibrium states of the plasma. The first method proceeds via a function parameterization (FP) technique, which uses in-vessel magnetic measurements to calculate up to 66 equilibrium parameters. The second method applies an interpretative equilibrium code (DIVA) for a best fit to a different set of magnetic measurements. Cross-checks with the measured particle influxes from the inner heat shield and the divertor region and with visible camera images of the scrape-off layer are made. (author) 3 refs., 3 figs

Ratcheting problems for ITER [International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Majumdar, S.

1991-01-01

Because of the presence of high cyclic thermal stress, pressure-induced primary stress, and disruption-induced high cyclic primary stress, ratcheting of the first wall poses a serious challenge to the designers of ITER (International Thermonuclear Experimental Reactor). Existing design tools such as the Bree diagram in the ASME Boiler and Pressure Vessels Code, are not directly applicable to ITER, because of important differences in geometry and loading modes. Available alternative models for ratcheting are discussed and new Bree diagrams, that are more relevant for fusion reactor applications, are proposed. 9 refs., 17 figs

A basic study on the ITER tritium storage vessel design and components

International Nuclear Information System (INIS)

Chung, H. S.; Ahn, D. H.; Kim, K. R.; Yim, S. P.; Paek, S. W.; Lee, M. S.; Lee, S. H.; Shim, M. H.

2006-01-01

The ZrCo getter beds are built of a primary vessel which contains the ZrCo powder mixed with Cu spheres of less than one mm diameter and of a secondary outer vessel. The purpose of the secondary outer vessel is to capture permeated or leaked tritium and to present a good thermal insulation when properly evacuated. A third volume, a helium filled loop, is installed in the primary volume to remove the decay heat and is used to perform tritium accountancy measurements

Multiple Depots Vehicle Routing Problem in the Context of Total Urban Traffic Equilibrium

Directory of Open Access Journals (Sweden)

Dongxu Chen

2017-01-01

Full Text Available A multidepot VRP is solved in the context of total urban traffic equilibrium. Under the total traffic equilibrium, the multidepot VRP is changed to GDAP (the problem of Grouping Customers + Estimating OD Traffic + Assigning traffic and bilevel programming is used to model the problem, where the upper model determines the customers that each truck visits and adds the trucks’ trips to the initial OD (Origin/Destination trips, and the lower model assigns the OD trips to road network. Feedback between upper model and lower model is iterated through OD trips; thus total traffic equilibrium can be simulated.

Status of Conceptual Design Progress for ITER Sector Sub-assembly Tools

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung Kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Ki Hak; Robert, Shaw [ITER Organization, Paul lez Durance (France)

2010-05-15

The ITER (International Thermonuclear Experimental Reactor) Tokamak assembly tools are purpose-built tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the ITER organization, Korea has carried out the conceptual design of assembly tools. The 40 .deg. sector assemblies sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. In-pit assembly tools are the purpose-built assembly tools for the completion of final sector assembly at Tokamak hall. The 40 .deg. sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and bracing tools. The process of the ITER sector sub-assembly at assembly hall and status of research and development are described in this paper. The ITER Tokamak device is composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The 40 .deg. sectors are sub-assembled at assembly hall respectively and then 9 sectors which sub-assembled at assembly hall are finally assembled at Tokamak hall. As a basic assembly component, the assembly strategy and tools for the 40 .deg. sector sub-assembly and final assembly at inpit should be developed to satisfy the basic assembly requirements of the ITER Tokamak device. Accordingly, the purpose-built assembly tools should be designed and manufactured considering assembly plan, available space, safety, easy operation, efficient maintenance, and so on. The 40 .deg. sector assembly tools are classified into 2 groups. One group is the sub-assembly tools including upending tool, lifting tool, sub-assembly tool, VV supports and bracing tools used at assembly hall and the other group is the in

Vibration control of an IVVS long-reach deployer using unknown visual features from inside the ITER vessel

International Nuclear Information System (INIS)

Dubus, G.; David, O.; Measson, Y.

2010-01-01

The In-Vessel Viewing System (IVVS) project assumes that a long reach deployer equipped with a probe penetrates the ITER chamber to perform periodic inspections. By giving the operator the capability and flexibility to examine unplanned targets, man-in-the-loop technology would be very helpful. But vibrations due to the high flexibility of the structure are probably the main problem in such a master-slave mode, which therefore needs the integration of a high level compensation scheme. However the ITER RH equipment will be confronted with strong electromagnetic interferences as well as a cumulated radiation dose up to several MGy. Short of costly developments, these constraints limit the use of dedicated electronics such as accelerometers or strain gauges. Our main idea is to control the vibrational behaviour of the flexible carrier without considering any extra sensor apart from its embedded probe. In this pre-study we propose to use the kind of rad-hardened viewing system already developed for the AIA demonstrator in order to feed an oscillation observer with visual information. The visual data are extracted from the environment without a priori knowledge of the examined scene. Our approach is quite open-ended and can be extended to other flexible systems. Moreover it has been designed to damp the oscillatory behaviour of the arm whatever its origins may be. As a consequence it should yield good performance when vibrations result from a critical trajectory imposed by the operator, from an interaction with the environment, or from internal dynamics of the carried process, e.g. the rotating prism of the IVVS 3D Inspection System. Experimental results validate the proposed strategy.

Vibration control of an IVVS long-reach deployer using unknown visual features from inside the ITER vessel

Energy Technology Data Exchange (ETDEWEB)

Dubus, G., E-mail: gregory.dubus@f4e.europa.e [Fusion for Energy, Remote Handling group, Josep Pla 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); David, O.; Measson, Y. [CEA LIST, Interactive Robotics Unit, 18 route du Panorama, BP6, Fontenay-aux-Roses F-92265 (France)

2010-12-15

The In-Vessel Viewing System (IVVS) project assumes that a long reach deployer equipped with a probe penetrates the ITER chamber to perform periodic inspections. By giving the operator the capability and flexibility to examine unplanned targets, man-in-the-loop technology would be very helpful. But vibrations due to the high flexibility of the structure are probably the main problem in such a master-slave mode, which therefore needs the integration of a high level compensation scheme. However the ITER RH equipment will be confronted with strong electromagnetic interferences as well as a cumulated radiation dose up to several MGy. Short of costly developments, these constraints limit the use of dedicated electronics such as accelerometers or strain gauges. Our main idea is to control the vibrational behaviour of the flexible carrier without considering any extra sensor apart from its embedded probe. In this pre-study we propose to use the kind of rad-hardened viewing system already developed for the AIA demonstrator in order to feed an oscillation observer with visual information. The visual data are extracted from the environment without a priori knowledge of the examined scene. Our approach is quite open-ended and can be extended to other flexible systems. Moreover it has been designed to damp the oscillatory behaviour of the arm whatever its origins may be. As a consequence it should yield good performance when vibrations result from a critical trajectory imposed by the operator, from an interaction with the environment, or from internal dynamics of the carried process, e.g. the rotating prism of the IVVS 3D Inspection System. Experimental results validate the proposed strategy.

Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas

International Nuclear Information System (INIS)

Tamai, H.; Fujita, T.; Kikuchi, M.

2006-01-01

JT-60SA, the former JT-60SC and NCT, a superconducting tokamak positioned as the satellite machine of ITER, collaborating with Japan and EU fusion community, aims at contribution to ITER and DEMO through the demonstration of advanced plasma operation scenario and the plasma applicability test with advanced materials. After the discussions in JA-EU Satellite Tokamak Working Group in 2005, the increased heating power, higher heat removal capacity for the plasma facing components, improvement of the radiation shielding, the remote handling system for the maintenance of in-vessel components, and related equipment are planed to be additionally installed. With such full equipment towards the increased heating power of 41 MW (34 MW-NBI and 7 MW-ECH) with 100 s, the prospective plasma performances, analysed by the equilibrium and transport analysis codes, are rather improved in the view point of the contribution to ITER and DEMO relevant research. Accessibility for higher heating power in a higher density region enables the lower normalized Larmor radius and normalized collision frequency close to the reactor relevant plasma with the ITER-similar configuration of single null divertor plasma with the aspect ratio of A = 3.1, elongation of k95 = 1.7, triangularity of d95 (q95) in the plasma current of I p = 3.5 MA, toroidal magnetic field of B T = 2.59 T and the major radius of Rp=3.16 m. The increases in the electron temperature, beam driven and bootstrap current fraction by the increase of the power of Negative ion based NBI (10 MW) reduce the loop voltage and contribute to increase the maximum plasma current of ITER similar shape. Full non-inductive current drive controllability is also extended into the high density and high plasma current operation towards high beta plasma. Flexibility in aspect ratio and shape parameter is kept the same as NCT, i.e. a double null divertor plasma with A = 2.6, k95 = 1.83, d95 = 0.57, I p = 5.5 MA, B T = 2.72 T, and R p = 3.01 m which

Deuterium in-vessel retention characterisation through the use of particle balance on Tore Supra

International Nuclear Information System (INIS)

Bucalossi, J.; Brosset, C.; Pegourie, B.; Tsitrone, E.; Dufour, E.; Eckedahl, A.; Geraud, A.; Goniche, M.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; Vallet, J.C.; Vartanian, S.

2007-01-01

Fuel retention inside plasma facing components will be a crucial issue not only in fusion reactors of the future, but also in ITER. The estimation of the fraction of the fuel which remains trapped inside the vessel is quite a difficult task. Particle balance analysis provides information for the whole vacuum chamber as a function of time and can be use to monitor the tritium in-vessel retention in real-time. On Tore Supra with a careful choice and position of pressure sensors, proper calibration procedures, the accuracy of the balance is around 10%. Particle balance analysis have been performed on many long pulse discharges and deuterium in-vessel retention has been found to be a constant around 5 x 10 20 D/s after several minutes of plasma. The evolution of the retention rate with plasma parameters indicates that deuterium bulk implantation and diffusion could dominate codeposition with carbon atoms. Particle balance is a powerful tool that should be implemented in ITER

Neutron cameras for ITER

International Nuclear Information System (INIS)

Johnson, L.C.; Barnes, C.W.; Batistoni, P.

1998-01-01

Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from 16 N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with 16 N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins

Retinal biometrics based on Iterative Closest Point algorithm.

Science.gov (United States)

Hatanaka, Yuji; Tajima, Mikiya; Kawasaki, Ryo; Saito, Koko; Ogohara, Kazunori; Muramatsu, Chisako; Sunayama, Wataru; Fujita, Hiroshi

2017-07-01

The pattern of blood vessels in the eye is unique to each person because it rarely changes over time. Therefore, it is well known that retinal blood vessels are useful for biometrics. This paper describes a biometrics method using the Jaccard similarity coefficient (JSC) based on blood vessel regions in retinal image pairs. The retinal image pairs were rough matched by the center of their optic discs. Moreover, the image pairs were aligned using the Iterative Closest Point algorithm based on detailed blood vessel skeletons. For registration, perspective transform was applied to the retinal images. Finally, the pairs were classified as either correct or incorrect using the JSC of the blood vessel region in the image pairs. The proposed method was applied to temporal retinal images, which were obtained in 2009 (695 images) and 2013 (87 images). The 87 images acquired in 2013 were all from persons already examined in 2009. The accuracy of the proposed method reached 100%.

Elastoplastic analysis of surface cracks in pressure vessels using slip-line theory

International Nuclear Information System (INIS)

Keskinen, R.P.

1983-01-01

The paper considers the aspects of engineering application of SLF theory to long surface cracks in pressure vessels. Green's upper-bound SLF for a bend specimen with deep wedge-shaped notch of small flank angle is adopted to analyse the remaining ligament of the cracked section. The SLF involves only one unknown variable, i.e., the radius of a circular slip-line arc, which can be evaluated from the equilibrium condition across the ligament. The stress distribution across the ligament is easily computed by Hencky's theorem and the respective stress resultants produce the boundary conditions for the solution of the neighboring elastic material. The elastic solution readily yields the rotation of the crack edges, COA, and it in turn geometrically defines the applied CTOD. Comparison has proved their relation to the stress resultants identical with that following from the customary single plastic hinge model when Tresca's yield condition prevails and the tensile side plastic constraint factor of the hinge model is chosen as 1.7. The SLF approach is demonstrated for an internal circumferential surface crack subjected to thermal gradient and axial load representative of overpressurization and emergency cooling conditions of a pressure vessel. Analytical formulas relating COA and CTOD to applied loading are derived and CTOD-R curve based stable crack propagation is solved iteratively. Generic numerical results are presented for COA and CTOD under arbitrary loading combination. The risk of crack growth initiation appears to increase with the linear dimensions of the pressure vessel, but remains small for a chosen BWR application. For a long axial surface crack the approach agrees with a previous plastic hinge analysis by Ranta-Maunus et al. suggesting instability under certain combinations of thermal gradient and internal pressure. (orig./HP)

Maintenance schemes for the ITER neutral beam test facility

International Nuclear Information System (INIS)

Zaccaria, P.; Dal Bello, S.; Marcuzzi, D.; Masiello, A.; Coniglio, A.; Antoni, V.; Cordier, J.J.; Hemsworth, R.; Jones, T.; Di Pietro, E.; Mondino, P.L.

2004-01-01

The ITER neutral beam test facility (NBTF) is planned to be built, after the approval of the ITER construction and the choice of the ITER site, with the agreement of the ITER International Team and of the JA and RF participant teams. The key purpose is to progressively increase the performance of the first ITER injector and to demonstrate its reliability at the maximum operation parameters: power delivered to the plasma 16.5 MW, beam energy 1 MeV, accelerated D - ion current 40 A, pulse length 3600 s. Several interventions for possible modifications and for maintenance are expected during the early operation of the ITER injector in order to optimize the beam generation, aiming and steering. The maintenance scheme and the related design solutions are therefore a very important aspect to be considered for the NBTF design. The paper describes consistently the many interrelated aspects of the design, such as the optimisation of the vessel and cryopump geometry, in order to get a better maintenance flexibility, an easier man access and a larger access for diagnostic and monitoring. (authors)

An overview of the ITER project

International Nuclear Information System (INIS)

Holtkamp, N.

2007-01-01

The ITER Project Team now coming together in Cadarache is currently being shaped from the old, preserving the legacy of technical know-how built up in the ITER Joint Central Team since 1992. It is particularly strong initially in the most urgent areas, related to long lead items - magnets, the main vessel and the buildings - as well as in work related to licensing. But it also incorporates new functional needs - financial, administrative, and procurement - and ties in the needs of future users during operation. Since the bulk of the procurement for ITER will be provided in kind, efforts have been strengthened to define better the share of responsibilities with the Parties' Domestic Agencies. The procurement cost sharing is being transferred into realistic technical splitting of the work, with agreements between the Parties to demonstrate production of the necessary quality, and how to handle any shortcomings. The design has evolved since originally conceived and valued 5 years ago. Design reviews of specific procurements will therefore start in September 2006 to ensure the current manufacturing and design assumptions continue to satisfy requirements. This paper reviews the current status of development of the ITER project, covering organisational and technical issues

Plasma scram in ITER L-mode ignited plasmas

International Nuclear Information System (INIS)

Villar Colome, J.; Johner, J.; Ane, J.M.

1995-01-01

The security of ITER will depend on the capability of the system in rapidly extinguishing the 1.5 GW of nominal fusion power without disruption. The local RLW transport model is used to simulate such a Plasma Scram. The conditions for a passively secure operation point in steady-state are discussed in terms of particle exhaust. The time scales of the process should determine the power supplies of both equilibrium coils and central solenoid. (authors). 6 refs., 4 figs., 2 tabs

A numerical solution for a toroidal plasma in equilibrium

International Nuclear Information System (INIS)

Hintz, E.; Sudano, J.P.

1982-01-01

The iterative techniques alternating direction implicit (ADI), sucessive ove-relaxation (SOR) and Gauss-Seidel are applied to a nonlinear elliptical second order differential equation (Grand-Shafranov). This equation was solve with the free boundary conditions plasma-vacuum interface over a rectangular section in cylindrical coordinates R and Z. The current density profile, plasma pressure profile, magnetic and isobaric surfaces are numerically determined for a toroidal plasma in equilibrium. (L.C.) [pt

Erosion evaluation capability of the IVVS for ITER applications

Energy Technology Data Exchange (ETDEWEB)

Pollastrone, Fabio, E-mail: fabio.pollastrone@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Ferri de Collibus, Mario; Florean, Marco; Francucci, Massimo; Mugnaini, Giampiero; Neri, Carlo; Rossi, Paolo [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Dubus, Gregory; Damiani, Carlo [Fusion For Energy c/Josep Pla, 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

2014-10-15

Highlights: •High resolution laser radar range images for hostile environment (IVVS). •Evaluation of the erosion on the surface scanned by IVVS laser radar. •Erosion evaluation procedure and software. •Test and results of the erosion evaluation procedure. -- Abstract: In ITER it is foreseen the use of the In Vessel Viewing System (IVVS), whose scanning head is a 3D laser imaging system able to obtain high-resolution intensity and range images in hostile environments. The IVVS will be permanently installed into a port extension, therefore it has to be compliant with ITER primary vacuum requirements. In the frame of a Fusion for Energy Grant, an investigation of the expected IVVS metrology performances was required to evaluate the device capability to detect erosions on ITER first wall and divertor and to estimate the amount of eroded material. In ENEA Frascati laboratories, an IVVS probe prototype was developed along with a method and a computational procedure applied to a reference erosion plate target simulating ITER vessel components and their possible erosions. Experimental tests were carried out by this system performing several scans of the reference target with different incidence angles, estimating the eroded volume and comparing this volume with its true value. A dedicated study has been also done by changing the power of the laser source; a discussion about the quality of the 3D laser images is reported. The main results obtained during laboratory tests and data processing are presented and discussed.

Progress in the design of the ITER Neutral Beam cell Remote Handling System

Energy Technology Data Exchange (ETDEWEB)

Shuff, R., E-mail: robin.shuff@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Van Uffelen, M.; Damiani, C. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Tesini, A.; Choi, C.-H. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Meek, R. [Oxford Technologies Limited, 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom)

2014-10-15

The ITER Neutral Beam cell will include a suite of Remote Handling equipment for maintenance tasks. This paper summarises the current status and recent developments in the design of the ITER Neutral Beam Remote Handling System. Its concept design was successfully completed in July 2012 by CCFE in the frame of a grant agreement with F4E, in collaboration with the ITER Organisation, including major systems like monorail crane, Beam Line Transporter, beam source equipment, upper port and neutron shield equipment and associated tooling. Research and development activities are now underway on the monorail crane radiation hardened on-board control system and first of a kind remote pipe and lip seal maintenance tooling for the beam line vessel, reported in this paper.

Progress in the design of the ITER Neutral Beam cell Remote Handling System

International Nuclear Information System (INIS)

Shuff, R.; Van Uffelen, M.; Damiani, C.; Tesini, A.; Choi, C.-H.; Meek, R.

2014-01-01

The ITER Neutral Beam cell will include a suite of Remote Handling equipment for maintenance tasks. This paper summarises the current status and recent developments in the design of the ITER Neutral Beam Remote Handling System. Its concept design was successfully completed in July 2012 by CCFE in the frame of a grant agreement with F4E, in collaboration with the ITER Organisation, including major systems like monorail crane, Beam Line Transporter, beam source equipment, upper port and neutron shield equipment and associated tooling. Research and development activities are now underway on the monorail crane radiation hardened on-board control system and first of a kind remote pipe and lip seal maintenance tooling for the beam line vessel, reported in this paper

Manufacturing progress on the first sector and lower ports for ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Ahn, H.J., E-mail: hjahn@nfri.re.kr [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kim, H.S.; Kim, G.H.; Park, C.K.; Hong, G.H.; Jin, S.W.; Lee, H.G.; Jung, K.J. [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Lee, J.S.; Kim, T.S.; Won, J.G.; Roh, B.R.; Park, K.H. [Hyundai Heavy Industries Co. Ltd., Ulsan 682-792 (Korea, Republic of); Sa, J.W.; Choi, C.H.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France)

2016-11-01

Highlights: • All manufacturing drawings of the first sector of VV have been completed. • Full scale mock-ups have been constructed to verify fabrication procedure. • Qualifications for welding and forming are done and for NDE are ongoing. • Manufacturing progress is around 40% for the sector and LPSE up to the end of 2015. - Abstract: Manufacturing design of Korean sectors and ports for the ITER Vacuum Vessel (VV) has been developed to comply with the tight tolerance and severe inspection requirements. The first VV sector and lower ports are being fabricated slowly under strict regulations after verification using several real scale mock-ups and qualifications for welding, forming and NDE. During three years after start of fabrication, manufacturing progress on four poloidal segments of the first sector is that (1) all inner shells were welded, (2) forgings for complicate components have been machined, (3) port stubs and poloidal T-ribs were assembled, and (4) machined components are welded on the inner shells by narrow-gap TIG welding and electron beam welding. The progress of lower ports is that (1) inner shells of stub extensions were bent and treated with heat, (2) T-ribs were fabricated and examined by qualified phased array UT, (3) supporting pads and gussets have been machined, and (4) inner shells are assembled with T-ribs and machined forgings. The progress rate of manufacturing is around 40% up to the end of 2015 for the first sector and lower port stub extensions.

Assistance tools for generic definition of ITER maintenance tasks and scenarios in advanced supervisory control systems

International Nuclear Information System (INIS)

Zieba, Stéphane; Russotto, François-Xavier; Da Silva Simoes, Max; Measson, Yvan

2013-01-01

Highlights: ► Improve supervisory control systems for ITER in-vessel and hot cell maintenance. ► Optimize remote handling operations effectiveness, reliability and safety. ► Provide a generic description of the maintenance tasks and scenarios. ► Development of context-based assistances for operators and supervisor. ► Improvement of operator's situation awareness. -- Abstract: This paper concerns the improvement of supervisory control systems in the context of remote handling for the maintenance tasks in ITER. This work aims at providing a single formalism and tools to define in a generic way the ITER maintenance tasks and scenarios for in-vessel and hot cell operations. A three-layered approach is proposed to model these tasks and scenarios. Physical actions are defined for the scene elements. From these physical actions, behaviours are defined to represent high-level functionalities. Finally, interaction modes define the way that behaviours are achieved in terms of human–machine interactions. Case study concerning the blanket maintenance procedure is discussed concerning the contributions of the descriptive model and the context-based assistances to the activities of supervisory control

Effect of density fluctuations on ECCD in ITER and TCV

Directory of Open Access Journals (Sweden)

Coda S.

2012-09-01

Full Text Available Density fluctuations near the edge of tokamak plasmas can affect the propagation of electron cyclotron (EC waves. In the present paper, the EC wave propagation in a fluctuating equilibrium is determined using the ray-tracing code C3PO. The evolution of the electron distribution function is calculated self-consistently with the EC wave damping using the 3-D Fokker-Planck solver LUKE. The cumulative effect of fluctuations results in a significant broadening of the current profile combined with a fluctuating power deposition profile. This mechanism improves the simulation of fully non-inductive EC discharges in the TCV tokamaks. Predictive simulations for ITER show that density fluctuations could make the stabilization of NTMs in ITER more challenging.

Two-dimensional over-all neutronics analysis of the ITER device

Science.gov (United States)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

Two-dimensional over-all neutronics analysis of the ITER device

International Nuclear Information System (INIS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi.

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR) and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li 2 O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No.5 and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design. (author)

Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 2. Comprehending the divertor structure

International Nuclear Information System (INIS)

Suzuki, Satoshi; Akiba, Masato; Saito, Masakatsu

2006-01-01

Divertor is given the largest heat load in the in-vessel components of fusion machine. The functions and conditions of divertor are stated from the point of view of thermal and structural dynamics. The way of thinking of structure design of divertor of JT-60 and the ITER (International Thermonuclear Experimental Reactor) is explained. As the conditions of divertor, the materials for large heat load, heat removal, pressure boundary, control of damage, and thermal stress/strain are considered. The divertor has to be changed periodically. The materials are required the heat removal function for high heat load. CuCrZr will be used to cooling tube and heat sink, and CFC materials for the surface. The cross section of ITER, a part of divertor, heat load of divertor and other components, the thermal conductivity of CFC and metal materials, conditions of cooling water for divertor of BWR, PWR and ITER, the thermal stress produced on rod, vertical target of ITER, structure of cooling tube, distribution of temperature and critical heart flux of inner wall of cooling tube, and fatigue clack of cooling tube are shown. (S.Y.)

A new estimation method for nuclide number densities in equilibrium cycle

International Nuclear Information System (INIS)

Seino, Takeshi; Sekimoto, Hiroshi; Ando, Yoshihira.

1997-01-01

A new method is proposed for estimating nuclide number densities of LWR equilibrium cycle by multi-recycling calculation. Conventionally, it is necessary to spend a large computation time for attaining the ultimate equilibrium state. Hence, the cycle in nearly constant fuel composition has been considered as an equilibrium state which can be achieved by a few of recycling calculations on a simulated cycle operation under a specific fuel core design. The present method uses steady state fuel nuclide number densities as the initial guess for multi-recycling burnup calculation obtained by a continuously fuel supplied core model. The number densities are modified to be the initial number densities for nuclides of a batch supplied fuel. It was found that the calculated number densities could attain to more precise equilibrium state than that of a conventional multi-recycling calculation with a small number of recyclings. In particular, the present method could give the ultimate equilibrium number densities of the nuclides with the higher mass number than 245 Cm and 244 Pu which were not able to attain to the ultimate equilibrium state within a reasonable number of iterations using a conventional method. (author)

The WEST programme: Minimizing technology and operational risks of a full actively cooled tungsten divertor on ITER

Energy Technology Data Exchange (ETDEWEB)

Grosman, André, E-mail: andre.grosman@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Bucalossi, Jérôme; Doceul, Louis [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, Frédéric [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Lipa, Manfred [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Merola, Mario [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Missirlian, Marc [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Pitts, Richard A. [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Samaille, Franck; Tsitrone, Emmanuelle [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France)

2013-10-15

Highlights: ► The WEST programme is a unique opportunity to experience the industrial scale manufacture of tungsten plasma-facing components similar to the ITER divertor ones. ► In Tore Supra, it will bring important know how for actively cooled W divertor operation. ► This can be done by a reasonable modification of the Tore Supra tokamak. ► A fast implementation of the project would make this information available in due time. ► This allows a significant contribution to the W ITER divertor risk minimization in its manufacturing and operation phase. -- Abstract: The WEST programme consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. This is obtained by inserting in vessel coils to create the X point while adapting the in-vessel elements to this new geometry. This will allow the full tungsten divertor technology to be used on ITER to be tested in anticipation of its use on ITER under relevant heat loading conditions and pulse duration. The early manufacturing of a significant industrial series of ITER-similar W plasma-facing units will contribute to the ITER divertor manufacturing risk mitigation and to that associated with early W divertor plasma operation on ITER.

ITER diagnostic system: Vacuum interface

Energy Technology Data Exchange (ETDEWEB)

Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

2013-10-15

Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

ITER diagnostic system: Vacuum interface

International Nuclear Information System (INIS)

Patel, K.M.; Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L.; Drevon, J.M.; Encheva, A.; Kashchuk, Y.; Maquet, Ph.; Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J.

2013-01-01

Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10 −7 Pa, irrespective of plasma operation, and a leak rate of less than 10 −10 Pa m 3 s −1 . In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions will be described

ITER operating limit definition criteria

International Nuclear Information System (INIS)

Ciattaglia, S.; Barabaschi, P.; Carretero, J.A.; Chiocchio, S.; Hureau, D.; Girard, J.Ph.; Gordon, C.; Portone, A.; Rodrigo, L. Rodriguez; Roldan, C.; Saibene, G.; Uzan-Elbez, J.

2009-01-01

The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction. OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent. This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits.

ITER Organization - 2012 Annual Report, 2012 Financial Statements

International Nuclear Information System (INIS)

2013-01-01

In its first part, this report gives an overview of the main activities and events regarding the ITER organization, the ITER project baseline, the construction of seismic foundations, the licensing decree in France, the procurement arrangements, the manufacturing of the ITER vacuum vessel, the research and development for prototype development, the management of member contributions in France, the creation of new positions as far as staffing is concerned. The second part presents the various highlights for the year and by department: Office of the Director-General, Legal Affairs, International Audit, ITER Council Secretariat, Bureau of International Cooperation, Department for ITER Project (Directorate for Central Integration and Engineering, Directorate for Tokamak, Directorate for CODAC, Heating and Diagnostics, Directorate for Buildings and Site Infrastructure, Directorate for Central Engineering and Plant, Directorate for Plasma Operation), Department for Safety Quality and Security, Department for Administration). The next parts contain tables and charts which present staffing and financial data, presentations of procurement highlights and data for domestic agencies (China, Europe, India, Japan, Korea, Russia, USA) in terms of R and D and manufacturing, of contracts. The last part presents and comments the financial statements for 2012

2-D Reflectometer Modeling for Optimizing the ITER Low-field Side Reflectometer System

International Nuclear Information System (INIS)

Kramer, G.J.; Nazikian, R.; Valeo, E.J.; Budny, R.V.; Kessel, C.; Johnson, D.

2005-01-01

The response of a low-field side reflectometer system for ITER is simulated with a 2?D reflectometer code using a realistic plasma equilibrium. It is found that the reflected beam will often miss its launch point by as much as 40 cm and that a vertical array of receiving antennas is essential in order to observe a reflection on the low-field side of ITER

European Technological Effort in Preparation of ITER Construction

International Nuclear Information System (INIS)

Andreani, Roberto

2005-01-01

Europe has started since the '80s with the preparatory work done on NET, the Next European Torus, the successor of JET, to prepare for the construction of the next generation experiment on the road to the fusion reactor. In 2000 the European Fusion Development Agreement (EFDA) has been signed by sixteen countries, including Switzerland, not a member of the Union. Now the signatory countries have increased to twenty-five. A vigorous programme of design and R and D in support of ITER construction has been conducted by EFDA through the coordinated effort of the national institutes and laboratories supported financially, in the framework of the VI European Framework Research Programme (2002-2006), by contracts of association with EURATOM. In the last three years, with the expenditure of 160 M[Euro], the accent has been particularly put on the preparation of the industrial manufacturing activities of components and systems for ITER. Prototypes and manufacturing methods have been developed in all the main critical areas of machine construction with the objective of providing sound and effective solutions: vacuum vessel, toroidal field coils, poloidal field coils, remote handling equipment, plasma facing components and divertor components, electrical power supplies, generators and power supplies for the Heating and Current Drive Systems and other minor subsystems.Europe feels to be ready to host the ITER site and to provide adequate support and guidance for the success of construction to our partners in the ITER collaboration, wherever needed

Calculation code NIRVANA for free boundary MHD equilibrium

International Nuclear Information System (INIS)

Ninomiya, Hiromasa; Suzuki, Yasuo; Kameari, Akihisa

1975-03-01

The calculation method and code of solving the free boundary problem for MHD equilibrium has been developed. Usage of the code ''NIRVANA'' is described. The toroidal plasma current density determined as a function of the flux function PSI is substituted by a group of the ring currents, whereby the equation of MHD equilibrium is transformed into an integral equation. Either of the two iterative methods is chosen to solve the integral equation, depending on the assumptions made of the plasma surface points. Calculation of the magnetic field configurations is possible when the plasma surface coincides self-consistently with the magnetic flux including the separatrix points. The code is usable in calculation of the circular or non-circular shell-less Tokamak equilibrium. (auth.)

Conceptual design of the hot cell facility universal docking station at ITER

International Nuclear Information System (INIS)

Dammann, A.; Benchikhoune, M.; Friconneau, J.P.; Ivanov, V.; Lemee, A.; Martins, J.P.; Tamassy, G.

2011-01-01

Between main shutdowns of the ITER machine, in-vessel components and Iter Remote Maintenance System (IRMS) are transferred between the Tokamak complex and the Hot Cell Facility using different types of sealed casks. Transfer Casks have different physical interfaces with the Vacuum Vessel, which need to be the same at the docking stations of the HCF. It means that in-vessel components and IRMS are cleaned in the same cells, which is in fact not convenient. Furthermore, logistic studies showed that the use rate of the cells is very inhomogeneous. In order to have dedicated cell for decontamination of Remote Handling tools, in order to increase the operability efficiency and to removes the hot cell docking operation from the critical path, the concept of a universal docking station has been investigated. Based on an existing design, the work was focused on a review of requirements, the re-design and the integration within the HCF layout. The universal docking station has been proposed and is now integrated in HCF design.

Conceptual design of the hot cell facility universal docking station at ITER

Energy Technology Data Exchange (ETDEWEB)

Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Benchikhoune, M.; Friconneau, J.P.; Ivanov, V. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Lemee, A. [SOGETI High Tech, 180 Rue Rene Descartes, 13851 Aix en Provence (France); Martins, J.P. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Tamassy, G. [SOGETI High Tech, 180 Rue Rene Descartes, 13851 Aix en Provence (France)

2011-10-15

Between main shutdowns of the ITER machine, in-vessel components and Iter Remote Maintenance System (IRMS) are transferred between the Tokamak complex and the Hot Cell Facility using different types of sealed casks. Transfer Casks have different physical interfaces with the Vacuum Vessel, which need to be the same at the docking stations of the HCF. It means that in-vessel components and IRMS are cleaned in the same cells, which is in fact not convenient. Furthermore, logistic studies showed that the use rate of the cells is very inhomogeneous. In order to have dedicated cell for decontamination of Remote Handling tools, in order to increase the operability efficiency and to removes the hot cell docking operation from the critical path, the concept of a universal docking station has been investigated. Based on an existing design, the work was focused on a review of requirements, the re-design and the integration within the HCF layout. The universal docking station has been proposed and is now integrated in HCF design.

Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

Directory of Open Access Journals (Sweden)

G. De Temmerman

2017-08-01

Full Text Available As a licensed nuclear facility, ITER must limit the in-vessel tritium (T retention to reduce the risks of potential release during accidents, the inventory limit being set at 1kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513K for the FW and 623K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

International Nuclear Information System (INIS)

Gargiulo, L.; Cordier, J.J.; Friconneau, J.P.; Grisolia, C.; Palmer, J.D.; Perrot, Y.; Samaille, F.

2007-01-01

The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools

Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

Energy Technology Data Exchange (ETDEWEB)

Gargiulo, L. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)], E-mail: laurent.gargiulo@cea.fr; Cordier, J.J. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Friconneau, J.P. [CEA-LIST Robotics and Interactive Systems Unit, BP6 F-92265 Fontenay aux Roses Cedex (France); Grisolia, C. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Palmer, J.D. [EFDA CSU, Max-Planck-Institut fuer Plasma Physik Boltzmannstr. 2, D-85748 Garching (Germany); Perrot, Y. [CEA-LIST Robotics and Interactive Systems Unit, BP6 F-92265 Fontenay aux Roses Cedex (France); Samaille, F. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)

2007-10-15

The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools.

Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

Science.gov (United States)

Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

2017-12-01

The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

Test facility TIMO for testing the ITER model cryopump

International Nuclear Information System (INIS)

Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

2001-01-01

Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L' Air Liquide in the form of a NET contract. (author)

Test facility TIMO for testing the ITER model cryopump

International Nuclear Information System (INIS)

Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

1999-01-01

Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L'Air Liquide in the form of a NET contract. (author)

Development of thick wall welding and cutting tools for ITER

International Nuclear Information System (INIS)

Nakahira, Masataka; Takahashi, Hiroyuki; Akou, Kentaro; Koizumi, Koichi

1998-01-01

The Vacuum Vessel, which is a core component of International Thermonuclear Experimental Reactor (ITER), is required to be exchanged remotely in a case of accident such as superconducting coil failure. The in-vessel components such as blanket and divertor are planned to be exchanged or fixed. In these exchange or maintenance operations, the thick wall welding and cutting are inevitable and remote handling tools are necessary. The thick wall welding and cutting tools for blanket are under developing in the ITER R and D program. The design requirement is to weld or cut the stainless steel of 70 mm thickness in the narrow space. Tungsten inert gas (TIG) arc welding, plasma cutting and iodine laser welding/cutting are selected as primary option. Element welding and cutting tests, design of small tools to satisfy space requirement, test fabrication and performance tests were performed. This paper reports the tool design and overview of welding and cutting tests. (author)

Development of thick wall welding and cutting tools for ITER

Energy Technology Data Exchange (ETDEWEB)

Nakahira, Masataka; Takahashi, Hiroyuki; Akou, Kentaro; Koizumi, Koichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-04-01

The Vacuum Vessel, which is a core component of International Thermonuclear Experimental Reactor (ITER), is required to be exchanged remotely in a case of accident such as superconducting coil failure. The in-vessel components such as blanket and divertor are planned to be exchanged or fixed. In these exchange or maintenance operations, the thick wall welding and cutting are inevitable and remote handling tools are necessary. The thick wall welding and cutting tools for blanket are under developing in the ITER R and D program. The design requirement is to weld or cut the stainless steel of 70 mm thickness in the narrow space. Tungsten inert gas (TIG) arc welding, plasma cutting and iodine laser welding/cutting are selected as primary option. Element welding and cutting tests, design of small tools to satisfy space requirement, test fabrication and performance tests were performed. This paper reports the tool design and overview of welding and cutting tests. (author)

Accuracy analysis of hybrid parallel robot for the assembling of ITER

Energy Technology Data Exchange (ETDEWEB)

Wang Yongbo [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland); The State Key Laboratory of Mechanical Transmission, Chongqing University (China); Pessi, Pekka [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland); Wu Huapeng [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland)], E-mail: huapeng@lut.fi; Handroos, Heikki [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland)

2009-06-15

This paper presents a novel mobile parallel robot, which is able to carry welding and machining processes from inside the international thermonuclear experimental reactor (ITER) vacuum vessel (VV). The kinematics design of the robot has been optimized for ITER access. To improve the accuracy of the parallel robot, the errors caused by the stiffness and manufacture process have to be compensated or limited to a minimum value. In this paper kinematics errors and stiffness modeling are given. The simulation results are presented.

Accuracy analysis of hybrid parallel robot for the assembling of ITER

International Nuclear Information System (INIS)

Wang Yongbo; Pessi, Pekka; Wu Huapeng; Handroos, Heikki

2009-01-01

This paper presents a novel mobile parallel robot, which is able to carry welding and machining processes from inside the international thermonuclear experimental reactor (ITER) vacuum vessel (VV). The kinematics design of the robot has been optimized for ITER access. To improve the accuracy of the parallel robot, the errors caused by the stiffness and manufacture process have to be compensated or limited to a minimum value. In this paper kinematics errors and stiffness modeling are given. The simulation results are presented.

Project status of manufacturing of European toroidal coils ITER. Validation tests

International Nuclear Information System (INIS)

Pando, F.; Felipe, A.; Madorran, A.; Pallisa, J.; Dormicch, O.; Valle, N.; D'Urzo, C.; Marin, M.; Pesenti, P.; Lucas, J.; Moreno, N.; Bonito-Oliva, A.; Harrison, R.; Bellesia, B.; Cornelis, M.; Cornella, J.

2015-01-01

The toroidal field coils are the ITER magnets responsible for confining the plasma inside the vacuum vessel. The consortium formed by IBERDROLA Ingenieria y Construccion, ASG Superconductors y ELYTT Energy is the responsible for the supply of 10 coils that the european agency F4E has to supply for the ITER project. At present, the coils are been manufactured in La Spezia (Italy), after the qualification of all the manufacturing process and the sucessfull manufacturing of a full scale prototype. (Author)

Project status of manufacturing of European toroidal coils ITER. Validation tests; Estado del proyecto de fabricacion de las bobinas toroidales european para el ITER. Ensayos de validacion

Energy Technology Data Exchange (ETDEWEB)

Pando, F.; Felipe, A.; Madorran, A.; Pallisa, J.; Dormicch, O.; Valle, N.; D' Urzo, C.; Marin, M.; Pesenti, P.; Lucas, J.; Moreno, N.; Bonito-Oliva, A.; Harrison, R.; Bellesia, B.; Cornelis, M.; Cornella, J.

2015-07-01

The toroidal field coils are the ITER magnets responsible for confining the plasma inside the vacuum vessel. The consortium formed by IBERDROLA Ingenieria y Construccion, ASG Superconductors y ELYTT Energy is the responsible for the supply of 10 coils that the european agency F4E has to supply for the ITER project. At present, the coils are been manufactured in La Spezia (Italy), after the qualification of all the manufacturing process and the sucessfull manufacturing of a full scale prototype. (Author)

Application of morphological bit planes in retinal blood vessel extraction.

Science.gov (United States)

Fraz, M M; Basit, A; Barman, S A

2013-04-01

The appearance of the retinal blood vessels is an important diagnostic indicator of various clinical disorders of the eye and the body. Retinal blood vessels have been shown to provide evidence in terms of change in diameter, branching angles, or tortuosity, as a result of ophthalmic disease. This paper reports the development for an automated method for segmentation of blood vessels in retinal images. A unique combination of methods for retinal blood vessel skeleton detection and multidirectional morphological bit plane slicing is presented to extract the blood vessels from the color retinal images. The skeleton of main vessels is extracted by the application of directional differential operators and then evaluation of combination of derivative signs and average derivative values. Mathematical morphology has been materialized as a proficient technique for quantifying the retinal vasculature in ocular fundus images. A multidirectional top-hat operator with rotating structuring elements is used to emphasize the vessels in a particular direction, and information is extracted using bit plane slicing. An iterative region growing method is applied to integrate the main skeleton and the images resulting from bit plane slicing of vessel direction-dependent morphological filters. The approach is tested on two publicly available databases DRIVE and STARE. Average accuracy achieved by the proposed method is 0.9423 for both the databases with significant values of sensitivity and specificity also; the algorithm outperforms the second human observer in terms of precision of segmented vessel tree.

NET in-vessel vehicle system

International Nuclear Information System (INIS)

Jones, H.

1991-02-01

The CFFTP/Spar In-vessel Vehicle System concept for in-vessel remote maintenance of the NET/ITER machine is described. It comprises a curved deployable boom, a vehicle which can travel on the boom and an end effector or work unit mounted on the vehicle. The stowed boom, vehicle, and work unit are inserted via the equatorial access port of the torus. Following insertion the boom is deployed and locked in place. The vehicle may then travel along the boom to transport the work unit to any desired location. A novel feature of the concept is the deployable boom. When fully deployed, it closely resembles a conventional curved truss structure in configuration and characteristics. However, the joints of the truss structure are hinged so that it can fold into a compact package, of less than 20% of deployed volume for storage, transportation and insertion into the torus. A full-scale 2-metre long section of this boom was produced for demonstration purposes. As part of the concept definition the work unit for divertor handling was studied to demonstrate that large payloads could be manipulated within the confines of the torus using the in-vessel vehicle system. Principal advantages of the IVVS are its high load capacity and rigidity, low weight and stowed volume, simplicity of control and operation, and its relatively high speed of transportation

ITER EDA newsletter. V. 3, no. 1. (International Thermonuclear Experimental Reactor Engineering Design Activities)

International Nuclear Information System (INIS)

1994-01-01

This issues of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the Fourth ITER Management Advisory Committee Meeting (MAC) held at San Diego, USA, 13-14 January, 1994, a Technical Committee Meeting on Plasma Equilibrium and Control held at Naka, Japan, 9-12 November 1993, and a Technical Committee Meeting on Radio-Frequency Heating and Current Drive held in Garching, Germany, 21-26 October 1993

Results on the ITER Technology R and D

International Nuclear Information System (INIS)

1999-01-01

The ITER Engineering Design Activities (EDA) have passed their originally planned six years by approval of the ITER Final Design Report at a meeting of the ITER Council held in July, 1998. The four Parties (EU, Japan, Russia, and USA) had hoped to make a decision for its construction by end of the EDA. However, the financial environment of these Parties were not optimistic to directly start construction of the device scooped in the Report. The ITER Technology R and D has been conducted by cooperation of these four Parties to provide data base and demonstrate technical feasibility on the ITER design. It contains, not only component technologies on tokamak reactor core, but also peripheral system technologies such as heating and current drive technique, remote maintenance technique, tritium technology, fuel air-in-taking/-exhausting technique, measurement diagnosis element technique, safety, and so on. Above all, seven large R and D projects are identified to demonstrate technical feasibility of manufacturing and system tests. They were planned to have scales capable of extrapolating to the ITER and of carrying out by joint efforts of a plural Parties. These projects were relating to superconducting magnet technology; vacuum vessel technology, blanket technology, divertor technology, and remote maintenance technology, among which three projects were promoted under leading of Japan. This report was prepared so as to enable to understand outline of results obtained under the seven projects on the ITER Technology R and D. (G.K.)

Main maintenance operations for Test Blanket Systems in ITER TBM port cells

Energy Technology Data Exchange (ETDEWEB)

Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Friconneau, J.-P.; Giancarli, L.M.; Gotewal, K.K.; Iseli, M.; Kim, B.Y.; Levesy, B.; Martins, J.-P.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Siarras, A. [Sogetti, Parc de la Duranne, 13857 Aix-en-Provence (France); Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues.

Progress of the ECRH Upper Launcher design for ITER

International Nuclear Information System (INIS)

Strauss, D.; Aiello, G.; Bruschi, A.; Chavan, R.; Farina, D.; Figini, L.; Gagliardi, M.; Garcia, V.; Goodman, T.P.; Grossetti, G.; Heemskerk, C.; Henderson, M.A.; Kasparek, W.; Krause, A.; Landis, J.-D.; Meier, A.; Moro, A.; Platania, P.; Plaum, B.; Poli, E.

2014-01-01

The design of the ITER ECRH system provides 20 MW millimeter wave power for central plasma heating and MHD stabilization. The system consists of an array of 24 gyrotrons with power supplies coupled to a set of transmission lines guiding the beams to the four upper and the equatorial launcher. The front steering upper launcher design described herein has passed successfully the preliminary design review, and it is presently in the final design stage. The launcher consists of a millimeter wave system and steering mechanism with neutron shielding integrated into an upper port plug with the plasma facing blanket shield module (in-vessel) and a set of ex-vessel waveguides connecting the launcher to the transmission lines. Part of the transmission lines are the ultra-low loss CVD torus diamond windows and a shutter valve, a miter bend section and the feedthroughs integrated in the plug closure plate. These components are connected by corrugated waveguides and form together the first confinement system (FCS). In-vessel, the millimeter-wave system includes a quasi-optical beam propagation system including four mirror sets and a front steering mirror. The millimeter wave system is integrated into a specifically optimized upper port plug providing structural stability to withstand plasma disruption forces and the high heat load from the plasma side with a dedicated blanket shield module. A recent update in the ITER interface definition has resulted in the recession of the upper port plug first wall panels, which is now integrated into the design. Apart from the millimeter wave system the upper port plug houses also a set of shield blocks which provide neutron shielding. An overview of the actual ITER ECRH Upper Launcher is given together with some highlights of the design

Behaviour of the ASDEX pressure gauge at high neutral gas pressure and applications for ITER

International Nuclear Information System (INIS)

Scarabosio, A.; Haas, G.

2008-01-01

The ASDEX Pressure Gauge is, at present, the main candidate for in-vessel neutral pressure measurement in ITER. Although the APG output is found to saturate at around 15 Pa, below the ITER requirement of 20 Pa. We show, here, that with small modifications of the gauge geometry and potentials settings we can achieve satisfactory behaviour up to 30 Pa at 6 T

Overview of magnetic control in ITER

Energy Technology Data Exchange (ETDEWEB)

Zabeo, L., E-mail: luca.zabeo@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Ambrosino, G., E-mail: ambrosin@unina.it [CREATE/Universitá di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’informazione, Naples (Italy); Cavinato, M., E-mail: mario.cavinato@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla 2, Torres Diagonal Litoral - B3, 08019 Barcelona (Spain); Gribov, Y., E-mail: yuri.gribov@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Kavin, A., E-mail: kavina@sintez.niiefa.spb.su [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Lukash, V., E-mail: lukash@nfi.kiae.ru [Kurchatov Institute, Moscow (Russian Federation); Mattei, M., E-mail: massimiliano.mattei@unina2.it [CREATE/Seconda Universitá di Napoli, Dip. Ingegneria Industriale e dell’informazione, Naples (Italy); Pironti, A., E-mail: pironti@unina.it [CREATE/Seconda Universitá di Napoli, Dip. Ingegneria Industriale e dell’informazione, Naples (Italy); Snipes, J.A., E-mail: joseph.snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Vayakis, G., E-mail: george.vayakis@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Winter, A., E-mail: axel.winter@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France)

2014-05-15

ITER is targeting Q = 10 with 500 MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination. Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios. Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design. Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the

Loads due to stray microwave radiation in ITER

Energy Technology Data Exchange (ETDEWEB)

Oosterbeek, Johan W. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Eindhoven University of Technology, P.O. Box 513, 5600 AZ Eindhoven (Netherlands); Udintsev, Victor S.; Gandini, Franco [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Hirsch, Matthias; Laqua, Heinrich P. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, D-17489 Greifswald (Germany); Maassen, Nick [Eindhoven University of Technology, P.O. Box 513, 5600 AZ Eindhoven (Netherlands); Ma, Yunxing; Polevoi, Alexei; Sirinelli, Antoine; Vayakis, George; Walsh, Mike J. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2015-10-15

High-power microwaves generated by gyrotrons will be extensively used in ITER for a variety of purposes such as assisting plasma breakdown, plasma heating, current drive, tearing mode suppression and as a probing beam for the Collective Thomson Scattering diagnostic. In a number of these schemes absorption of the microwaves by the plasma will not be full and in some cases there could be no absorption at all. This may result in a directed beam with a high microwave power flux or – depending on location and plasma conditions – an approximately isotropic microwave power field. The contribution of electron cyclotron emission to these power densities is briefly discussed. Exposure to in-vessel components leads to absorption by metals and ceramics. In this paper microwave power densities are estimated and, following a brief review of absorption, thermal loads on in-vessel components are assessed. The paper is concluded by a discussion of the current approach to control such loads.

Intelligent controller of a flexible hybrid robot machine for ITER assembly and maintenance

International Nuclear Information System (INIS)

Al-saedi, Mazin I.; Wu, Huapeng; Handroos, Heikki

2014-01-01

Highlights: • Studying flexible multibody dynamic of hybrid parallel robot. • Investigating fuzzy-PD controller to control a hybrid flexible hydraulically driven robot. • Investigating ANFIS-PD controller to control a hybrid flexible robot. Compare to traditional PID this method gives better performance. • Using the equilibrium of reaction forces between the parallel and serial parts of hybrid robot to control the serial part hydraulically driven. - Abstract: The assembly and maintenance of International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is highly challenging since the tasks performed by the robot involve welding, material handling, and machine cutting from inside the VV. To fulfill the tasks in ITER application, this paper presents a hybrid redundant manipulator with four DOFs provided by serial kinematic axes and six DOFs by parallel mechanism. Thus, in machining, to achieve greater end-effector trajectory tracking accuracy for surface quality, a robust control of the actuators for the flexible link has to be deduced. In this paper, the intelligent control of a hydraulically driven parallel robot part based on the dynamic model and two control schemes have been investigated: (1) fuzzy-PID self tuning controller composed of the conventional PID control and with fuzzy logic; (2) adaptive neuro-fuzzy inference system-PID (ANFIS-PID) self tuning of the gains of the PID controller, which are implemented independently to control each hydraulic cylinder of the parallel robot based on rod position predictions. The obtained results of the fuzzy-PID and ANFIS-PID self tuning controller can reduce more tracking errors than the conventional PID controller. Subsequently, the serial component of the hybrid robot can be analyzed using the equilibrium of reaction forces at the universal joint connections of the hexa-element. To achieve precise positional control of the end effector for maximum precision machining, the hydraulic cylinder should

Intelligent controller of a flexible hybrid robot machine for ITER assembly and maintenance

Energy Technology Data Exchange (ETDEWEB)

Al-saedi, Mazin I., E-mail: mazin.al-saedi@lut.fi; Wu, Huapeng; Handroos, Heikki

2014-10-15

Highlights: • Studying flexible multibody dynamic of hybrid parallel robot. • Investigating fuzzy-PD controller to control a hybrid flexible hydraulically driven robot. • Investigating ANFIS-PD controller to control a hybrid flexible robot. Compare to traditional PID this method gives better performance. • Using the equilibrium of reaction forces between the parallel and serial parts of hybrid robot to control the serial part hydraulically driven. - Abstract: The assembly and maintenance of International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is highly challenging since the tasks performed by the robot involve welding, material handling, and machine cutting from inside the VV. To fulfill the tasks in ITER application, this paper presents a hybrid redundant manipulator with four DOFs provided by serial kinematic axes and six DOFs by parallel mechanism. Thus, in machining, to achieve greater end-effector trajectory tracking accuracy for surface quality, a robust control of the actuators for the flexible link has to be deduced. In this paper, the intelligent control of a hydraulically driven parallel robot part based on the dynamic model and two control schemes have been investigated: (1) fuzzy-PID self tuning controller composed of the conventional PID control and with fuzzy logic; (2) adaptive neuro-fuzzy inference system-PID (ANFIS-PID) self tuning of the gains of the PID controller, which are implemented independently to control each hydraulic cylinder of the parallel robot based on rod position predictions. The obtained results of the fuzzy-PID and ANFIS-PID self tuning controller can reduce more tracking errors than the conventional PID controller. Subsequently, the serial component of the hybrid robot can be analyzed using the equilibrium of reaction forces at the universal joint connections of the hexa-element. To achieve precise positional control of the end effector for maximum precision machining, the hydraulic cylinder should

IVVS actuating system compatibility test to ITER gamma radiation conditions

Energy Technology Data Exchange (ETDEWEB)

Rossi, Paolo, E-mail: paolo.rossi@enea.it [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F. [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Baccaro, S.; Piegari, A. [ENEA CR Casaccia, 301 Via Anguillarese, 00123 Santa Maria di Galeria, Rome (Italy); Damiani, C.; Dubus, G. [Fusion For Energy c/Josep Pla, n° 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

2013-10-15

Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

IVVS actuating system compatibility test to ITER gamma radiation conditions

International Nuclear Information System (INIS)

Rossi, Paolo; Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F.; Baccaro, S.; Piegari, A.; Damiani, C.; Dubus, G.

2013-01-01

Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

Remote handling demonstration of ITER blanket module replacement

International Nuclear Information System (INIS)

Kakudate, S.; Nakahira, M.; Oka, K.; Taguchi, K.; Obara, K.; Tada, E.; Shibanuma, K.; Tesini, A.; Haange, R.; Maisonnier, D.

2001-01-01

In ITER, the in-vessel components such as blanket are to be maintained or replaced remotely since they will be activated by 14 MeV neutrons, and a complete exchange of shielding blanket with breeding blanket is foreseen after the Basic Performance Phase. The blanket is segmented into about seven hundred modules to facilitate remote maintainability and allow individual module replacement. For this, the remote handing equipment for blanket maintenance is required to handle a module with a dead weight of about 4 tonne within a positioning accuracy of a few mm under intense gamma radiation. According to the ITER R and D program, a rail-mounted vehicle manipulator system was developed and the basic feasibility of this system was verified through prototype testing. Following this, development of full-scale remote handling equipment has been conducted as one of the ITER Seven R and D Projects aiming at a remote handling demonstration of the ITER blanket. As a result, the Blanket Test Platform (BTP) composed of the full-scale remote handling equipment has been completed and the first integrated performance test in March 1998 has shown that the fabricate remote handling equipment satisfies the main requirements of ITER blanket maintenance. (author)

Availability analysis of the ITER blanket remote handling system

International Nuclear Information System (INIS)

Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

2015-01-01

The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

Status of Design and Manufacturing of ITER 1st batch Assembly Tools

Energy Technology Data Exchange (ETDEWEB)

Bae, Jin Ho; Nam, Kyoun Go; Chung, Si Kun; Ha, Min Su [ITER Korea National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Geun Hong [ITER Organization, St Paul lez Durance (France)

2016-05-15

The ITER tokamak assembly tools are purpose-built and specially designed to complete the ITER tokamak machine which includes; Vacuum Vessel (VV), VV Thermal Shield (VVTS), Toroidal Field Coil (TFC) and other components contained in the cryostat. KODA has carried out the preliminary and final design of these assembly tools. This paper shows that the current status, first quarter of the 2016, including manufacturing of ITER 1st batch assembly tools and briefly summarized the design process through design work of Sector Sub-assembly Tool (SSAT) that is most important tool representing ITER 1st batch assembly tools. KODA (Korea Domestic Agency) should provide 128 kinds of the purpose-built assembly tools for ITER Tokamak machine, and the ITER 1st batch assembly tools are split into 3 groups. The FDR for Group A was performed in December 2014, and design of SSAT has been verified by FE analysis and engineering calculation using EN cords. The SSAT is now under manufacturing phase to meet the ITER milestone. After factory acceptance test of SSAT on end of 2016, the 1st SSAT will be delivered and arrived in ITER site on second quarter of the 2017.

Modeling of Eddy current distribution and equilibrium reconstruction in the SST-1 Tokamak

International Nuclear Information System (INIS)

Banerjee, Santanu; Sharma, Deepti; Radhakrishnana, Srinivasan; Daniel, Raju; Shankara Joisa, Y.; Atrey, Parveen Kumar; Pathak, Surya Kumar; Singh, Amit Kumar

2015-01-01

Toroidal continuity of the vacuum vessel and the cryostat leads to the generation of large eddy currents in these passive structures during the Ohmic phase of the steady state superconducting tokamak SST-1. This reduces the magnitude of the loop voltage seen by the plasma as also delays its buildup. During the ramping down of the Ohmic transformer current (OT), the resultant eddy currents flowing in the passive conductors play a crucial role in governing the plasma equilibrium. Amount of this eddy current and its distribution has to be accurately determined such that this can be fed to the equilibrium reconstruction code as an input. For the accurate inclusion of the effect of eddy currents in the reconstruction, the toroidally continuous conducting structures like the vacuum vessel and the cryostat with large poloidal cross-section and any other poloidal field (PF) coil sitting idle on the machine are broken up into a large number of co-axial toroidal current carrying filaments. The inductance matrix for this large set of toroidal current carrying conductors is calculated using the standard Green's function and the induced currents are evaluated for the OT waveform of each plasma discharge. Consistency of this filament model is cross-checked with the 11 in-vessel and 12 out-vessel toroidal flux loop signals in SST-1. Resistances of the filaments are adjusted to reproduce the experimental measurements of these flux loops in pure OT shots and shots with OT and vertical field (BV). Such shots are taken routinely in SST-1 without the fill gas to cross-check the consistency of the filament model. A Grad-Shafranov (GS) equation solver, named as IPREQ, has been developed in IPR to reconstruct the plasma equilibrium through searching for the best-fit current density profile. Ohmic transformer current (OT), vertical field coil current (BV), currents in the passive filaments along with the plasma pressure (p) and current (I p ) profiles are used as inputs to the IPREQ

Assessment of radiation maps during activated divertor moving in the ITER building

International Nuclear Information System (INIS)

Ying Dongchuan; Zeng Qin; Qiu Yuefeng; Dang Tongqiang; Wu Yican; Loughlin, Michael

2011-01-01

As the main interface components between plasma and vacuum vessel, the divertor is foreseen to be removed to the hot cell for refurbishment during the 20 years of ITER operation. During this process, the activated divertor will cause a large increase of radiation in the ITER building. 3D analysis has been performed to assess the radiation maps throughout the ITER building for assisting the shielding design for personnel and sensitive equipment. The activation of the divertor has been determined by coupled neutron transport and inventory calculations, radiation maps have been obtained from gamma transport calculations. The neutron and gamma transport calculations have been performed by MCNP5 code with FENDL2.1library. The inventory calculations have been performed by FISPACT2007 code with EAF-2007 library. The results of these 3D decay gamma radiation maps are presented by pictures in this paper, including the biological dose maps and maps of heat deposition in electronic equipment.

Magnetic and electrical properties of ITER vacuum vessel steels

International Nuclear Information System (INIS)

Mergia, K.; Apostolopoulos, G.; Gjoka, M.; Niarchos, D.

2007-01-01

Full text of publication follows: Ferritic steel AISI 430 is a candidate material for the lTER vacuum vessel which will be used to limit the ripple in the toroidal magnetic field. The magnetic and electrical properties and their temperature dependence in the temperature range 300 - 900 K of AISI 430 ferritic stainless steels are presented. The temperature variation of the coercive field, remanence and saturation magnetization as well as electrical resistivity and the effect of annealing on these properties is discussed. (authors)

Real time equilibrium reconstruction algorithm in EAST tokamak

International Nuclear Information System (INIS)

Wang Huazhong; Luo Jiarong; Huang Qinchao

2004-01-01

The EAST (HT-7U) superconducting tokamak is a national project of China on fusion research, with a capability of long-pulse (∼1000 s) operation. In order to realize a long-duration steady-state operation of EAST, some significant capability of real-time control is required. It would be very crucial to obtain the current profile parameters and the plasma shapes in real time by a flexible control system. As those discharge parameters cannot be directly measured, so a current profile consistent with the magnetohydrodynamic equilibrium should be evaluated from external magnetic measurements, based on a linearized iterative least square method, which can meet the requirements of the measurements. The arithmetic that the EFIT (equilibrium fitting code) is used for reference will be given in this paper and the computational efforts are reduced by parameterizing the current profile linearly in terms of a number of physical parameters. In order to introduce this reconstruction algorithm clearly, the main hardware design will be listed also. (authors)

ITER diagnostics: Maintenance and commissioning in the hot cell test bed

International Nuclear Information System (INIS)

Walker, C.I.; Barnsley, R.; Costley, A.E.; Gottfried, R.; Haist, B.; Itami, K.; Kondoh, T.; Loesser, G.D.; Palmer, J.; Sugie, T.; Tesini, A.; Vayakis, G.

2005-01-01

In-vessel diagnostic equipment in ITER integrated in six equatorial and 12 upper ports, 16 divertor cassettes and five lower ports is designed to be removed in modules and then repaired, tested and commissioned in the same location at the ITER hot cell. The repair requirements and tests on these components are described along with design features that facilitate repair. The testing establishes the repair strategy, qualifies the refurbishment work and finally checks the mechanical and diagnostic function before the return of the modules. At the hot cell, a dummy port is provided for tests of mechanical and vacuum integrity as well as commissioning of the diagnostic equipment. The scope of the hot cell maintenance and commissioning activities is summarised and an overview of the integration of the diagnostic equipment is given

Rotation and neoclassical ripple transport in ITER

Science.gov (United States)

Paul, E. J.; Landreman, M.; Poli, F. M.; Spong, D. A.; Smith, H. M.; Dorland, W.

2017-11-01

Neoclassical transport in the presence of non-axisymmetric magnetic fields causes a toroidal torque known as neoclassical toroidal viscosity (NTV). The toroidal symmetry of ITER will be broken by the finite number of toroidal field coils and by test blanket modules (TBMs). The addition of ferritic inserts (FIs) will decrease the magnitude of the toroidal field ripple. 3D magnetic equilibria in the presence of toroidal field ripple and ferromagnetic structures are calculated for an ITER steady-state scenario using the variational moments equilibrium code (VMEC). Neoclassical transport quantities in the presence of these error fields are calculated using the stellarator Fokker-Planck iterative neoclassical conservative solver (SFINCS). These calculations fully account for E r , flux surface shaping, multiple species, magnitude of ripple, and collisionality rather than applying approximate analytic NTV formulae. As NTV is a complicated nonlinear function of E r , we study its behavior over a plausible range of E r . We estimate the toroidal flow, and hence E r , using a semi-analytic turbulent intrinsic rotation model and NUBEAM calculations of neutral beam torque. The NTV from the \\vert{n}\\vert = 18 ripple dominates that from lower n perturbations of the TBMs. With the inclusion of FIs, the magnitude of NTV torque is reduced by about 75% near the edge. We present comparisons of several models of tangential magnetic drifts, finding appreciable differences only for superbanana-plateau transport at small E r . We find the scaling of calculated NTV torque with ripple magnitude to indicate that ripple-trapping may be a significant mechanism for NTV in ITER. The computed NTV torque without ferritic components is comparable in magnitude to the NBI and intrinsic turbulent torques and will likely damp rotation, but the NTV torque is significantly reduced by the planned ferritic inserts.

Progress of IRSN R&D on ITER Safety Assessment

Science.gov (United States)

Van Dorsselaere, J. P.; Perrault, D.; Barrachin, M.; Bentaib, A.; Gensdarmes, F.; Haeck, W.; Pouvreau, S.; Salat, E.; Seropian, C.; Vendel, J.

2012-08-01

The French "Institut de Radioprotection et de Sûreté Nucléaire" (IRSN), in support to the French "Autorité de Sûreté Nucléaire", is analysing the safety of ITER fusion installation on the basis of the ITER operator's safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.

Multistep Hybrid Iterations for Systems of Generalized Equilibria with Constraints of Several Problems

Directory of Open Access Journals (Sweden)

Lu-Chuan Ceng

2014-01-01

Full Text Available We first introduce and analyze one multistep iterative algorithm by hybrid shrinking projection method for finding a solution of the system of generalized equilibria with constraints of several problems: the generalized mixed equilibrium problem, finitely many variational inclusions, the minimization problem for a convex and continuously Fréchet differentiable functional, and the fixed-point problem of an asymptotically strict pseudocontractive mapping in the intermediate sense in a real Hilbert space. We prove strong convergence theorem for the iterative algorithm under suitable conditions. On the other hand, we also propose another multistep iterative algorithm involving no shrinking projection method and derive its weak convergence under mild assumptions.

Fabrication of a full-size mock-up for inboard 10o section of ITER vacuum vessel thermal shield

International Nuclear Information System (INIS)

Chung, W.; Nam, K.; Noh, C.H.; Kang, D.K.; Kang, S.M.; Oh, Y.G.; Choi, S.W.; Kang, S.H.; Utin, Y.; Ioki, K.; Her, N.; Yu, J.

2011-01-01

A full-scale mock-up of VVTS inboard section was made in order to validate its manufacturing processes before manufacturing the vacuum vessel thermal shield (VVTS) for ITER tokamak. VVTS inboard 10 o section consists of 20 mm shells on which cooling tubes are welded and flange joints that connect adjacent thermal shield sectors. The whole VVTS inboard is divided into two by bisectional flange joint located at the center. All the manufacturing processes except silver coating were tested and verified in the fabrication of mock-up. For the forming and the welding, pre-qualification tests were conducted to find proper process conditions. Shell thickness change was measured after bending, forming and buffing processes. Shell distortion was adjusted after the welding. Welding was validated by non-destructive examination. Bisectional flange joint was successfully assembled by inserting pins and tightening with bolt/nut. Bolt hole margin of 2 mm for sector flange was revealed to be sufficient by successful sector assembly of upper and lower parts of mock-up. Handling jig was found to be essential because the inboard section was flexible. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner.

TEA: A CODE CALCULATING THERMOCHEMICAL EQUILIBRIUM ABUNDANCES

Energy Technology Data Exchange (ETDEWEB)

Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver, E-mail: jasmina@physics.ucf.edu [Planetary Sciences Group, Department of Physics, University of Central Florida, Orlando, FL 32816-2385 (United States)

2016-07-01

We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature–pressure pairs. We tested the code against the method of Burrows and Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows and Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.

TEA: A CODE CALCULATING THERMOCHEMICAL EQUILIBRIUM ABUNDANCES

International Nuclear Information System (INIS)

Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver

2016-01-01

We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature–pressure pairs. We tested the code against the method of Burrows and Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows and Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.

Numerical simulation of plasma vertical position stabilization in ITER

International Nuclear Information System (INIS)

Astapkovich, A.M.; Sadakov, S.N.

1992-01-01

The paper deals with numerical simulation of plasma vertical position stabilization in ITER. The calculations are performed using EDDY C-2 code by the method of direct numerical simulation of transient electromagnetic processes taking into account the evolution of plasma position, cross-section shape and full plasma current. When simulating free vertical plasma drift in ITER with twin passive stabilization loops, it was shown that account of the effects of cross-section deformation and plasma current alternations results in almost two fold degradation of passive stabilization parameters as compared to the calculations for 'rigid displacement' model. In terms of methodology, the account of the effects of cross section deformation and plasma current alternations requires clarification of the definitions for reverse increment of vertical instability and for stability margin coefficient. The simulation of plasma pinch return to equilibrium position after the closure of control coils allows to assess the required parameters of active control system and demonstrate the effect of screen current reverse in twin loops. The obtained results were used to develop the ITER conceptual design and affected the choice of the concept of twin passive loops and new positron of control coils as the basis approaches. 11 refs.; 12 figs.; 1 tab

ITER Safety and Licensing

International Nuclear Information System (INIS)

Girard, J-.P; Taylor, N.; Garin, P.; Uzan-Elbez, J.; GULDEN, W.; Rodriguez-Rodrigo, L.

2006-01-01

The site for the construction of ITER has been chosen in June 2005. The facility will be implemented in Europe, south of France close to Marseille. The generic safety scheme is now under revision to adapt the design to the host country regulation. Even though ITER will be an international organization, it will have to comply with the French requirements in the fields of public and occupational health and safety, nuclear safety, radiation protection, licensing, nuclear substances and environmental protection. The organization of the central team together with its partners organized in domestic agencies for the in-kind procurement of components is a key issue for the success of the experimentation. ITER is the first facility that will achieve sustained nuclear fusion. It is both important for the experimental one-of-a-kind device, ITER itself, and for the future of fusion power plants to well understand the key safety issues of this potential new source of energy production. The main safety concern is confinement of the tritium, activated dust in the vacuum vessel and activated corrosion products in the coolant of the plasma-facing components. This is achieved in the design through multiple confinement barriers to implement the defence in depth approach. It will be demonstrated in documents submitted to the French regulator that these barriers maintain their function in all postulated incident and accident conditions. The licensing process started by examination of the safety options. This step has been performed by Europe during the candidature phase in 2002. In parallel to the final design, and taking into account the local regulations, the Preliminary Safety Report (RPrS) will be drafted with support of the European partner and others in the framework of ITER Task Agreements. Together with the license application, the RPrS will be forwarded to the regulatory bodies, which will launch public hearings and a safety review. Both processes must succeed in order to

Modelling controlled VDE's and ramp-down scenarios in ITER

Science.gov (United States)

Lodestro, L. L.; Kolesnikov, R. A.; Meyer, W. H.; Pearlstein, L. D.; Humphreys, D. A.; Walker, M. L.

2011-10-01

Following the design reviews of recent years, the ITER poloidal-field coil-set design, including in-vessel coils (VS3), and the divertor configuration have settled down. The divertor and its material composition (the latter has not been finalized) affect the development of fiducial equilibria and scenarios together with the coils through constraints on strike-point locations and limits on the PF and control systems. Previously we have reported on our studies simulating controlled vertical events in ITER with the JCT 2001 controller to which we added a PID VS3 circuit. In this paper we report and compare controlled VDE results using an optimized integrated VS and shape controller in the updated configuration. We also present our recent simulations of alternate ramp-down scenarios, looking at the effects of ramp-down time and shape strategies, using these controllers. This work performed under the auspices of the U.S. Department of Energy by LLNL under Contract DE-AC52-07NA27344.

Numerical simulation on bake-out of the ITER diagnostic upper port plug

International Nuclear Information System (INIS)

Pak, S.; Pitcher, C.S.; Kalish, M.R.; Cheon, M.S.; Seon, C.R.; Lee, H.G.

2010-01-01

The diagnostic upper port plug in ITER is fixed to the upper port of the vacuum vessel as a cantilevered beam with bolts and forms a primary vacuum boundary. It needs to be baked out for outgassing before normal operation. This study calculated the required bake-out time and the transient thermal stress during baking for the diagnostic upper port plug. The calculation was done through numerical simulation. The analysis took into consideration the gradual temperature increase of working fluid. In order to look into the effect of radiation heat transfer from the upper port plug to the vacuum vessel port, the upper vacuum vessel port was included in this analysis.

Time-dependent free boundary equilibrium and resistive diffusion in a tokamak plasma

International Nuclear Information System (INIS)

Selig, G.

2012-12-01

In a Tokamak, in order to create the necessary conditions for nuclear fusion to occur, a plasma is maintained by applying magnetic fields. Under the hypothesis of an axial symmetry of the tokamak, the study of the magnetic configuration at equilibrium is done in two dimensions, and is deduced from the poloidal flux function. This function is solution of a non linear partial differential equation system, known as equilibrium problem. This thesis presents the time dependent free boundary equilibrium problem, where the circuit equations in the tokamak coils and passive conductors are solved together with the Grad-Shafranov equation to produce a dynamic simulation of the plasma. In this framework, the Finite Element equilibrium code CEDRES has been improved in order to solve the aforementioned dynamic problem. Consistency tests and comparisons with the DINA-CH code on an ITER vertical instability case have validated the results. Then, the resistive diffusion of the plasma current density has been simulated using a coupling between CEDRES and the averaged one-dimensional diffusion equation, and it has been successfully compared with the integrated modeling code CRONOS. (author)

Disruption modeling in support of ITER

International Nuclear Information System (INIS)

Bandyopadhyay, I.

2015-01-01

Plasma current disruptions and Vertical Displacement Events (VDEs) are one of the major concerns in any tokamak as they lead to large electromagnetic forces to tokamak first wall components and vacuum vessel. Their occurrence also means disruption to steady state operations of tokamaks. Thus future fusion reactors like ITER must ensure that disruptions and VDEs are minimized. However, since there is still finite probability of their occurrence, one must be able to characterize disruptions and VDEs and able to predict, for example, the plasma current quench time and halo current amplitude, which mainly determine the magnitude of the electromagnetic forces. There is a concerted effort globally to understand and predict plasma and halo current evolution during disruption in tokamaks through MHD simulations. Even though Disruption and VDEs are often 3D MHD perturbations in nature, presently they are mostly simulated using 2D axisymmetric MHD codes like the Tokamak Simulation Code (TSC) and DINA. These codes are also extensively benchmarked against experimental data in present day tokamaks to improve these models and their ability to predict these events in ITER. More detailed 3D models like M3D are only recently being developed, but they are yet to be benchmarked against experiments, as also they are massively computationally exhaustive

ITER articulated inspection arm (AIA): R and d progress on vacuum and temperature technology for remote handling

International Nuclear Information System (INIS)

Perrot, Y.; Cordier, J.J.; Friconneau, J.P.; Gargiulo, L.; Martin, E.; Palmer, J.D.; Tesini, A.

2005-01-01

This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R and D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER. It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs. To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 o C). The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility. As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system. This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module

ITER articulated inspection arm (AIA): R and d progress on vacuum and temperature technology for remote handling

Energy Technology Data Exchange (ETDEWEB)

Perrot, Y. [Robotics and Interactive Systems Unit-CEA/LIST, BP6 F-92265 Fontenay aux Roses Cedex (France)]. E-mail: yann.perrot@cea.fr; Cordier, J.J. [DRFC-CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Friconneau, J.P. [Robotics and Interactive Systems Unit-CEA/LIST, BP6 F-92265 Fontenay aux Roses Cedex (France); Gargiulo, L. [DRFC-CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Martin, E. [ITER International Team, Boltzmannstrasse 2, 85748 Garching (Germany); Palmer, J.D. [EFDA CSU Garching, Boltzmannstrasse 2, 85748 Garching (Germany); Tesini, A. [ITER International Team, ITER Naka Joint Work Site, 801-1, Muouyama, Naka-machi, Naka-gun, Iberaki-ken 311-0193 (Japan)

2005-11-15

This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R and D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER. It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs. To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 {sup o}C). The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility. As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system. This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module.

ITER, a major step toward nuclear fusion energy; ITER, une etape majeure vers l'energie de fusion

Energy Technology Data Exchange (ETDEWEB)

Ikeda, K.; Holtkamp, N.; Pick, M.; Gauche, F.; Garin, P.; Bigot, B.; Luciani, J.F.; Mougniot, J.C.; Watteau, J.P.; Saoutic, B.; Becoulet, A.; Libeyre, P.; Beaumont, B.; Simonin, A.; Giancarli, L.; Rosenvallon, S.; Gastaldi, O.; Marbach, G.; Boudot, C.; Ioki, K.; Mitchell, N.; Girard, J.Ph.; Giraud, B.; Lignini, F.; Giguet, E.; Bofusch, E.; Friconneau, J.P.; Di Pace, L.; Pampin, R.; Cook, I.; Maisonnier, D.; Campbell, D.; Hayward, J.; Li Puma, A.; Norajitra, P.; Sardain, P.; Tran, M.Q.; Ward, D.; Moslang, A.; Carre, F.; Serpantie, J.P

2007-01-15

This document gathers together a series of articles dedicated to ITER. They are organized into 5 parts. The first part describes the potential of fusion as a source of energy that will be able to face the challenge of a continuously increasing demand. After a reminder of the main fusion reactions and the conditions to obtain fusion, the second part focuses on the magnetic fusion based concepts with a special emphasis on the tokamak configuration. In the third part the main components of ITER are described: first the plasma facing components, then the vacuum vessel, the superconducting magnets and the heating systems. In the fourth part short papers concerning ITER safety, the maintenance through remote handling systems, the tritium breeding blanket, are given, along with a full article on the waste management. It is interesting to notice that the nuclear wastes will represent: -) between 1600 and 3800 tons of housekeeping and process wastes produced during the 20 years of operation of ITER (20% very low level waste, 75% low or medium activity with short life and 5% medium activity with long life), -) about 750 tons from component replacement during ITER active operation, and -) about 30000 tons from the decommissioning of ITER. The last part presents the European concepts for a power plant based on a fusion reactor. A basic design is given along with a state of the art of the research on the materials that will be used for the structures. It is highlighted that synergies between fission and fusion technologies exist in at least 4 areas: nuclear design code system, high temperature materials, safety approach, and in-service inspection, maintenance and dismantling. (A.C.)

Combining monoenergetic extrapolations from dual-energy CT with iterative reconstructions. Reduction of coil and clip artifacts from intracranial aneurysm therapy

Energy Technology Data Exchange (ETDEWEB)

Winklhofer, Sebastian; Baltsavias, Gerasimos; Michels, Lars; Valavanis, Antonios [University of Zurich, Department of Neuroradiology, University Hospital Zurich, Zurich (Switzerland); Hinzpeter, Ricarda; Stocker, Daniel; Alkadhi, Hatem [University of Zurich, Institute of Diagnostic and Interventional Radiology, University Hospital Zurich, Zurich (Switzerland); Burkhardt, Jan-Karl; Regli, Luca [University of Zurich, Department of Neurosurgery, University Hospital Zurich, Zurich (Switzerland)

2018-03-15

To compare and to combine iterative metal artifact reduction (MAR) and virtual monoenergetic extrapolations (VMEs) from dual-energy computed tomography (DECT) for reducing metal artifacts from intracranial clips and coils. Fourteen clips and six coils were scanned in a phantom model with DECT at 100 and 150SnkVp. Four datasets were reconstructed: non-corrected images (filtered-back projection), iterative MAR, VME from DECT at 120 keV, and combined iterative MAR + VME images. Artifact severity scores and visibility of simulated, contrast-filled, adjacent vessels were assessed qualitatively and quantitatively by two independent, blinded readers. Iterative MAR, VME, and combined iterative MAR + VME resulted in a significant reduction of qualitative (p < 0.001) and quantitative clip artifacts (p < 0.005) and improved the visibility of adjacent vessels (p < 0.05) compared to non-corrected images, with lowest artifact scores found in combined iterative MAR + VME images. Titanium clips demonstrated less artifacts than Phynox clips (p < 0.05), and artifact scores increased with clip size. Coil artifacts increased with coil size but were reducible when applying iterative MAR + VME compared to non-corrected images. However, no technique improved the severe artifacts from large, densely packed coils. Combining iterative MAR with VME allows for an improved metal artifact reduction from clips and smaller, loosely packed coils. Limited value was found for large and densely packed coils. (orig.)

Investigation of the influence of divertor recycling on global plasma confinement in JET ITER-like wall

NARCIS (Netherlands)

Tamain, P.; Joffrin, E.; Bufferand, H.; Jarvinen, A.; Brezinsek, S.; Ciraolo, G.; Delabie, E.; Frassinetti, L.; Giroud, C.; Groth, M.; Lipschultz, B.; Lomas, P.; Marsen, S.; Menmuir, S.; Oberkofler, M.; Stamp, M.; Wiesen, S.; JET-EFDA Contributors,

2015-01-01

Abstract The impact of the divertor geometry on global plasma confinement in type I ELMy H-mode has been investigated in the JET tokamak equipped with ITER-Like Wall. Discharges have been performed in which the position of the strike-points was changed while keeping the bulk plasma equilibrium

Design and integration of lower ports for ITER diagnostic systems

Energy Technology Data Exchange (ETDEWEB)

Casal, Natalia, E-mail: Natalia.casal@iter.org [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Bertalot, Luciano; Cheng, Hao; Drevon, Jean Marc; Duckworth, Philip; Giacomin, Thibaud; Guirao, Julio; Iglesias, Silvia [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Kochergin, Mikhail [IOFFE Institute, Saint Petersburg (Russian Federation); Martin, Alex [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); McCarron, Eddie [Oxford Technologies Ltd., Abingdon (United Kingdom); Mokeev, Alexander [Russian Federation Domestic Agency, Moscow (Russian Federation); Mota, Fernando [CIEMAT, Madrid (Spain); Penot, Christophe; Portales, Mickael [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Kitazawa, Sin-iti [Japanese Domestic Agency, Naka (Japan); Sky, Jack [Oxford Technologies Ltd., Abingdon (United Kingdom); Suarez, Alejandro; Udintsev, Victor; Utin, Yuri [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); and others

2015-10-15

Highlights: • Lower port structures are in its conceptual design phase. • Electromagnetic and seismic loads, will dominate all other mechanical loads. • Design allows diagnostics support, neutron shielding while and signals transmission. • Installation and maintenance operations are fully remote handling compatible. - Abstract: All around the ITER vacuum vessel, forty-four ports will provide access to the vacuum vessel for remote handling operations, diagnostic systems, heating, and vacuum systems: 18 upper ports, 17 equatorial ports, and 9 lower ports. Among the lower ports, three of them will be used for the remote handling installation of the ITER divertor. Once the divertor is in place, these ports will host various diagnostic systems mounted in the so-called diagnostic racks. The diagnostic racks must allow the support and cooling of the diagnostics, extraction of the required diagnostic signals, and providing access and maintainability while minimizing the leakage of radiation toward the back of the port where the humans are allowed to enter. A fully integrated inner rack, carrying the near plasma diagnostic components, will be an stainless steel structure, 4.2 m long, with a maximum weight of 10 t. This structure brings water for cooling and baking at maximum temperature of 240 °C and provides connection with gas, vacuum and electric services. Additional racks (placed away from plasma and not requiring cooling) may be required for the support of some particular diagnostic components. The diagnostics racks and its associated ex vessel structures, which are in its conceptual design phase, are being designed to survive the lifetime of ITER of 20 years. This paper presents the current state of development including interfaces, diagnostic integration, operation and maintenance, shielding requirements, remote handling, loads cases and discussion of the main challenges coming from the severe environment and engineering requirements.

Vessel eddy current characteristics in SST-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

Jana, Subrata; Pradhan, Subrata, E-mail: pradhan@ipr.res.in; Dhongde, Jasraj; Masand, Harish