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Sample records for mftf vacuum vessel

  1. Changing MFTF vacuum environment

    International Nuclear Information System (INIS)

    Margolies, D.; Valby, L.

    1982-12-01

    The Mirror Fusion Test Facility (MFTF) vacuum vessel will be about 60m long and 10m in diameter at the widest point. The allowable operating densities range from 2 x 10 9 to 5 x 10 10 particles per cc. The maximum leak rate of 10 - 6 tl/sec is dominated during operation by the deliberately injected cold gas of 250 tl/sec. This gas is pumped by over 1000 square meters of cryopanels, external sorption pumps and getters. The design and requirements have changed radically over the past several years, and they are still not in final form. The vacuum system design has also changed, but more slowly and less radically. This paper discusses the engineering effort necessary to meet these stringent and changing requirements. Much of the analysis of the internal systems has been carried out using a 3-D Monte Carlo computer code, which can estimate time dependent operational pressures. This code and its use will also be described

  2. Changing MFTF vacuum environment

    International Nuclear Information System (INIS)

    Margolies, D.; Valby, L.

    1982-01-01

    The Mirror Fusion Test Facility (MFTF) vaccum vessel will be about 60m long and 10m in diameter at the widest point. The allowable operating densities range from 2 x 10 9 to 5 x 10 10 particles per cc. The maximum leak rate of 10 -6 tl/sec is dominated during operation by the deliberately injected cold gas of 250 tl/sec. This gas is pumped by over 1000 square meters of cryopanels, external sorbtion pumps and getters. The design and requirements have changed radically over the past several years, and they are still not in final form. The vacuum system design has also changed, but more slowly and less radically. This paper discusses the engineering effort necessary to meet these stringent and changing requirements. Much of the analysis of the internal systems has been carried out using a 3-D Monte Carlo computer code, which can estimate time dependent operational pressures. This code and its use will also be described

  3. Structural design considerations in the Mirror Fusion Test Facility (MFTF-B) vacuum vessel

    International Nuclear Information System (INIS)

    Vepa, K.; Sterbentz, W.H.

    1981-01-01

    In view of favorable results from the Tandem Mirror Experiment (TMX) also at LLNL, the MFTF project is now being rescoped into a large tandem mirror configuration (MFTF-B), which is the mainline approach to a mirror fusion reactor. This paper concerns itself with the structural aspects of the design of the vessel. The vessel and its intended functions are described. The major structural design issues, especially those influenced by the analysis, are described. The objectives of the finite element analysis and their realization are discussed at length

  4. Static and dynamic analyses on the MFTF [Mirror Fusion Test Facility]-B Axicell Vacuum Vessel System: Final report

    International Nuclear Information System (INIS)

    Ng, D.S.

    1986-09-01

    The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory (LLNL) is a large-scale, tandem-mirror-fusion experiment. MFTF-B comprises many highly interconnected systems, including a magnet array and a vacuum vessel. The vessel, which houses the magnet array, is supported by reinforced concrete piers and steel frames resting on an array of foundations and surrounded by a 7-ft-thick concrete shielding vault. The Pittsburgh-Des Moines (PDM) Corporation, which was awarded the contract to design and construct the vessel, carried out fixed-base static and dynamic analyses of a finite-element model of the axicell vessel and magnet systems, including the simulation of various loading conditions and three postulated earthquake excitations. Meanwhile, LLNL monitored PDM's analyses with modeling studies of its own, and independently evaluated the structural responses of the vessel in order to define design criteria for the interface members and other project equipment. The assumptions underlying the finite-element model and the behavior of the axicell vessel are described in detail in this report, with particular emphasis placed on comparing the LLNL and PDM studies and on analyzing the fixed-base behavior with the soil-structure interaction, which occurs between the vessel and the massive concrete vault wall during a postulated seismic event. The structural members that proved sensitive to the soil effect are also reevaluated

  5. MFTF-α+T end cell vacuum vessel and nuclear shield trade studies

    International Nuclear Information System (INIS)

    Kirchner, J.

    1984-01-01

    Three separate and distinct vacuum vessel and nuclear shield trade studies were performed in series. The studies are: vacuum topology, nuclear shield location and composition, and water bulk shield location and material selection

  6. Design of the MFTF external vacuum system

    International Nuclear Information System (INIS)

    Holl, P.M.

    1979-01-01

    As a result of major experiment success in the LLL mirror program on start-up and stabilization of plasmas in minimum-B magnetic geometry, a Mirror Fusion Test Facility (MFTF) is under construction. Completion is scheduled for September, 1981. MFTF will be used to bridge the gap between present day small mirror experiments and future fusion-reactor activity based on magnetic mirrors. The focal point of the Mirror Fusion Test Facility is the 35 foot diameter by 60 foot long vacuum vessel which encloses the superconducting magnets. High vacuum conditions in the vessel are required to establish and maintain a plasma, and to create and deliver energetic neutral atoms to heat the plasma at the central region

  7. Plasma modeling of MFTF-B and the sensitivity to vacuum conditions

    International Nuclear Information System (INIS)

    Porter, G.D.; Rensink, M.

    1984-01-01

    The Mirror Fusion Test Facility (MFTF-B) is a large tandem mirror device currently under construction at Lawrence Livermore National Laboratory. The completed facility will consist of a large variety of components. Specifically, the vacuum vessel that houses the magnetic coils is basically a cylindrical vessel 60 m long and 11 m in diameter. The magnetics system consists of some 28 superconducting coils, each of which is located within the main vacuum vessel. Twenty of these coils are relatively simple solenoidal coils, but the remaining eight are of a more complicated design to provide an octupole component to certain regions of the magnetic field. The vacuum system is composed of a rough vacuum chain, used to evacuate the vessel from atmospheric pressure, and a high vacuum system, used to maintain good vacuum conditions during a plasma shot. High vacuum pumping is accomplished primarily by cryogenic panels cooled to 4.5 0 K. The MFTF-B coil set is shown together with typical axial profiles of magnetic field (a), electrostatic potential (b), and plasma density (c). The plasma is divided into nine regions axially, as labelled on the coil set in Figure 1. The central cell, which is completely azimuthally symmetric, contains a large volume plasma that is confined by a combination of the magnetic fields and the electrostatic potentials in the yin-yang cell

  8. Dynamic testing of MFTF containment-vessel structural system

    International Nuclear Information System (INIS)

    Weaver, H.J.; McCallen, D.B.; Eli, M.W.

    1982-01-01

    Dynamic (modal) testing was performed on the Magnetic Fusion Test Facility (MFTF) containment vessel. The seismic design of this vessel was heavily dependent upon the value of structural damping used in the analysis. Typically for welded steel vessels, a value of 2 to 3% of critical is used. However, due to the large mass of the vessel and magnet supported inside, we felt that the interaction between the structure and its foundation would be enhanced. This would result in a larger value of damping because vibrational energy in the structure would be transferred through the foundation into the surrounding soil. The dynamic test performed on this structure (with the magnet in place) confirmed this later theory and resulted in damping values of approximately 4 to 5% for the whole body modes. This report presents a brief description of dynamic testing emphasizing the specific test procedure used on the MFTF-A system. It also presents an interpretation of the damping mechanisms observed (material and geometric) based upon the spatial characteristics of the modal parameters

  9. Vacuum vessel for the tandem Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Gerich, J.W.

    1986-01-01

    In 1980, the US Department of Energy gave the Lawrence Livermore National Laboratory approval to design and build a tandem Mirror Fusion Test Facility (MFTF-B) to support the goals of the National Mirror Program. We designed the MFTF-B vacuum vessel both to maintain the required ultrahigh vacuum environment and to structurally support the 42 superconducting magnets plus auxiliary internal and external equipment. During our design work, we made extensive use of both simple and complex computer models to arrive at a cost-effective final configuration. As part of this work, we conducted a unique dynamic analysis to study the interaction of the 32,000-tonne concrete-shielding vault with the 2850-tonne vacuum vessel system. To maintain a vacuum of 2 x 10 -8 torr during the physics experiments inside the vessel, we designed a vacuum pumping system of enormous capacity. The vacuum vessel (4200-m 3 internal volume) has been fabricated and erected, and acceptance tests have been completed at the Livermore site. The rest of the machine has been assembled, and individual systems have been successfully checked. On October 1, 1985, we began a series of integrated engineering tests to verify the operation of all components as a complete system

  10. Vacuum distilling vessel

    Energy Technology Data Exchange (ETDEWEB)

    Reik, H

    1928-12-27

    Vacuum distilling vessel for mineral oil and the like, characterized by the ring-form or polyconal stiffeners arranged inside, suitably eccentric to the casing, being held at a distance from the casing by connecting members of such a height that in the resulting space if necessary can be arranged vapor-distributing pipes and a complete removal of the residue is possible.

  11. NCSX Vacuum Vessel Fabrication

    International Nuclear Information System (INIS)

    Viola ME; Brown T; Heitzenroeder P; Malinowski F; Reiersen W; Sutton L; Goranson P; Nelson B; Cole M; Manuel M; McCorkle D.

    2005-01-01

    The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. This paper describes the manufacturing of the vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120 o vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. The port extensions are welded on, leak checked, cut off within 1-inch of the vessel surface at MTM and then reattached at PPPL, to accommodate assembly of the close-fitting modular coils that surround the vessel. The 120 o vessel segments are formed by welding two 60 o segments together. Each 60 o segment is fabricated by welding ten press-formed panels together over a collapsible welding fixture which is needed to precisely position the panels. The vessel is joined at assembly by welding via custom machined 8-inch (20.3 cm) wide spacer ''spool pieces''. The vessel must have a total leak rate less than 5 X 10 -6 t-l/s, magnetic permeability less than 1.02(micro), and its contours must be within 0.188-inch (4.76 mm). It is scheduled for completion in January 2006

  12. PDX vacuum vessel stress analysis

    International Nuclear Information System (INIS)

    Nikodem, Z.D.

    1975-01-01

    A stress analysis of PDX vacuum vessel is described and the summary of results is presented. The vacuum vessel is treated as a toroidal shell of revolution subjected to an internal vacuum. The critical buckling pressure is calculated. The effects of the geometrical discontinuity at the juncture of toroidal shell head and cylindrical outside wall, and the concavity of the cylindrical wall are examined. An effect of the poloidal field coil supports and the vessel outside supports on the stress distribution in the vacuum vessel is determined. A method evaluating the influence of circular ports in the vessel wall on the stress level in the vessel is outlined

  13. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Hagiwara, Koji; Imura, Yasuya.

    1979-01-01

    Purpose: To provide constituted method for easily performing baking of vacuum vessel, using short-circuiting segments. Constitution: At the time of baking, one turn circuit is formed by the vacuum vessel and short-circuiting segments, and current transformer converting the one turn circuit into a secondary circuit by the primary coil and iron core is formed, and the vacuum vessel is Joule heated by an induction current from the primary coil. After completion of baking, the short-circuiting segments are removed. (Kamimura, M.)

  14. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Kurita, Gen-ichi; Onozuka, Masaki; Suzuki, Masaru.

    1997-01-01

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and γ rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  15. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Kurita, Gen-ichi [Japan Atomic Energy Research Inst., Tokyo (Japan); Onozuka, Masaki; Suzuki, Masaru

    1997-07-31

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and {gamma} rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  16. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Nagashima, Keisuke; Suzuki, Masaru; Onozuka, Masaki.

    1997-01-01

    A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

  17. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Nagashima, Keisuke [Japan Atomic Energy Research Inst., Tokyo (Japan); Suzuki, Masaru; Onozuka, Masaki

    1997-07-11

    A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

  18. Vacuum vessel for plasma devices

    International Nuclear Information System (INIS)

    Yamada, Masao; Taguchi, Masami.

    1975-01-01

    Object: To permit effective utility of the space in the inner and outer sides of the container wall and also permit repeated assembly for use. Structure: Vacuum vessel wall sections are sealed together by means of welding bellows, and also flange portions formed at the end of the wall sections are coupled together by bolts and are sealed together with a seal ring and a seal cap secured by welding. (Nakamura, S.)

  19. MFTF-. cap alpha. + T progress report

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, W.D. (ed.)

    1985-04-01

    Early in FY 1983, several upgrades of the Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory (LLNL) were proposed to the fusion community. The one most favorably received was designated MFTF-..cap alpha..+T. The engineering design of this device, guided by LLNL, has been a principal activity of the Fusion Engineering Design Center during FY 1983. This interim progress report represents a snapshot of the device design, which was begun in FY 1983 and will continue for several years. The report is organized as a complete design description. Because it is an interim report, some parts are incomplete; they will be supplied as the design study proceeds. As described in this report, MFTF-..cap alpha..+T uses existing facilities, many MFTF-B components, and a number of innovations to improve on the physics parameters of MFTF-B. It burns deuterium-tritium and has a central-cell Q of 2, a wall loading GAMMA/sub n/ of 2 MW/m/sup 2/ (with a central-cell insert module), and an availability of 10%. The machine is fully shielded, allows hands-on maintenance of components outside the vacuum vessel 24 h after shutdown, and has provisions for repair of all operating components.

  20. MFTF-α + T progress report

    International Nuclear Information System (INIS)

    Nelson, W.D.

    1985-04-01

    Early in FY 1983, several upgrades of the Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory (LLNL) were proposed to the fusion community. The one most favorably received was designated MFTF-α+T. The engineering design of this device, guided by LLNL, has been a principal activity of the Fusion Engineering Design Center during FY 1983. This interim progress report represents a snapshot of the device design, which was begun in FY 1983 and will continue for several years. The report is organized as a complete design description. Because it is an interim report, some parts are incomplete; they will be supplied as the design study proceeds. As described in this report, MFTF-α+T uses existing facilities, many MFTF-B components, and a number of innovations to improve on the physics parameters of MFTF-B. It burns deuterium-tritium and has a central-cell Q of 2, a wall loading GAMMA/sub n/ of 2 MW/m 2 (with a central-cell insert module), and an availability of 10%. The machine is fully shielded, allows hands-on maintenance of components outside the vacuum vessel 24 h after shutdown, and has provisions for repair of all operating components

  1. 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.

  2. 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

  3. Performance of the MFTF magnet cryogenic power leads

    International Nuclear Information System (INIS)

    VanSant, J.H.

    1983-01-01

    The cryogenic power lead system for the MFTF superconducting magnets has been acceptance tested and operated with the magnets. This system, which includes 5-m-long superconducting buses, 1.5-m-long vapor-cooled transition leads, external warm buses, and a cryostack, can conduct up to 6000 A (dc) and operate adiabatically for long periods. We present both design details and performance data; our MFTF version is an example of a reliable lead system for large superconducting magnets contained in a much larger vacuum vessel

  4. Baking results of KSTAR vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported.

  5. Baking results of KSTAR vacuum vessel

    International Nuclear Information System (INIS)

    Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M.

    2009-01-01

    The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported

  6. TORE SUPRA vacuum vessel and shield manufacturing

    International Nuclear Information System (INIS)

    Blateyron, J.; Lepez, R.

    1984-01-01

    TORE SUPRA vacuum vessel and vacuum chamber shield manufacturing in progress at Jeumont-Schneider consists of three main phases: - Detail engineering and manufacturing fixture construction; - Prototype section manufacturing and process preparation; - Construction of the 6 production modules. The welding techniques adopted, call for three special automatic processes: TIG, MIG and PLASMA welding which guarantee mechanical strength, vacuum tightness and absence of distortion. Production of the modules began July 1984. (author)

  7. 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.)

  8. Structural Analysis of the NCSX Vacuum Vessel

    International Nuclear Information System (INIS)

    Fred Dahlgren; Art Brooks; Paul Goranson; Mike Cole; Peter Titus

    2004-01-01

    The NCSX (National Compact Stellarator Experiment) vacuum vessel has a rather unique shape being very closely coupled topologically to the three-fold stellarator symmetry of the plasma it contains. This shape does not permit the use of the common forms of pressure vessel analysis and necessitates the reliance on finite element analysis. The current paper describes the NCSX vacuum vessel stress analysis including external pressure, thermal, and electro-magnetic loading from internal plasma disruptions and bakeout temperatures of up to 400 degrees centigrade. Buckling and dynamic loading conditions are also considered

  9. Device of supporting a vacuum plasma vessel

    International Nuclear Information System (INIS)

    Kanoi, Minoru; Hori, Yasuro.

    1980-01-01

    Purpose: To improve the earthquake-resistance of a vacuum plasma vessel by equalizing the natural vibrations of a vibrating system formed by supporting mechanisms of the respective sectors of the vessel. Constitution: The vacuum plasma vessel is constructed of bellows interposed among a plurality of thick sector-like rings and the rings, which are respectively supported by supporting mechanisms. Thus, the vibrating systems are divided into the rings interposed with the bellows, arms as the supporting mechanisms, and posts. The natural vibrations of these vibrating systems are equalized to each other by suitably adjusting the configurations and the sized of the arms and the posts or the weight or the like of the rings. Therefore, the respective rings become vibrated at the natural vibrations equal to each other so as to largely reduce the stresses produced at both ends of the bellows. Accordingly, it can remarkably improve the earthquake-resistance of the entire plasma vessel. (Sekiya, K.)

  10. Initial conditioning of the TFTR vacuum vessel

    International Nuclear Information System (INIS)

    Dylla, H.F.; Blanchard, W.R.; Krawchuk, R.B.; Hawryluk, R.J.; Owens, D.K.

    1984-01-01

    We report on the initial conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel prior to the initiation of first plasma discharges, and during subsequent operation with high power ohmically-heated plasmas. Following evacuation of the 86 m 3 vessel with the 10 4 1/s high vacuum pumping system, the vessel was conditioned by a 15 A dc glow discharge in H 2 at a pressure of 5 mTorr. Rapid-pulse discharge cleaning was used subsequently to preferentially condition the graphite plasma limiters. The effectiveness of the discharge cleaning was monitored by measuring the exhaust rates of the primary discharge products (CO/C 2 H 4 , CH 4 , and H 2 O). After 175 hours of glow discharge treatment, the equivalent of 50 monolayers of C and O was removed from the vessel, and the partial pressures of impurity gases were reduced to the range of 10 -9 -10 -10 Torr

  11. Expanding plasma jet in a vacuum vessel

    International Nuclear Information System (INIS)

    Chutov, Yu.I.; Kravchenko, A.Yu.; Yakovetskij, V.S.

    1998-01-01

    The paper deals with numerical calculations of parameters of a supersonic quasi-neutral argon plasma jet expanding into a cylindrical vacuum vessel and interacting with its inner surface. A modified method of large particles was used, the complex set of hydrodynamic equations being broken into simpler components, each of which describes a separate physical process. Spatial distributions of the main parameters of the argon plasma jet were simulated at various times after the jet entering the vacuum vessel, the parameters being the jet velocity field, the full plasma pressure, the electron temperature, the temperature of heavy particles, and the degree of ionization. The results show a significant effect of plasma jet interaction on the plasma parameters. The jet interaction with the vessel walls may result e.g. in excitation of shock waves and rotational plasma motions. (J.U.)

  12. 2XIIB vacuum vessel: a unique design

    International Nuclear Information System (INIS)

    Hibbs, S.M.; Calderon, M.O.

    1975-01-01

    The 2XIIB mirror confinement experiment makes unique demands on its vacuum system. The confinement coil set encloses a cavity whose surface is comprised of both simple and compound curves. Within this cavity and at the core of the machine is the operating vacuum which is on the order of 10 -9 Torr. The vacuum container fits inside the cavity, presenting an inside surface suitable for titanium getter pumping and a means of removing the heat load imposed by incandescent sublimator wires. In addition, the cavity is constructed of nonmagnetic and nonconducting materials (nonmetals) to avoid distortion of the pulsed confinement field. It is also isolated from mechanical shocks induced in the machine's main structure when the coils are pulsed. This paper describes the design, construction, and operation of the 2XIIB high-vacuum vessel that has been performing successfully since early 1974

  13. The TPX vacuum vessel and in-vessel components

    International Nuclear Information System (INIS)

    Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

    1994-01-01

    The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

  14. Structural analysis of the KSTAR vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    In, Sang Ryul; Yoon, Byeong Joo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    Structure analysis of the vacuum vessel for the KSTAR tokamak which, is in the end phase of the conceptual design have been performed. Mechanical stresses and deformations of the vessel produced by constant forces due to atmospheric pressure, dead weight, fluid pressure, etc and various transient electromagnetic forces induced during tokamak operations were calculated as well as modal characteristics and buckling properties were investigated. Influences of the temperature gradient and the constraint condition of the support on the thermal stress and deformation of the vessel were analyzed. The thermal stress due to the temperature distribution on the vessel as supplying the N{sub 2} gas of 400 deg C through poloidal channels according to the recent baking concept were calculated. No severe problem in the robustness of the vessel was found when applying the constant pressures on the vessel. However the mechanical stress due to the EM force induced by halo currents flowing on the vessel and the plasma facing components (PFCs) far exceeded the allowable limit. Some reinforcing components should be added on the boundary of the PFC support and the vessel, and that of the vessel support and the vessel. A steep temperature gradient in the vicinity of the inlet and oulet of the heating gas produced a thermal stress much higher than allowable. It is necessary to make the temperature of the vessel as uniform as possible and to develop a new support concept which is flexible enough to accommodate a thermal expansion of a few cm while sufficiently strong to resist mechanical impacts. (author). 5 refs., 41 figs., 9 tabs.

  15. Status of the ITER vacuum vessel construction

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C.H.; Sborchia, C.; Ioki, K.; Giraud, B.; Utin, Yu.; Sa, J.W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wang, X., E-mail: xiaoyuwww@gmail.com [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Teissier, P.; Martinez, J.M.; Le Barbier, R.; Jun, C.; Dani, S.; Barabash, V.; Vertongen, P.; Alekseev, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Jucker, P.; Bayon, A. [F4E, c/ Josep Pla, n. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Pathak, H.; Raval, J. [ITER-India, IPR, A-29, Electronics Estate, GIDC, Sector-25, Gandhinagar 382025 (India); Ahn, H.J. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2014-10-15

    Highlights: • Final design of the ITER vacuum vessel (VV). • Procurement of the ITER VV. • Manufacturing results of real scale mock-ups. • Manufacturing status of the VV in domestic agencies. - Abstract: The ITER vacuum vessel (VV) is under manufacturing by four domestic agencies after completion of engineering designs that have been approved by the Agreed Notified Body (ANB). Manufacturing designs of the VV have been being completed, component by component, by accommodating requirements of the RCC-MR 2007 edition. Manufacturing of the VV first sector has been started in February 2012 in Korea and in-wall shielding in May 2013 in India. EU will start manufacturing of its first sector from September 2013 and Russia the upper port by the end of 2013. All DAs have manufactured several mock-ups including real-size ones to justify/qualify and establish manufacturing techniques and procedures.

  16. Vacuum vessel for a nuclear fusion device

    International Nuclear Information System (INIS)

    Watanabe, Takashi; Sato, Hiroshi; Owada, Koro.

    1976-01-01

    Object: To provide a reinforcing member on a bellows portion to reduce a stress at the bellows portion thereby increasing the strength of a vessel. Structure: A vacuum vessel for a nuclear fusion device has a bellows portion and a wall thick portion. A support extended toward the bellows portion is secured inside of a toroidal section in order to reduce the stress at the bellows portion. An insulator is interposed between the support and the bellows portion and is retained on the support by a bolt. Since the stress may be reduced by the support, the wall thick of the bellows portion may be decreased to sufficiently secure the low electric resistance value. (Yoshihara, H.)

  17. Seismic analysis of the MFTF facility

    International Nuclear Information System (INIS)

    Maslenikov, O.R.; Johnson, J.J.; Tiong, L.W.; Mraz, M.J.

    1985-01-01

    Seismic analyses were performed on the Mirror Fusion Test Facility (MFTF-B) located at the Lawrence Livermore National Laboratory, Livermore, CA. The three major structures studied were the vacuum vessel, the concrete shielding vault, and the steel frame enclosure building. The analyses performed on these structures ranged from fixed-base response spectrum analyses to soil-structure interaction analyses including the effects of structure-to-structure interaction and foundation flexibility. The results of these studies showed that the presence of the vault significantly affects the response of the vessel; that modeling the flexibility of the vault footing is important when studying forces near the base of the wall; and that the vault had very little effect on the building response. (orig.)

  18. Shielding performance of the NET vacuum vessel

    International Nuclear Information System (INIS)

    Arkuszewski, J.J.; Jaeger, J.F.

    1988-01-01

    To corroborate 1-D deterministic shielding calculations on the Next European Torus (NET) vacuum vessel/shield and shielding blanket, 3-D Monte Carlo calculations have been done with the MCNP code. This should provide information on the poloidal and the toroidal variations. Plasma source simulation and the geometrical model are described, as are other assumptions. The calculations are based on the extended plasma power of 714 MW. The results reported here are the heat deposition in various parts of the device, on the one hand, and the neutron and photon currents at the outer boundary of the vacuum vessel, on the other hand. The latter are needed for the detailed design of the super-conducting magnetic coils. A reasonable statistics has been obtained on the outboard side of the torus, though this cannot be said for the inboard side. The inboard is, however, much more toroidally symmetric than the outboard, so that other methods could be applied such as 2-D deterministic calculations, for instance. (author) 4 refs., 44 figs., 42 tabs

  19. 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)

  20. TPX vacuum vessel transient thermal and stress conditions

    International Nuclear Information System (INIS)

    Feldshteyn, Y.; Dinkevich, S.; Feng, T.; Majumder, D.

    1995-01-01

    The TPX vacuum vessel provides the vacuum boundary for the plasma and the mechanical support for the internal components. Another function of the vacuum vessel is to contain neutron shielding water in the double wall space during normal operation. This double wall space serves as a heat reservoir for the entire vacuum vessel during bakeout. The vacuum vessel and the internal components are subjected to thermal stresses induced by a nonuniform temperature distribution within the structure during bakeout. A successful Conceptual Design Review in March 1993 has established superheated steam as the heating source of the vacuum vessel. A transient bakeout mode of the vacuum vessel and in-vessel components has been analyzed to evaluate transient period duration, proper temperature level, actual thermal stresses and performance of the steam equipment. Thermally, the vacuum vessel structure may be considered as an adiabatic system because it is perfectly insulated by the strong surrounding vacuum and multiple layers of superinsulation. Important aspects of the analysis are described herein

  1. 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.)

  2. Fabrication of the vacuum vessel for JT-60 machine upgrade

    International Nuclear Information System (INIS)

    Uchikawa, T.; Takanabe, K.; Tsujimura, S.; Ue, K.; Oka, K.; Kuri, S.; Ioki, K.; Namiki, K.; Suzuki, Y.; Horliike, H.; Ninomiya, H.; Yamamoto, M.; Neyatani, Y.; Ando, T.; Matsukawa, M.

    1992-01-01

    The JT-60 tokamak was upgraded to double the plasma current to 6 MA. In the JT-60 machine upgrade (JT-60U), the vacuum vessel and poloidal field (PF) coils were renewed. The new vacuum vessel features a three-dimensionally curved, thin double-skin torus with multi-arc D-shaped cross section. The double-skin structure is strengthened with square pipes placed in between the outer and inner skins. Fabrication and site installation of the vessel was smoothly completed in February, 1991. This paper describes an overview of the JT-60U vacuum vessel construction

  3. 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

  4. Report on the experience with the Supervisory Control and Diagnostics System (SCDS) of MFTF-B

    International Nuclear Information System (INIS)

    Wyman, R.H.

    1983-01-01

    The Supervisory Control and Diagnostics System (SCDS) of MFTF is a multiprocessor computer system using graphics oriented displays with touch sensitive panels as the primary operator interface. Late in the calendar year 1981 the system was used to control an integrated test of the vacuum vessel, vacuum system, cryogenics system and the superconducting magnet of MFTF. Since the completion of those tests and starting in early calendar 1983 the system has been used for control of the neutral beam test facility at LLNL. This paper presents a short overview of SCDS for the purpose of orientation and then proceeds to describe the difficulties encountered in these preliminary encounters with reality. The band-aids used to hold things together as disaster threatened as well as the long-term solutions to the problems will be discussed. Finally, we will present some comments on system costs and management philosophy

  5. Device for supporting the vacuum vessel of a thermonuclear device

    International Nuclear Information System (INIS)

    Sato, Hiroshi.

    1980-01-01

    Purpose: To hold a vacuum vessel securely at a predetermined position. Constitution: A vacuum vessel is supported on its one side to the standard mounting location of a support frame by way of a pin junction. The vacuum vessel is provided at its upper and lower positions with movable mounting portions, which are connected by way of connecting rods to fixed mounting locations on the upper and lower frames. The fixed mounting locations are disposed on a vertical plane including the axis of the torus center. This arrangement enables to hold even a large vacuum vessel at an exact predetermined position even under high temperature conditions without limiting the container's thermal expansion relative to the changes in temperature, thereby providing an extremely high rigidity against electromagnetic forces, earthquakes, etc. (Furukawa, Y.)

  6. Design and manufacturing of vacuum vessel of TPE-RX

    Energy Technology Data Exchange (ETDEWEB)

    Sago, H.; Kaguchi, H.; Orita, J.; Ishigami, Y. [Mitsubishi Heavy Industries Ltd., Kobe (Japan); Urata, K. [Mitsubishi Heavy Industries Ltd. (Japan). Nuclear Energy Systems Engineering Center; Hasegawa, M. [Mitsubishi Electric Co. (Japan). Nuclear Fusion Development; Yagi, Y.; Hirano, Y.; Shimada, T.; Sekine, S.; Sakakita, H. [Electrotechnical Lab. (Japan)

    1998-07-01

    Construction of a new, large reversed field pinch (RFP) machine called TPE-RX was complete at the end of 1997 as a successor of the previous TPE-1RM20 machine at the Electrotechnical Laboratory (ETL). RFP configuration has been successfully obtained in March 1998. This paper introduces structural design and manufacturing of the vacuum vessel of TPE-RX. The support positions were decided by structural analyses. The structural integrity of the vacuum vessel was evaluated by inelastic analyses. (author)

  7. Design and manufacturing of vacuum vessel of TPE-RX

    International Nuclear Information System (INIS)

    Sago, H.; Kaguchi, H.; Orita, J.; Ishigami, Y.; Urata, K.

    1998-01-01

    Construction of a new, large reversed field pinch (RFP) machine called TPE-RX was complete at the end of 1997 as a successor of the previous TPE-1RM20 machine at the Electrotechnical Laboratory (ETL). RFP configuration has been successfully obtained in March 1998. This paper introduces structural design and manufacturing of the vacuum vessel of TPE-RX. The support positions were decided by structural analyses. The structural integrity of the vacuum vessel was evaluated by inelastic analyses. (author)

  8. Depressurization as a means of leak checking large vacuum vessels

    International Nuclear Information System (INIS)

    Callis, R.W.; Langhorn, A.; Petersen, P.I.; Ward, C.; Wesley, J.

    1985-01-01

    A common problem associated with large vacuum vessels used in magnetic confinement fusion experiments is that leak checking is hampered by the inaccessibility to most of the vacuum vessel surface. This inaccessibility is caused by the close proximity of magnetic coils, diagnostics and, for those vessels that are baked, the need to completely surround the vessel with a thermal insulation blanket. These obstructions reduce the effectiveness of the standard leak checking method of using a mass spectrometer and spraying a search gas such as helium on the vessel exterior. Even when the presence of helium is detected, its entry point into the vessel cannot always be pinpointed. This paper will describe a method of overcoming this problem. By slightly depressurizing the vessel, an influx of helium through the leak is created. The leak site can then be identified by personnel within the vessel using standard sniffing procedures. There are two conditions which make this method of leak checking practical. First, the vessel need only be depressurized 2 psi, thus allowing personnel inside to perform the sniffing operation. Second, the sniffing probe used (Leybold--Heraus ''Quick Test'') could detect a change in helium concentration as small as 100 ppb, which allows for faster scanning of the vessel inferior. Use of this technique to find an elusive 10 -3 Torrxl/s leak in the Doublet III tokamak vacuum vessel will be presented

  9. 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

  10. 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.)

  11. MFTF sensor verification computer program

    International Nuclear Information System (INIS)

    Chow, H.K.

    1984-01-01

    The design, requirements document and implementation of the MFE Sensor Verification System were accomplished by the Measurement Engineering Section (MES), a group which provides instrumentation for the MFTF magnet diagnostics. The sensors, installed on and around the magnets and solenoids, housed in a vacuum chamber, will supply information about the temperature, strain, pressure, liquid helium level and magnet voltage to the facility operator for evaluation. As the sensors are installed, records must be maintained as to their initial resistance values. Also, as the work progresses, monthly checks will be made to insure continued sensor health. Finally, after the MFTF-B demonstration, yearly checks will be performed as well as checks of sensors as problem develops. The software to acquire and store the data was written by Harry Chow, Computations Department. The acquired data will be transferred to the MFE data base computer system

  12. Baking of SST-1 vacuum vessel modules and sectors

    International Nuclear Information System (INIS)

    Pathan, Firozkhan S; Khan, Ziauddin; Yuvakiran, Paravastu; George, Siju; Ramesh, Gattu; Manthena, Himabindu; Shah, Virendrakumar; Raval, Dilip C; Thankey, Prashant L; Dhanani, Kalpesh R; Pradhan, Subrata

    2012-01-01

    SST-1 Tokamak is a steady state super-conducting tokamak for plasma discharge of 1000 sec duration. The plasma discharge of such long time duration can be obtained by reducing the impurities level, which will be possible only when SST-1 vacuum chamber is pumped to ultra high vacuum. In order to achieve UHV inside the chamber, the baking of complete vacuum chamber has to be carried out during pumping. For this purpose the C-channels are welded inside the vacuum vessel. During baking of vacuum vessel, these welded channels should be helium leak tight. Further, these U-channels will be in accessible under operational condition of SST-1. So, it will not possible to repair if any leak is developed during experiment. To avoid such circumstances, a dedicated high vacuum chamber is used for baking of the individual vacuum modules and sectors before assembly so that any fault during welding of the channels will be obtained and repaired. This paper represents the baking of vacuum vessel modules and sectors and their temperature distribution along the entire surface before assembly.

  13. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

    Mardani, M.; Amrollahi, R.; Koohestani, S.

    2012-01-01

    Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

  14. Design features of the solenoid magnets for the central cell of the MFTF-B

    International Nuclear Information System (INIS)

    Wohlwend, J.W.; Tatro, R.E.; Ring, D.S.

    1981-01-01

    The 14 superconducting solenoid magnets which form the central cell of the MFTF-B are being designed and fabricated by General Dynamics for the Lawrence Livermore National Laboratory. Each solenoid coil has a mean diameter of five meters and contains 600 turns of a proven conductor type. Structural loading resulting from credible fault events, cooldown and warmup requirements, and manufacturing processes consistent with other MFTF-B magnets have been considered in the selection of 304 LN as the structural material for the magnet. The solenoid magnets are connected by 24 intercoil beams and 20 solid struts which resist the longitudinal seismic and electromagnetic attractive forces and by 24 hanger/side supports which react magnet dead weight and seismic loads. A modular arrangement of two solenoid coils within a vacuum vessel segment allow for sequential checkout and installation

  15. Integration of ITER in-vessel diagnostic components in the vacuum vessel

    International Nuclear Information System (INIS)

    Encheva, A.; Bertalot, L.; Macklin, B.; Vayakis, G.; Walker, C.

    2009-01-01

    The integration of ITER in-vessel diagnostic components is an important engineering activity. The positioning of the diagnostic components must correlate not only with their functional specifications but also with the design of the major parts of ITER torus, in particular the vacuum vessel, blanket modules, blanket manifolds, divertor, and port plugs, some of which are not yet finally designed. Moreover, the recently introduced Edge Localised Mode (ELM)/Vertical Stability (VS) coils mounted on the vacuum vessel inner wall call for not only more than a simple review of the engineering design settled down for several years now, but also for a change in the in-vessel distribution of the diagnostic components and their full impact has yet to be determined. Meanwhile, the procurement arrangement (a document defining roles and responsibilities of ITER Organization and Domestic Agency(s) (DAs) for each in-kind procurement including technical scope of work, quality assurance requirements, schedule, administrative matters) for the vacuum vessel must be finalized. These make the interface process even more challenging in terms of meeting the vacuum vessel (VV) procurement arrangement's deadline. The process of planning the installation of all the ITER diagnostics and integrating their installation into the ITER Integrated Project Schedule (IPS) is now underway. This paper covers the progress made recently on updating and issuing the interfaces of the in-vessel diagnostic components with the vacuum vessel, outlines the requirements for their attachment and summarises the installation sequence.

  16. Testing program for burning plasma experiment vacuum vessel bolted joint

    International Nuclear Information System (INIS)

    Hsueh, P.K.; Khan, M.Z.; Swanson, J.; Feng, T.; Dinkevich, S.; Warren, J.

    1992-01-01

    As presently designed, the Burning Plasma Experiment vacuum vessel will be segmentally fabricated and assembled by bolted joints in the field. Due to geometry constraints, most of the bolted joints have significant eccentricity which causes the joint behavior to be sensitive to joint clamping forces. Experience indicates that as a result of this eccentricity, the joint will tend to open at the side closest to the applied load with the extent of the opening being dependent on the initial preload. In this paper analytical models coupled with a confirmatory testing program are developed to investigate and predict the non-linear behavior of the vacuum vessel bolted joint

  17. Vacuum vessel of thermonuclear device and manufacturing method thereof

    International Nuclear Information System (INIS)

    Kurita, Genichi; Nagashima, Keisuke; Uchida, Takaho; Shibui, Masanao; Ebisawa, Katsuyuki; Nakagawa, Satoshi.

    1997-01-01

    The present invention provides a vacuum vessel of a thermonuclear device using, as a material of a plasma vacuum vessel, a material to be less activated and having excellent strength as well as a manufacturing method thereof. Namely, the vacuum vessel is made of titanium or a titanium alloy. In addition, a liner layer comprising a manganese alloy, nickel alloy, nickel-chromium alloy or aluminum or aluminum alloy is formed. With such a constitution, the wall substrate made of titanium or a titanium alloy can be isolated by the liner from hydrogen or plasmas. As a result, occlusion of hydrogen to titanium or the titanium alloy can be prevented thereby enabling to prevent degradation of the material of the wall substrate of the vacuum vessel. In addition, since the liner layer has relatively high electric resistance, a torus circumferential resistance value required for plasma ignition can be ensured by using it together with the vessel wall made of titanium alloy. (I.S.)

  18. Vacuum vessels for the LHC magnets arrive at CERN

    CERN Multimedia

    2001-01-01

    The first batch of pre-series vacuum vessels for the LHC dipole magnets has just been delivered to CERN. The vessels are components of the cryostats and will provide the thermal insulation for the superconducting magnets. The first batch of vacuum vessels for the LHC dipole magnets with the team taking part at CERN in ordering and installing them. Left to right : Claude Hauviller, Monique Dupont, Lloyd Williams, Franck Gavin, Alain Jacob, Christophe Vuitton, Davide Bozzini, Laure Sandri, Mikael Sjoholm and André de Saever. In 2006 all that will be seen of the LHC superconducting dipoles in the LHC tunnel will be a line of over 1230 blue cylindrical vacuum vessels. Ten vessels, each weighing 4 tonnes, are already at CERN. On 6 July the first batch of pre-series vessels reached the Lab-oratory from the firm SIMIC Spa whose works are near Savona in north-western Italy. Despite appearances, these 15-metre long, 1-metre diameter blue tubes are much more sophisticated than sections of a run-of-the-mill...

  19. Welding distortion control in double walled KSTAR vacuum vessel fabrication

    International Nuclear Information System (INIS)

    Oh, D. W.; Lee, G. T.; Kim, H. K.; Yang, H. L.; Bak, J. S.

    2004-01-01

    The KSTAR(Korea Superconducting Tokamak Advanced Research) vacuum vessel is designed to be a double walled structure made of 12mm thick 316LN stainless steel with a D shaped cross-section about 4 m height. Vacuum vessel was pre-fabricated in two parts, 180 degree and 157.5 degree sectors in toroidal direction to meet the transportation purpose. These two parts have to be welded on site with ±2mm allowable fabrication tolerances. 1/3 scaled mock-up model was used to estimate the welding distortion and to ensure the weld quality of vacuum vessel. Gas Tungsten Arc Welding(GTAW), which has been approved by procedure qualification test, was used during mock-up test and vacuum vessel site fabrication. Welding distortion could be managed by allowing for distortion in opposite direction, by applying high restraint using lots of strong backs, by controlling the welding heat input with symmetrical welding sequence. The integrity of the site welding joint was assured by radiographic test, ultrasonic test and leak test with helium detecting method

  20. Analysis of effective electrical parameters for CFETR vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xufeng; Xu, Weiwei, E-mail: wwxu@ipp.ac.cn; Du, Shuangsong; Zheng, Jinxing

    2016-11-15

    Highlights: • The eddy current distribution and variation of CFETR vacuum vessel during plasma disruption have been calculated. • Effective electrical parameters can be derived from the eddy current characters. • The method for eddy current and effective electrical parameters is suit for the complex shell with arbitrary shape. - Abstract: The electrical parameters of CFETR (China Fusion Engineering Test Reactor) vacuum vessel are very important to the design of control system and power supply system. Effective electrical parameters are relevant to the dynamic of eddy current. For complex structure, the distribution of eddy current can’t be obtained by analytical form. A method is presented to solve the eddy current of the vacuum vessel in this paper. The effective electrical parameters can be got from the eddy current distribution and variation. The time constant of the CFETR vacuum vessel is derived from the decay characteristics of the eddy current. And the effective resistance and inductance can be derived from the viewpoint of energy for a certain distribution of eddy current.

  1. Structural design and manufacturing of TPE-RX vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sago, H.; Orita, J.; Kaguchi, H.; Ishigami, Y. [Mitsubishi Heavy Ind. Ltd., Kobe (Japan); Urata, K.; Kudough, F. [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan); Hasegawa, M.; Oyabu, I. [Mitsubishi Electric Co., Tokyo (Japan); Yagi, Y.; Sekine, S.; Shimada, T.; Hirano, Y.; Sakakita, H.; Koguchi, H. [Electrotechnical Laboratory, Tsukuba (Japan)

    1999-10-01

    TPE-RX is a newly constructed, large-sized reversed field pinch (RFP) machine installed at the Electrotechnical Laboratory of the Ministry of International Trade and Industry in Japan. This is the third largest RFP in the world. Major and minor radii of the plasma are 1.72 and 0.45 m, respectively. TPE-RX aims to optimize plasma confinement up to 1 MA. RFP plasma configuration was successfully obtained in March 1998. This paper reports the structural design and manufacturing of the vacuum vessel of TPE-RX. The supporting system on the bellows sections of the vessel was designed based on a detailed finite element method. The integrity of the vacuum vessel against a plasma disruption has been confirmed using dynamic inelastic analyses. (orig.)

  2. Structural design and manufacturing of TPE-RX vacuum vessel

    International Nuclear Information System (INIS)

    Sago, H.; Orita, J.; Kaguchi, H.; Ishigami, Y.; Urata, K.; Kudough, F.; Hasegawa, M.; Oyabu, I.; Yagi, Y.; Sekine, S.; Shimada, T.; Hirano, Y.; Sakakita, H.; Koguchi, H.

    1999-01-01

    TPE-RX is a newly constructed, large-sized reversed field pinch (RFP) machine installed at the Electrotechnical Laboratory of the Ministry of International Trade and Industry in Japan. This is the third largest RFP in the world. Major and minor radii of the plasma are 1.72 and 0.45 m, respectively. TPE-RX aims to optimize plasma confinement up to 1 MA. RFP plasma configuration was successfully obtained in March 1998. This paper reports the structural design and manufacturing of the vacuum vessel of TPE-RX. The supporting system on the bellows sections of the vessel was designed based on a detailed finite element method. The integrity of the vacuum vessel against a plasma disruption has been confirmed using dynamic inelastic analyses. (orig.)

  3. New baking system for the RFX vacuum vessel

    International Nuclear Information System (INIS)

    Collarin, P.; Luchetta, A.; Sonato, P.; Toigo, V.; Zaccaria, P.; Zollino, G.

    1996-01-01

    A heating system based on eddy currents has been developed for the vacuum vessel of the RFX Reversed Field Pinch device. After a testing phase, carried out at low power, the final power supply system has been designed and installed. It has been used during last year to bake out the vessel and the graphite first wall up to 320 degree C. Recently the heating system has been completed with a control system that allows for baking sessions with an automatic control of the vacuum vessel temperature and for pulse sessions with a heated first wall. After the description of the preliminary analyses and tests, and of the main characteristics of the power supply and control systems, the experimental results of the baking sessions performed during last year are presented. 6 refs., 7 figs

  4. New baking system for the RFX vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Collarin, P.; Luchetta, A.; Sonato, P.; Toigo, V.; Zaccaria, P.; Zollino, G. [Universita di Padova (Italy)

    1996-12-31

    A heating system based on eddy currents has been developed for the vacuum vessel of the RFX Reversed Field Pinch device. After a testing phase, carried out at low power, the final power supply system has been designed and installed. It has been used during last year to bake out the vessel and the graphite first wall up to 320{degree}C. Recently the heating system has been completed with a control system that allows for baking sessions with an automatic control of the vacuum vessel temperature and for pulse sessions with a heated first wall. After the description of the preliminary analyses and tests, and of the main characteristics of the power supply and control systems, the experimental results of the baking sessions performed during last year are presented. 6 refs., 7 figs.

  5. Manufacturing the MFTF magnet

    International Nuclear Information System (INIS)

    Dalder, E.N.C.; Hinkle, R.E.; Hodges, A.J.

    1980-01-01

    The Mirror Fusion Test Facility (MFTF) is a large mirror program experiment for magnetic fusion energy. It will combine and extend the near-classical plasma confinement achieved in 2XIIB with advanced neutral-beam and magnet technologies. The product of ion density and confinement time will be improved more than an order of magnitude, while the superconducting magnet weight will be extrapolated from 15 tons in Baseball II to 375 tons in MFTF. Recent reactor studies show that the MFTF will traverse much of the distance in magnet technology towards the reactor regime

  6. Manufacture, testing and assembly preparation of the JET vacuum vessel

    International Nuclear Information System (INIS)

    Arbez, J.; Baeumel, S.; Dean, J.R.; Duesling, G.; Froger, C.; Hemmerich, J.L.; Walravens, M.; Walter, K.; Winkel, T.

    1983-01-01

    To reach the target pressure of 10 -9 mbar, JET's double-walled Inconel vacuum vessel is being manufactured and assembled in clean conditions and with meticulous leak detection. Each octant (1/8 of the torus) is baked in an oven to 520 0 C and leak tested at 350 0 C to reveal leaks as small as 10 -9 mbar l/s, which are repaired. In service the vessel will be baked periodically to 500 0 C by CO 2 passing between its walls. The single-walled ports will be electrically heated. (author)

  7. Fabrication progress of the ITER vacuum vessel sector in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.C., E-mail: bckim@nfri.re.kr [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Lee, Y.J.; Hong, K.H.; Sa, J.W.; Kim, H.S.; Park, C.K.; Ahn, H.J.; Bak, J.S.; Jung, K.J. [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Park, K.H.; Roh, B.R.; Kim, T.S.; Lee, J.S.; Jung, Y.H.; Sung, H.J.; Choi, S.Y.; Kim, H.G.; Kwon, I.K.; Kwon, T.H. [Hyundai Heavy Industries Co. Ltd., Dong-gu, Ulsan (Korea, Republic of)

    2013-10-15

    Highlights: ► Fabrication of ITER vacuum vessel sector full scale mock-up to develop fabrication procedures. ► The welding and nondestructive examination techniques conform to RCC-MR. ► The preparation of real manufacturing of ITER vacuum vessel sector. -- Abstract: As a participant of ITER project, ITER Korea has to supply two ITER vacuum vessel sectors (Sector no. 6, no. 1) of total nine ITER VV sectors. After the procurement arrangement with ITER Organization, ITER Korea made the contract with Hyundai Heavy Industries (HHI) for fabrication of two sectors. Then the start of the manufacturing design was initiated from January 2010. HHI made three real scale R and D mock-ups to verify the critical fabrication feasibility issues on electron beam welding, 3D forming, welding distortion and achievable tolerances. The documentation according to IO and the French nuclear safety regulation requirement, the qualification of welding and nondestructive examination procedures conform to RCC-MR 2007 were proceed in parallel. The mass production of raw material was done after receiving ANB (agreed notified body) verification of product/parts and shop qualification. The manufacturing drawing, manufacturing and inspection plan of VV sector with supporting fabrication procedures are also verified by ANB, accordingly the first cutting and forming of plates for VV sector fabrication started from February 2012. This paper reports the latest fabrication progress of ITER vacuum vessel Sector no. 6 that will be assembled as the first sector in the ITER pit. The overall fabrication route, R and D mock-up fabrication results with forming and welding distortion analysis, qualification status of welding and nondestructive examination (NDE) are also presented.

  8. 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).

  9. ATLAS Supplier Award for the ECT Vacuum Vessels

    CERN Multimedia

    Jenni, P

    On 12 February the Netherlands firm Schelde Exotech was awarded the ATLAS Supplier Award for the construction of the two vacuum vessels for the ATLAS End- Cap Toroid (ECT) magnets. ATLAS Supplier Award ceremonies have now become something of a tradition. For the third consecutive year, ATLAS has given best supplier awards for the most exceptional contributors to the construction of the detector. The Netherlands firm Schelde Exotech has just received the award for the construction of the two vacuum vessels for the ECTs. With a diameter of 11 metres and a volume of 550 cubic metres, the ECT vacuum vessels are obviously impressive in scale. They consist of large aluminium plates and a stainless steel central bore tube. In order to obtain the required undulations, the firm had to develop a special assembly and welding technique. Despite the chambers' imposing size, a very high degree of precision has been achieved in their geometry. Moreover, the chambers, which were delivered in July 2002 to CERN, were built i...

  10. 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).

  11. MFTF TOTAL benchmark

    International Nuclear Information System (INIS)

    Choy, J.H.

    1979-06-01

    A benchmark of the TOTAL data base management system as applied to the Mirror Fusion Test Facility (MFTF) data base was implemented and run in February and March of 1979. The benchmark was run on an Interdata 8/32 and involved the following tasks: (1) data base design, (2) data base generation, (3) data base load, and (4) develop and implement programs to simulate MFTF usage of the data base

  12. 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

  13. JT-60SA vacuum vessel manufacturing and assembly

    Energy Technology Data Exchange (ETDEWEB)

    Masaki, Kei, E-mail: masaki.kei@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Shibama, Yusuke K.; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The design of the JT-60SA vacuum vessel body was completed with the demonstration of manufacturing procedure by the mock-up fabrication of the 20 Degree-Sign upper half of VV. Black-Right-Pointing-Pointer The actual VV manufacturing has started since November 2009. Black-Right-Pointing-Pointer The first product of the VV 40 Degree-Sign sector was completed in May 2011. Black-Right-Pointing-Pointer A basic VV assembly scenario and procedure were studied to complete the 360 Degree-Sign VV including positioning method and joint welding. - Abstract: The JT-60SA vacuum vessel (VV) has a D-shaped poloidal cross section and a toroidal configuration with 10 Degree-Sign segmented facets. A double wall structure is adopted to ensure high rigidity at operational load and high toroidal one-turn resistance. The material is 316L stainless steel with low cobalt content (<0.05%). The design temperatures of the VV at plasma operation and baking are 50 Degree-Sign C and 200 Degree-Sign C, respectively. In the double wall, boric-acid water is circulated at plasma operation to reduce the nuclear heating of the superconducting magnets. For baking, nitrogen gas is circulated in the double wall after draining of the boric-acid water. The manufacturing of the VV started in November 2009 after a fundamental welding R and D and a trial manufacturing of 20 Degree-Sign upper half mock-up. The manufacturing of the first VV 40 Degree-Sign sector was completed in May 2011. A basic concept and required jigs of the VV assembly were studied. This paper describes the design and manufacturing of the vacuum vessel. A plan of VV assembly in torus hall is also presented.

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. Development of a master model concept for DEMO vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe, E-mail: peppe.digironimo@gmail.com [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy)

    2016-11-15

    Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

  20. Development of a master model concept for DEMO vacuum vessel

    International Nuclear Information System (INIS)

    Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea; Bachmann, Christian; Di Gironimo, Giuseppe

    2016-01-01

    Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

  1. 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)

  2. Manufacturing technology development for vacuum vessel and plasma facing components

    International Nuclear Information System (INIS)

    Laitinen, Arttu; Liimatainen, Jari; Hallila, Pentti

    2005-01-01

    Vacuum vessel and plasma facing components of the ITER construction including shield modules and primary first wall panels have great impact on the production costs and reliability of the installation. From the manufacturing technology point of view, accuracy of shape, properties of the various austenitic stainless steel/austenitic stainless steel interfaces or CuCrZr/austenitic stainless steel interfaces as well as those of the base materials are crucial for technical reliability of the construction. The current approach in plasma facing components has been utilisation of solid-HIP technology and solid-powder-HIP technology. Due to the large size of especially shield modules shape, control of the internal cavities and cooling channels is extremely demanding. This requires strict control of the raw materials and manufacturing parameters

  3. 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

  4. 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.

  5. Design study of a new vacuum vessel for Doublet III

    International Nuclear Information System (INIS)

    Rawls, J.M.; Davis, L.G.; Anderson, P.M.

    1980-10-01

    The principal thrust of the project was to examine a single design in enough depth to gain confidence in the feasibility and desirability of specific design features. However, a valuable spin-off of the project was to develop information of a more generic character to aid in future studies of possibilities for Doublet III. For example, we now feel that Doublet III can be reconfigured with any of a variety of new vacuum vessels, poloidal coil sets, and auxiliary heating systems within three years of project initiation, a period that is short compared to the time scale for developing a completely new facility. In addition, this can be accomplished at a fraction of the cost required to develop a comparable facility

  6. 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)

  7. 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)

  8. 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.

  9. Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel

    Science.gov (United States)

    Maheshwari, A.; Pathak, H. A.; Mehta, B. K.; Phull, G. S.; Laad, R.; Shaikh, M. S.; George, S.; Joshi, K.; Khan, Z.

    2017-04-01

    ITER Vacuum Vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with In-Wall Shielding Blocks (IWS) and Water. The main purpose of IWS is to provide neutron shielding during ITER plasma operation and to reduce ripple of Toroidal Magnetic Field (TF). Although In-Wall Shield Blocks (IWS) will be submerged in water in between the walls of the ITER Vacuum Vessel (VV), Outgassing Rate (OGR) of IWS materials plays a significant role in leak detection of Vacuum Vessel of ITER. Thermal Outgassing Rate of a material critically depends on the Surface Roughness of material. During leak detection process using RGA equipped Leak detector and tracer gas Helium, there will be a spill over of mass 3 and mass 2 to mass 4 which creates a background reading. Helium background will have contribution of Hydrogen too. So it is necessary to ensure the low OGR of Hydrogen. To achieve an effective leak test it is required to obtain a background below 1 × 10-8 mbar 1 s-1 and hence the maximum Outgassing rate of IWS Materials should comply with the maximum Outgassing rate required for hydrogen i.e. 1 x 10-10 mbar 1 s-1 cm-2 at room temperature. As IWS Materials are special materials developed for ITER project, it is necessary to ensure the compliance of Outgassing rate with the requirement. There is a possibility of diffusing the gasses in material at the time of production. So, to validate the production process of materials as well as manufacturing of final product from this material, three coupons of each IWS material have been manufactured with the same technique which is being used in manufacturing of IWS blocks. Manufacturing records of these coupons have been approved by ITER-IO (International Organization). Outgassing rates of these coupons have been measured at room temperature and found in acceptable limit to obtain the required Helium Background. On the basis of these measurements, test reports have been generated and got

  10. Dust processing device for inside of vacuum vessel of thermonuclear reactor

    International Nuclear Information System (INIS)

    Okumura, Atsushi; Tsujimura, Seiichi; Takahashi, Kenji; Ueda, Yasutoshi; Kuwata, Masayasu; Onozuka, Masaki.

    1995-01-01

    The device of the present invention can occasionally recover dusts in a vacuum vessel of a thermonuclear reactor. In addition, fine powdery dusts are never scattered to the vacuum vessel. Namely, a processing device main body comprises a locally sealed space in the vacuum vessel. A blow-up device blows up and floats dusts accumulated in the vacuum vessel to the processing device main body. A discharge plate electrically charges the floating dusts by discharge. An electrode collects the charged dusts. Collected dusts are recovered together with a pressurized gas through a dust recovering port to the outside of the processing device. With such a constitution, it is not necessary to release the vacuum vessel to the atmosphere and evacuate after the completion of the collection of the dusts on every time when the dusts are generated as in the prior art. It is no more necessary for an operator to enter into the vacuum vessel and recover the dusts. Since fine powdery dusts are never scattered in the vacuum vessel, no undesired effects are given to exhaustion facilities and instruments of the vacuum vessel. (I.S.)

  11. Dust processing device for inside of vacuum vessel of thermonuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Atsushi; Tsujimura, Seiichi; Takahashi, Kenji; Ueda, Yasutoshi; Kuwata, Masayasu; Onozuka, Masaki

    1995-05-02

    The device of the present invention can occasionally recover dusts in a vacuum vessel of a thermonuclear reactor. In addition, fine powdery dusts are never scattered to the vacuum vessel. Namely, a processing device main body comprises a locally sealed space in the vacuum vessel. A blow-up device blows up and floats dusts accumulated in the vacuum vessel to the processing device main body. A discharge plate electrically charges the floating dusts by discharge. An electrode collects the charged dusts. Collected dusts are recovered together with a pressurized gas through a dust recovering port to the outside of the processing device. With such a constitution, it is not necessary to release the vacuum vessel to the atmosphere and evacuate after the completion of the collection of the dusts on every time when the dusts are generated as in the prior art. It is no more necessary for an operator to enter into the vacuum vessel and recover the dusts. Since fine powdery dusts are never scattered in the vacuum vessel, no undesired effects are given to exhaustion facilities and instruments of the vacuum vessel. (I.S.).

  12. 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)

  13. 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)

  14. Materials requirements for the ITER vacuum vessel and in-vessel components - approaching the construction phase

    International Nuclear Information System (INIS)

    Barabash, V.; Ioki, K.; Pick, M.; Girard, J.P.; Merola, M.

    2007-01-01

    Full text of publication follows: The ITER activities are fully devoted toward its construction. In accordance with the ITER integrated project schedule, the procurement specifications for the manufacturing of the Vacuum Vessel should be prepared by March 2008 and the procurement specifications for the in-vessel components (first wall/blanket, divertor) by 2009. To update the design, considering design and technology evolution, the ITER Design Review has been launched. Among the various topics being discussed are the important issues related to selection of materials, material procurement, and assessment of performance during operation. The main requirements related to materials for the vacuum vessel and the in-vessel components are summarized in the paper. The specific licensing requirements are to be followed for structural materials of pressure and nuclear pressure equipment components for construction of ITER. In addition, the procurements in ITER will be done mostly 'in-kind' and it is assumed that materials for these components will be produced by different Parties. However, in accordance with the regulatory requirements and quality requirements for operation, common specifications and the general rules to fulfill these requirements are to be adopted. For some ITER components (e.g. first wall, divertor high heat flux components), the ultimate qualification of the joining technologies (Be/Cu, SS/Cu, CFC/Cu, W/Cu) is under final evaluation. Successful accomplishment of the qualification program will allow to proceed with procurements of the components for ITER. The criteria for acceptance of these components and materials after manufacturing are described and the main results will be reported. Additional materials issues, which come from the on-going manufacturing R and D program, will be also described. Finally, further materials activity during the construction phase, needs for final qualification and acceptance of materials are discussed. (authors)

  15. 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.

  16. Plasma discharge in ferritic first wall vacuum vessel of the Hitachi Tokamak HT-2

    International Nuclear Information System (INIS)

    Abe, Mitsushi; Nakayama, Takeshi; Asano, Katsuhiko; Otsuka, Michio

    1997-01-01

    A tokamak discharge with ferritic material first wall was tried successfully. The Hitachi Tokamak HT-2 had a stainless steel SUS304 vacuum vessel and modified to have a ferritic plate first wall for experiments to investigate the possibility of ferritic material usage in magnetic fusion devices. The achieved vacuum pressure and times used for discharge cleaning was roughly identical with the stainless steel first wall or the original HT-2. We concluded that ferritic material vacuum vessel is possible for tokamaks. (author)

  17. Neutral-beam aiming and calorimetry for MFTF-B

    International Nuclear Information System (INIS)

    Goldner, A.I.; Margolies, D.

    1981-01-01

    The vessel for the Tandem Mirror Fusion Test Facility (MFTF-B) will have up to eleven 0.5-s-duration neutral-beam injectors for the initial heating of the MFTF-B plasma. Knowing the exact alignment of the beams and their total power is critical to the performance of the experiment. Using prototype aiming and calorimetry systems on the High Voltage Test Stand (HVTS) at Lawrence Livermore National Laboratory (LLNL), we hope to prove our ability to obtain an aiming accuracy of +-1 cm at the plasma and a calorimetric accuracy of +-5% of the actual total beam energy

  18. 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

  19. 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)

  20. 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.

  1. 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)

  2. 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.

  3. 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)

  4. 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)

  5. Analysis of toroidal vacuum vessels for use in demonstration sized tokamak reactors

    International Nuclear Information System (INIS)

    Culbert, M.E.

    1978-07-01

    The vacuum vessel component of the tokamak fusion reactor is the subject of this study. The main objective of this paper was to provide guidance for the structural design of a thin wall externally pressurized toroidal vacuum vessel. The analyses are based on the available state-of-the-art analytical methods. The shortcomings of these analytical methods necessitated approximations and assumptions to be made throughout the study. A principal result of the study has been the identification of a viable vacuum vessel design for the Demonstration Tokamak Hybrid Reactor (DTHR) and The Next Step (TNS) Reactor

  6. Recovery process of wall condition in KSTAR vacuum vessel after temporal machine-vent for repair

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwang Pyo, E-mail: kpkim@nfri.er.ke; Hong, Suk-Ho; Lee, Hyunmyung; Song, Jae-in; Jung, Nam-Yong; Lee, Kunsu; Chu, Yong; Kim, Hakkun; Park, Kaprai; Oh, Yeong-Kook

    2015-10-15

    Highlights: • Efforts have been made to obtain vacuum condition that is essential for the plasma experiments. • For example, the vacuum vessel should be vented to repair in-vessel components such as diagnostic shutter, and PFC damaged by high energy plasma. • Here, we present the recovery process of wall condition in KSTAR after temporal machine-vent for repair. • It is found that an acceptable vacuum condition has been achieved only by plasma based wall conditioning techniques such as baking, GDC, and boronization. • This study was that the proper recovering method of the vacuum condition should be developed according to the severity of the accident. - Abstract: Efforts have been made to obtain vacuum condition that is essential for the plasma experiments. Under certain situations, for example, the vacuum vessel should be vented to repair in-vessel components such as diagnostic shutter, exchange of window for diagnostic equipment, and PFC damaged by high energy plasma. For the quick restart of the campaign, a recovery process was established to make the vacuum condition acceptable for the plasma experiment. In this paper, we present the recovery process of wall condition in KSTAR after temporal machine-vent for repair. It is found that an acceptable vacuum condition has been achieved only by plasma based wall conditioning techniques such as baking, GDC, and boronization. This study was that the proper recovering method of the vacuum condition should be developed according to the severity of the accident.

  7. Design description of the vacuum vessel for the Advanced Toroidal Facility

    International Nuclear Information System (INIS)

    Chipley, K.K.; Nelson, B.E.; Vinyard, L.M.; Williamson, D.F.

    1983-01-01

    The Advanced Toroidal Facility (ATF) will be a stellarator experiment to investigate improvements in toroidal confinement. The vacuum vessel for this facility will provide the appropriate evacuated region for plasma containment within the helical field (HF) coils. The vessel is designed to provide the maximum reasonable volume inside the HF coils and to provide the maximum reasonable access for future diagnostics. The vacuum vessel design is at an early phase and all of the details have not been completed. The heat transfer analysis and stress analysis completed during the conceptual design indicate that the vessel will not change drastically

  8. Design features of the A-cell and transition coils of MFTF-B

    International Nuclear Information System (INIS)

    Tatro, R.E.; Wohlwend, J.W.; Ring, D.S.

    1981-01-01

    The MFTF-B transition coil and A-cell magnet designs use variations of the copper-stabilized NbTi conductor developed by LLNL for the MFTF Yin-Yang magnets. This conductor will be wound on the one inch thick (12.7 mm) stainless steel coil forms using a two-axis winding machine similar to the existing LLNL Yin-Yang winding machine. After winding, covers will be placed over the coil and welded to the coil form to form a helium-tight jacket around the conductor. These jacketed coils are then enclosed in thick structural cases that react the large Lorentz forces on the magnets. The space between the coil jacket and case will be filled by a stainless steel bladder that will be injected with urethane. The injection bladder will provide cooling passages during cooldown as well as transmitting the Lorentz forces between the jacket and the case. The large self-equilibrating lobe-spreading forces on the magnets (29.10 6 lb, 127.0 MN) for the A-cell are reacted primarily through the thick 304 LN case into the external superstructure. The net Lorentz forces and the inertial forces on the magnet are reacted through support systems into the LLNL vacuum vessel structure

  9. A spheromak ignition experiment reusing Mirror Fusion Test Facility (MFTF) equipment

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1993-01-01

    Based on available experimental results and theory, a scenario is presented to achieve ohmic ignition in a spheromak by slow (∼ 10 sec.) helicity injection using power from the Mirror Fusion Test Facility (MFTF) substation. Some of the other parts needed (vacuum vessel, coils, power supplies, pumps, shielded building space) might also be obtained from MFTF or other salvage, as well as some components needed for intermediate experiments for additional verification of the concept (especially confinement scaling). The proposed ignition experiment would serve as proof-of-principle for the spheromak DT fusion reactor design published by Hagenson and Krakowski, with a nuclear island cost about ten times less than a tokamak of comparable power. Designs at even higher power density and lower cost might be possible using Christofilos' concept of a liquid lithium blanket. Since all structures would be protected from neutrons by the lithium blanket and the tritium inventory can be reduced by continuous removal from the liquid blanket, environmental and safety characteristics appear to be favorable

  10. MFTF magnet cryostability

    International Nuclear Information System (INIS)

    VanSant, J.H.

    1979-01-01

    A pair of large superconducting magnets will be installed in the Mirror Fusion Test Facility (MFTF), which is to begin operation in 1981. To ensure a stable superconducting state for the niobium-titanium (Nb-Ti) conductor, special consideration has been given to certain aspects of the magnet system design. These include the conductor, joints, coil assembly, vapor plenums, liquid-helium (LHe) supply system, and current leads. Heat transfer is the main consideration; i.e., the helium quality and temperature are limited so that the superconductor will perform satisfactorily in the magnet environment

  11. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

    2000-01-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

  12. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)

    2000-11-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.

  13. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

    2000-07-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

  14. MFTF-progress and promise

    International Nuclear Information System (INIS)

    Thomassen, K.I.

    1980-01-01

    The Mirror Fusion Test Facility (MFTF) has been in construction at Lawrence Livermore National Laboratory (LLNL) for 3 years, and most of the major subsystems are nearing completion. Recently, the scope of this project was expanded to meet new objectives, principally to reach plasma conditions corresponding to energy break-even. To fulfill this promise, the single-cell minimum-B mirror configuration will be replaced with a tandem mirror configuration (MFTF-B). The facility must accordingly be expanded to accomodate the new geometry. This paper briefly discusses the status of the major MFTF subsystems and describes how most of the technological objectives of MFTF will be demonstrated before we install the additional systems necessary to make the tandem. It also summarizes the major features of the expanded facility

  15. The vacuum vessel for the FTU device: design constraints and stress analysis

    International Nuclear Information System (INIS)

    Andreani, R.; Cecchini, A.; Gasparotto, M.; Lovisetto, L.; Migliori, S.; Pizzuto, A.

    1984-01-01

    The FTU vacuum vessel must withstand large electromagnetic loads due to the interactions between the eddy currents in the vessel and high magnetic fields of the machine, the atmospheric pressure and the severe thermal loads due to plasma losses and RF power not coupled to the plasma. In order to minimise the stresses on the vacuum chamber, an optimization of the wall thickness has been performed and, in order to assess the feasibility of the vessel, an extensive three dimensional finite element stress analysis has been developed. The main results obtained are illustrated. (author)

  16. Eddy currents in a nonperiodic vacuum vessel induced by axisymmetric plasma motion

    International Nuclear Information System (INIS)

    DeLucia, J.

    1985-12-01

    A method is described for calculating the two-dimensional trajectory of a vertically or horizontally unstable axisymmetric tokamak plasma in the presence of a resistive vacuum vessel. The vessel is not assumed to have toroidal symmetry. The plasma is represented by a current-filament loop that is free to move vertically and to change its major radius. Its position is evolved in time self-consistently with the vacuum vessel eddy currents. The plasma current, internal inductance, and poloidal beta can be specified functions of time so that eddy currents resulting from a disruption can be modeled. The vacuum vessel is represented by a set of current-filaments whose positions and orientations are chosen to model the dominant eddy current paths. Although the specific application is to TFTR, the present model is of general applicability. 7 refs., 4 figs., 2 tabs

  17. Electromagnetic forces on a metallic Tokamak vacuum vessel following a disruptive instability

    International Nuclear Information System (INIS)

    Eckhartt, D.

    1979-04-01

    During a 'hard' disruptive instability of a Tokamak plasma the current-carrying plasma is lost within a very short time, typically few milliseconds. If the plasma is contained in a metallic vacuum vessel, electric currents are set up in the vessel following the disappearance of the plasma current. These vessel currents together with the magnetic fields intersecting the vessel generate electromagnetic forces which appear as mechanical loads on the vessel. In the following note it is assumed that the vacuum vessel is surrounded by an 'outer equivalent' or 'flux-conserving' shell having a characteristic time of magnetic field penetration which is long compared to the time of existence of the vessel currents. This property defines the distribution of vessel current densities (and hence the load distribution) without referring to the exact mechanism or time sequence of events by which the plasma current is lost. Numerical examples of the electromagnetic force distribution from this model refer to parameters of the JET-device with the simplifying assumption of circular cross-sections for plasma current, vacuum vessel, and outer equivalent shell. (orig.)

  18. MFTF-B plasma-diagnostic system

    International Nuclear Information System (INIS)

    Throop, A.L.; Goerz, D.A.; Thomas, S.R.

    1981-01-01

    This paper describes the current design status of the plasma diagnostic system for MFTF-B. In this paper we describe the system requirement changes which have occurred as a result of the funded rescoping of the original MFTF facility into MFTF-B. We outline the diagnostic instruments which are currently planned, and present an overview of the diagnostic system

  19. MFTF exception handling system

    International Nuclear Information System (INIS)

    Nowell, D.M.; Bridgeman, G.D.

    1979-01-01

    In the design of large experimental control systems, a major concern is ensuring that operators are quickly alerted to emergency or other exceptional conditions and that they are provided with sufficient information to respond adequately. This paper describes how the MFTF exception handling system satisfies these requirements. Conceptually exceptions are divided into one of two classes. Those which affect command status by producing an abort or suspend condition and those which fall into a softer notification category of report only or operator acknowledgement requirement. Additionally, an operator may choose to accept an exception condition as operational, or turn off monitoring for sensors determined to be malfunctioning. Control panels and displays used in operator response to exceptions are described

  20. Structural analysis of the ITER Vacuum Vessel regarding 2012 ITER Project-Level Loads

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Jun, C.H.; Portafaix, C.; Choi, C.-H.; Ioki, K.; Sannazzaro, G.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Cambazar, M.; Corti, Ph.; Pinori, K.; Sfarni, S.; Tailhardat, O. [Assystem EOS, 117 rue Jacquard, L' Atrium, 84120 Pertuis (France); Borrelly, S. [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Albin, V.; Pelletier, N. [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France)

    2014-10-15

    Highlights: • ITER Vacuum Vessel is a part of the first barrier to confine the plasma. • ITER Vacuum Vessel as Nuclear Pressure Equipment (NPE) necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, conformance testing and quality assurance, i.e. Agreed Notified Body (ANB). • A revision of the ITER Project-Level Load Specification was implemented in April 2012. • ITER Vacuum Vessel Loads (seismic, pressure, thermal and electromagnetic loads) were summarized. • ITER Vacuum Vessel Structural Margins with regards to RCC-MR code were summarized. - Abstract: A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (LS) on the ITER Vacuum Vessel (VV) loads and the main structural margins required by the applicable French code, RCC-MR.

  1. Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

    International Nuclear Information System (INIS)

    Mozzillo, Rocco; Tarallo, Andrea; Marzullo, Domenico; Bachmann, Christian; Di Gironimo, Giuseppe; Mazzone, Giuseppe

    2016-01-01

    Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

  2. Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

    Energy Technology Data Exchange (ETDEWEB)

    Mozzillo, Rocco, E-mail: rocco.mozzillo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Tarallo, Andrea; Marzullo, Domenico [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Mazzone, Giuseppe [Unità Tecnica Fusione - ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

    2016-11-15

    Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

  3. Design of the Intersector Welding Robot for vacuum vessel assembly and maintenance

    International Nuclear Information System (INIS)

    Jones, L.; Dagenais, J.-F.; Daenner, W.; Maisonnier, D.

    2000-01-01

    Next Step Fusion Devices require on-site (field weld) joining of sectors of the thick-walled vacuum vessel for structural and vacuum integrity. EFDA (European Fusion Development Agreement) is supporting an R and D programme to investigate processes for assembly of the vacuum vessel and to carry out cutting, re-welding and inspection for remote sector replacement, forming part of the overall VV/blanket research effort. In order to direct the process end-effectors along the field joint zone, a track-mounted Intersector Welding Robot (IWR) on a mock-up of a region of the vacuum vessel has been designed and is described in this paper. A rail-mounted hexapod type robot offers six axes of motion over a limited work envelope with high payload to robot weight ratio. A solution to the production of reduced pressure local vacuum is the installation of short, lightweight segments bolted to each other and the vessel wall. The various process heads can be mounted using end-effectors of special design. To minimise the supply and interface problems for the IWR prototype, its motion control and electronic systems will be embedded locally. A laser scan with camera forms the on-line seam tracking capability to compensate for rail and seam deviations

  4. Evaluation of structural reliability for vacuum vessel under external pressure and electromagnetic force

    International Nuclear Information System (INIS)

    Minato, Akio

    1983-08-01

    Static and dynamic structural analyses of the vacuum vessel for a Swimming Pool Type Tokamak Reactor (SPTR) have been conducted under the external pressure (hydraulic and atmospheric pressure) during normal operation or the electromagnetic force due to plasma disruption. The reactor structural design is based on the concept that the adjacent modules of the vacuum vessel are not connected mechanically with bolts in the torus inboard region each other, so as to save the required space for inserting the remote handling machine for tightenning and untightenning bolts in the region and to simplify the repair and maintenance of the reactor. The structural analyses of the vacuum vessel have been carried out under the external pressure and the electromagnetic force and the structural reliability against the static and dynamic loads is estimated. The several configurations of the lip seal between the modules, which is required to make a plasma vacuum boundary, have been proposed and the structural strength under the forced displacements due to the deformation of the vacuum vessel is also estimated. (author)

  5. Fabrication of the vacuum vessel of the Spanish stellarator TJ-II

    International Nuclear Information System (INIS)

    Botija, J.; Blaumoser, M.; Cal, E. de la; Garcia, A.; Tabares, F.; Molleta, L.; Rigadello, D.; Dal Maso, S.; Bevilacqua, G.

    1995-01-01

    TJ-II is a medium size stellarator under construction in Madrid, Spain. Its major plasma radius is 1.5 m and its minor plasma dimensions are 0.2m by 0.4m. The toroidal magnetic field on the axis is 1T. The bean shaped helical plasma is contained in a stainless steel vacuum vessel with a total of 96 ports, including 8 manholes to have access to its interior. The vacuum vessel will be baked at 150 C. Its complicated geometry along with the high tolerance requirements make this component a difficult manufacturing challenge. (orig.)

  6. Integration of cooking and vacuum cooling of carrots in a same vessel

    Directory of Open Access Journals (Sweden)

    Luiz Gustavo Gonçalves Rodrigues

    2012-03-01

    Full Text Available Cooked vegetables are commonly used in the preparation of ready-to-eat foods. The integration of cooking and cooling of carrots and vacuum cooling in a single vessel is described in this paper. The combination of different methods of cooking and vacuum cooling was investigated. Integrated processes of cooking and vacuum cooling in a same vessel enabled obtaining cooked and cooled carrots at the final temperature of 10 ºC, which is adequate for preparing ready-to-eat foods safely. When cooking and cooling steps were performed with the samples immersed in boiling water, the effective weight loss was approximately 3.6%. When the cooking step was performed with the samples in boiling water or steamed, and the vacuum cooling was applied after draining the boiling water, water loss ranged between 15 and 20%, which caused changes in the product texture. This problem can be solved with rehydration using a small amount of sterile cold water. The instrumental textural properties of carrots samples rehydrated at both vacuum and atmospheric conditions were very similar. Therefore, the integrated process of cooking and vacuum cooling of carrots in a single vessel is a feasible alternative for processing such kind of foods.

  7. Structural analysis and manufacture for the vacuum vessel of experimental advanced superconducting tokamak (EAST) device

    International Nuclear Information System (INIS)

    Song Yuntao; Yao Damao; Wu Songata; Weng Peide

    2006-01-01

    The experimental advanced superconducting tokamak (EAST) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and is being constructed as the Chinese national nuclear fusion research project. The vacuum vessel, that is one of the key components, will have to withstand not only the electromagnetic force due to the plasma disruption and the Halo current, but also the pressure of boride water and the thermal stress due to the 250 deg. C baking out by the hot pressure nitrogen gas, or the 100 deg. C hot wall during plasma operation. This paper is a report of the mechanical analyses of the vacuum vessel. According to the allowable stress criteria of American Society of Mechanical Engineers, Boiler and Pressure Vessel Committee (ASME), the maximum integrated stress intensity on the vacuum vessel is 396 MPa, less than the allowable design stress intensity 3S m (441 MPa). At the same time, some key R and D issues are presented, which include supporting system, bellows and the assembly of the whole vacuum vessel

  8. Design, fabrication and test of double-wall vacuum vessel for JT-60U

    International Nuclear Information System (INIS)

    Uchikawa, Takashi; Ioki, Kimihiro; Ninomiya, Hiromasa.

    1994-01-01

    A double-wall vacuum vessel was designed and fabricated for JT-60U (an upgraded machine of JT-60), which has a plasma current up to 6 MA and a large plasma volume (100 m 3 ). A new concept of Inconel 625 all-welded structure was adopted to the vessel, that comprises an inner plate, square tubes and an outer plate. The vacuum vessel with a multi-arc D-shaped cross section was fabricated by using hot-sizing press. The electromagnetic and structural analysis has been performed for plasma disruption loads. Dynamic responses of the vessel were measured during plasma disruptions, and the observed displacement had a good agreement with the result of FEM analysis. (author)

  9. 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

  10. Modeling and measurement of the motion of the DIII-D vacuum vessel during vertical instabilities

    International Nuclear Information System (INIS)

    Reis, E.; Blevins, R.D.; Jensen, T.H.; Luxon, J.L.; Petersen, P.I.; Strait, E.J.

    1991-11-01

    The motions of the D3-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which were input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hertz that are sufficiently loud to be felt as well as heard by the D3-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds. The correlation of the theoretical and measured vessel currents, the dynamic measurements and analysis, and the acoustic measurements and analysis show that: (1) The physics model can predict vessel forces for selected values of plasma resistivity. The model also predicts poloidal and toroidal wall currents which agree with measured values; (2) The force-time history from the above model, used in conjunction with an axisymmetric structural model of the vessel, predicts vessel motions which agree well with measured values; (3) The above results, input to a simple acoustic model predicts the magnitude of sounds emitted from the vessel during disruptions which agree with acoustic measurements; (4) Correlation of measured vessel motions with structural analysis shows that a maximum vertical motion of the vessel up to 0.24 in will not overstress the vessel or its supports. 11 refs., 10 figs., 1 tab

  11. Neutronics studies for the design of the European DEMO vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Flammini, Davide, E-mail: davide.flammini@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Moro, Fabio; Pizzuto, Aldo [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Bachmann, Christian [EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching (Germany)

    2016-11-01

    Highlights: • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial plane. • Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. • Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. • Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. • Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel. - Abstract: The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm{sup 3}), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided.

  12. Temperature field and thermal stress analysis of the HT-7U vacuum vessel

    International Nuclear Information System (INIS)

    Song Yuntao; Yao Damao; Wu Songtao; Weng Peide

    2000-01-01

    The HT-7U vacuum vessel is an all-metal-welded double-wall interconnected with toroidal and poloidal stiffening ribs. The channels formed between the ribs and walls are filled with boride water as a nuclear shielding. On the vessel surface facing the plasma are installed cable-based Ohmic heaters. Prior to plasma operation the vessel is to be baked out and discharge cleaned at about 250 degree C. During baking out the non-uniformity of temperature distribution on the vacuum vessel will bring about serious thermal stress that can damage the vessel. In order to determine and optimize the design of the HT-7U vacuum vessel, a three-dimensional finite element model was performed to analyse its temperature field and thermal stress. the maximal thermal stress appeared on the round of lower vertical port and maximal deformation located just on the region between the upper vertical port and the horizontal port. The results show that the reinforced structure has a good capability of withstanding the thermal loads

  13. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    Science.gov (United States)

    Kumar, B. Ramesh; Gangradey, R.

    2012-11-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  14. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    International Nuclear Information System (INIS)

    Kumar, B Ramesh; Gangradey, R

    2012-01-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  15. Thermal structural analysis of SST-1 vacuum vessel and cryostat assembly using ANSYS

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Bedakihale, Vijay; Ranganath, Tata

    2009-01-01

    Steady state super-conducting tokamak-1 (SST-1) is a medium sized tokamak, which has been designed to produce a 'D' shaped double null divertor plasma and operate in quasi steady state (1000 s). SST-1 vacuum system comprises of plasma chamber (vacuum vessel, interconnecting rings, baking and cooling channels), and cryostat all made of SS 304L material designed to meet ultra high vacuum requirements for plasma generation and confinement. Prior to plasma shot and operation the vessel assembly is baked to 250/150 deg. C from room temperature and discharge cleaned to remove impurities/trapped gases from wall surfaces. Due to baking the non-uniform temperature pattern on the vessel assembly coupled with atmospheric pressure loading and self-weight give rise to high thermal-structural stresses, which needs to be analyzed in detail. In addition the vessel assembly being a thin shell vessel structure needs to be checked for critical buckling load caused by atmospheric and baking thermal loads. Considering symmetry of SST-1, 1/16th of the geometry is modeled for finite element (FE) analysis using ANSYS for different loading scenarios, e.g. self-weight, pressure loading considering normal operating conditions, and off-normal loads coupled with baking of vacuum vessel from room temperature 250 deg. C to 150 deg. C, buckling and modal analysis for future dynamic analysis. The paper will discuss details about SST-1 vacuum system/cryostat, solid and FE model of SST-1, different loading scenarios, material details and the stress codes used. We will also present the thermal structural results of FE analysis using ANSYS for various load cases being investigated and our observations under different loading conditions.

  16. Development of remote welding equipment and techniques for the TFTR vacuum vessel

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Larson, R.A.; Aldrich, W.C.

    1980-01-01

    In the event that the TFTR vacuum vessel is damaged or one of the toroidal field coils fails after the system has become substantially activated, it is necessary to remotely remove and replace the damaged section of the vessel using remote handling procedures. This paper describes a welding system developed through the final design stage to perform the remote welding necessary during the replacement operation. Information is presented describing the vessel configuration, the replacement sequence, the welding system requirements, welder configuration, supporting systems, the weld development program and future development requirements

  17. Calculation of voltages and currents induced in the vacuum vessel of ASDEX by plasma disruptions

    International Nuclear Information System (INIS)

    Preis, H.

    1978-01-01

    An approximation method is used to analyze the electromagnetic diffusion process induced in the walls of the ASDEX vacuum vessel by plasma disruptions. For this purpose the rotational-symmetric vessel is regarded as N = 82 circular conductors connected in parallel and inductively coupled with one another and with the plasma. The transient currents and voltages occurring in this circuit are calculated with computer programs. From the calculated currents it is possible to determine the time behavior of the distributions of the current density and magnetic force density in the vessel walls. (orig.) [de

  18. 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.

  19. 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.

  20. Modeling the Thermal Mechanical Behavior of a 300 K Vacuum Vessel that is Cooled by Liquid Hydrogen in Film Boiling

    International Nuclear Information System (INIS)

    Yang, S.Q.; Green, M.A.; Lau, W.

    2004-01-01

    This report discusses the results from the rupture of a thin window that is part of a 20-liter liquid hydrogen vessel. This rupture will spill liquid hydrogen onto the walls and bottom of a 300 K cylindrical vacuum vessel. The spilled hydrogen goes into film boiling, which removes the thermal energy from the vacuum vessel wall. This report analyzes the transient heat transfer in the vessel and calculates the thermal deflection and stress that will result from the boiling liquid in contact with the vessel walls. This analysis was applied to aluminum and stainless steel vessels

  1. Design and fabrication of the vacuum vessel for the Advanced Toroidal Facility

    International Nuclear Information System (INIS)

    Chipley, K.K.; Frey, G.N.

    1985-01-01

    The vacuum vessel for the Advanced Toroidal Facility (ATF) is a heavily contoured and very complex formed vessel that is specifically designed to allow for maximum plasma volume in a pure stellarator arrangement. The design of the facility incorporates an internal vessel that is closely fitted to the two helical field coils following the winding law theta = 1/6phi. Metallic seals have been incorporated throughout the system to minimize impurities. The vessel has been fabricated utilizing a comprehensive set of tooling fixtures specifically designed for the task of forming 6-mm stainless steel plate to the complex shape. Computer programs were used to develop a series of ribs that essentially form an internal mold of the vessel. Plates were press-formed with multiple compound curves, fitted to the fixture, and joined with full-penetration welds. 7 refs., 8 figs

  2. Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

    International Nuclear Information System (INIS)

    Joshi, Jaydeep; Yadav, Ashish; Gangadharan, Roopesh; Prasad, Rambilas; Ulahannan, Shino; Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun

    2015-01-01

    Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

  3. Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jaydeep, E-mail: Jaydeep.joshi@iter-india.org [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Yadav, Ashish; Gangadharan, Roopesh [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Prasad, Rambilas [Madan Mohan Malaviya University of Technology, Gorakhpur, Uttar Pradesh 273001 (India); Ulahannan, Shino [Airframe Aerodesigns Pvt. Ltd., HAL Airport Exit Road, Old Airport Road, Bengaluru 17 (India); Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India)

    2015-10-15

    Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

  4. 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.

  5. 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

  6. Electromagnetic loads and structural response of the CIT [Compact Ignition Tokamak] vacuum vessel to plasma disruptions

    International Nuclear Information System (INIS)

    Salem, S.L.; Listvinsky, G.; Lee, M.Y.; Bailey, C.

    1987-01-01

    Studies of the electromagnetic loads produced by a variety of plasma disruptions, and the resulting structural effects on the compact Ignition Tokamak (CIT) vacuum vessel (VV), have been performed to help optimize the VV design. A series of stationary and moving plasmas, with disruption rates from 0.7--10.0 MA/ms, have been analyzed using the EMPRES code to compute eddy currents and electromagnetic pressures, and the NASTRAN code to evaluate the structural response of the vacuum vessel. Key factors contributing to the magnitude of EM forces and resulting stresses on the vessel have been found to include disruption rate, and direction and synchronization of plasma motion with the onset of plasma current decay. As a result of these analyses, a number of design changes have been made, and design margins for the present 1.75 meter design have been improved over the original CIT configuration. 1 ref., 10 figs., 4 tabs

  7. Mechanical strength evaluation of the welded bellows for the ports of the JT-60 vacuum vessel

    International Nuclear Information System (INIS)

    Takatso, H.; Shimizu, M.; Yamamoto, M.

    1983-01-01

    Mechanical strength of the welded bellows for the ports of the JT-60 vacuum vessel was evaluated, laying the emphasis on the fatigue strength under the torsional electromagnetic force. The welded bellows were designed to be loaded with the forced deflection due to the relative displacement between the vacuum vessel and the external fixed point, the atmospheric pressure and the forced torsional angle due to the electromagnetic force. Stresses caused by the former two were estimated following the formulae proposed by the Kellogg Company. On the other hand, two formulae were established to estimate the stress caused by the last, after examining experimentally the behavior of the welded bellows under the torsional load; one is the shearing stress evaluation formula and the other is the axial bending stress evaluation formula. It was found that the welded bellows can easily buckle under the torsional load and the former formula corresponds to the case of non-buckling and the latter to the case of buckling. The present mechanical strength evaluation method was applied to the three kinds of the welded bellows to be used in the ports of the JT-60 vacuum vessel (neutral beam injection ports, vacuum pumping ports and the adjustable limiter ports) and it was confirmed that they have sufficient strength in the range of the design load conditions

  8. 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

  9. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    Energy Technology Data Exchange (ETDEWEB)

    Houry, M., E-mail: Michael.houry@cea.fr [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H. [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Kammerer, N.; Measson, Y. [CEA, LIST, F-92265 Fontenay-aux-Roses (France); Carrel, F.; Schoepff, V. [CEA, LIST, F-91191 Gif-sur-Yvette (France)

    2011-10-15

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  10. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    International Nuclear Information System (INIS)

    Houry, M.; Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H.; Kammerer, N.; Measson, Y.; Carrel, F.; Schoepff, V.

    2011-01-01

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  11. Mirror Fusion Test Facility (MFTF)

    International Nuclear Information System (INIS)

    Thomassen, K.I.

    1978-01-01

    A large, new Mirror Fusion Test Facility is under construction at LLL. Begun in FY78 it will be completed at the end of FY78 at a cost of $94.2M. This facility gives the mirror program the flexibility to explore mirror confinement principles at a signficant scale and advances the technology of large reactor-like devices. The role of MFTF in the LLL program is described here

  12. Design and Structural Analysis for the Vacuum Vessel of Superconducting Tokamak JT-60SC

    International Nuclear Information System (INIS)

    Kudo, Y.; Sakurai, S.; Masaki, K.; Urata, K.; Sasajima, T.; Matsukawa, M.; Sakasai, A.; Ishida, S.

    2003-01-01

    A modification of the JT-60 is planned to be a superconducting tokamak (JT-60SC) in order to establish steady-state operation of high beta plasma for 100 s, and to ensure the applicability of ferritic steel as a reduced activation material for reactor relevant break-even class plasmas. This paper describes the detailed design of the vacuum vessel, which has a unique structure for cost effective manufacturing, as well as structural analysis results for a feasibility study

  13. Studies on structural analysis related to the design of the JT-60 vacuum vessel

    International Nuclear Information System (INIS)

    Takatsu, Hideyuki

    1987-06-01

    Studies on structural analysis of a vacuum vessel of tokamak-type fusion devices are presented. The present studies are proposals for the structural analysis procedures of the tokamak-type fusion devices and are composed of five parts, each of which covers the fundamental area required for the structural analysis and design; stress analysis, dynamic response analysis, fatigue evaluation, buckling analysis and seismic analysis. Special attention is paid to the critical component, bellows and the critical load, electromagnetic forces. A new finite element method modeling technique is proposed for the stress analysis of U-shaped bellows, where the bellows is replaced by an orthotropic plate having the same stiffness as the bellows. The applicability of the present modeling technique is confirmed by verification tests. Dynamic response and fatigue of the vacuum vessel are critical issues of the structural analysis and design of the tokamak-type fusion devices. Detailed dynamic response analyses of the JT-60 vacuum vessel are presented paying special attention to the dynamic behavior of the U-shaped bellows, where the above-mentioned modeling technique of the U-shaped bellows is applied. A fatigue evaluation method of the vacuum vessel under the dynamic electromagnetic forces is proposed, which utilizes the results of the detailed dynamic response analysis. In the present method, fatigue evaluation method for random loads is applied. Torsional fatigue strength of the welded bellows is experimentally evaluated aiming the application to the port of the fusion device and it is shown that the welded bellows reveals elastic buckling and spiral distortion under a small angle of tortion. Two formulae are proposed to evaluate the stress of the welded bellows under the forced angle of tortion. (author)

  14. 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

  15. Blanket and vacuum vessel design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    Iida, H.; Minato, A.; Kitamura, K.

    1983-01-01

    The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li 2 O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

  16. Dynamic response of the JT-60 vacuum vessel under the electromagnetic forces

    International Nuclear Information System (INIS)

    Takatsu, H.; Shimizu, M.; Ohta, M.

    1982-01-01

    Dynamic response analyses of the JAERI Tokamak 60 (JT-60) vacuum vessel were carried out under three kinds of saddle-like electromagnetic forces. In the analysis, the dynamic response of the bellows was obtained by dividing it into three components; the first, caused by the forced deflection due to the displacement of an adjacent rigid ring; the second, caused by inertia force; and the third, caused by a saddle-like electromagnetic force. Eigenvalue analyses showed that the 20th mode is a typical rotation mode of the rigid ring around the major radius with a natural frequency of 46.3 Hz. From the results of the dynamic response analyses, the maximum displacement response of the rigid ring was 3.1 mm and remarkable dynamic response was observed in the case of plasma disruption with a time constant of 1 ms. In cases of start-up of the plasma current and plasma disruption with a time constant of 50 ms, the rigid ring vibrates quasi-statically. It is clear that the dynamic behavior of the vacuum vessel is governed mainly by the saddle-like electromagnetic force, with a smaller effect of the inverse saddle-like electromagnetic force on the dynamic response of the vacuum vessel. (orig.)

  17. Experimental and analytical investigations to air and steam ingress into the vacuum vessel of fusion reactors

    International Nuclear Information System (INIS)

    Kruessenberg, A.K.

    1996-12-01

    The basic fusion safety objective is the development of fusion power plants with features that protect individuals, society and the environment by establishing and maintaining an effective defence against radiological and other hazards. The most important specific principle is the establishment of three sequential levels of defence, characterized in priority order by prevention, protection and mitigation. The safety conscious selection of materials as one prevention feature gives the basis for the work described in this report. In order to protect the metallic first wall of fusion reactors from direct interaction with the plasma an extra armour is foreseen. Carbon offers the features low atomic number, high melting point, high thermal conductivity and good mechanical stability up to high temperatures making it to a favourite armour material. Looking on the safety behaviour of fusion reactors it has to be noted that carbon is unstable against oxidizing media like oxygen and steam at high temperatures und carbon has a high sorption capacity for radiologically important tritium. And tritium used as intermediate fuel in the actual reactor concepts is the one form radioactivity is present in fusion reactors. Accidents like loss of vacuum (LOVA) will lead to an air ingress into the vacuum vessel, oxidation of the hot carbon and a partial mobilization of the sorbed tritium. In a similar manner loss of coolant into vacuum (LOCIV) will lead to a water/steam ingress into the vacuum vessel, also accompanied by carbon oxidation and tritium release. (orig.)

  18. Temperature distributions in a Tokamak vacuum vessel of fusion reactor after the loss-of-vacuum-events occurred

    International Nuclear Information System (INIS)

    Takase, K.; Kunugi, T.; Shibata, M.; Seki, Y.

    1998-01-01

    If a loss-of-vacuum-event (LOVA) occurred in a fusion reactor, buoyancy-driven exchange flows would occur at breaches of a vacuum vessel (VV) due to the temperature difference between the inside and outside of the VV. The exchange flows may bring mixtures of activated materials and tritium in the VV to the outside through the breaches, and remove decay heat from the plasma-facing components of the VV. Therefore, the LOVA experiments were carried out under the condition that one or two breaches was opened and that the VV was heated to a maximum 200 C, using a small-scaled LOVA experimental apparatus. Air and helium gas were provided as working fluids. Fluid and wall temperature distributions in the VV were measured and the flow patterns in the VV were estimated by using these temperature distributions. It was found that: (1) the exchange mass in the VV depended on the breach positions; (2) the exchange flow at the single breach case became a counter-current flow when the breach was at the roof of the VV and a stratified flow when it was at the side wall; (3) and that at the double breach case, a one-way flow between two breaches was formed. (orig.)

  19. Lubricant coating of dowel for the ITER vacuum vessel gravity support

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.Y. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Ahn, H.J., E-mail: hjahn@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bak, J.S. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Choi, C.H.; Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Zauner, C. [KRP-Mechatec Engineering GbR, 85748 Garching b, Muenchen (Germany)

    2012-08-15

    The ITER vacuum vessel gravity supports located in the lower level shall sustain loads in radial, toroidal and vertical directions. The hinge type VVGS consists of two hinges, upper and lower blocks and dowels. In order to develop the design concept and verify the structural integrity of the hinge system, the design analysis has been performed in detail. Inclination of 15 Degree-Sign for the hinge based supporting system was introduced to provide centering force to make stable equilibrium state of the vacuum vessel. Due to this inclination the hinges are rotated by the radial expansion of the VV during operation and baking, respectively. If a dowel is seized in the hinge, the supporting system can be highly stressed due to the restrained displacement in the seized dowel. Therefore, solid lubricant coatings were suggested on dowels in order to avoid seizing in the sliding area. In this work, several sets of coupons were made with different coating materials to investigate the effect according to the selection of coating material. Also, a test facility was designed to cover the ITER relevant loading and boundary conditions, e.g. vacuum condition, temperature, contact pressure, cycles, etc. From those test results, the optimized coating method was found to avoid seizure of dowel in the ITER VVGS.

  20. Design and performance tests of gas circulation heating of JT-60U vacuum vessel

    International Nuclear Information System (INIS)

    Yotsuga, M.; Masuzaki, T.; Sago, H.; Nishikane, M.; Uchikawa, T.; Iritani, Y.; Murakami, T.; Horiike, H.; Neyatani, Y.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Miyachi, I.

    1992-01-01

    This paper reports that in the final stage of construction of the upgraded JT-60 device (JT-60U), baking tests of the vacuum vessel was performed. The vessel torus was heated-up to 300 degrees C by means of the nitrogen gas circulation system and electric heaters mounted on the outboard solid wall of the vessel. The design of the gas flow channels inside the double-wall structure of the vessel was done based on flow model tests, fluid analysis, and flow network analysis. The results of the baking tests were satisfactory. In maintaining 300 degrees C bake-out temperature, required heating power of the gas circulation system and outboard heaters was 520kW and 50kW, respectively. The temperature distribution over the vessel wall was within 300 ± 30 degrees C. It was also shown or suggested that heat-up and cool-down time is about 30 hours. The baking tests data have been reflected on operations for plasma experiments

  1. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  2. Analysis on ingress of coolant event in vacuum vessel using modified TRAC-BF1 code

    International Nuclear Information System (INIS)

    Ajima, Toshio; Kurihara, Ryoichi; Seki, Yasushi

    1999-08-01

    The Transient Reactor Analysis Code (TRAC-BF1) was modified on the basis of ICE experimental results so as to analyze the Ingress of Coolant Event (ICE) in the vacuum vessel of a nuclear fusion reactor. In the previous report, the TRAC-BF1 code, which was originally developed for the safety analysis of a light water reactor, had been modified for the ICE of the fusion reactor. And the addition of the flat structural plate model to the VESSEL component and arbitrary appointment of the gravity direction had been added in the TRAC-BF1 code. This TRAC-BF1 code was further modified. The flat structural plate model of the VESSEL component was enabled to divide in multi layers having different materials, and a part of the multi layers could take a buried heater into consideration. Moreover, the TRAC-BF1 code was modified to analyze under the low-pressure condition close to vacuum within range of the steam table. This paper describes additional functions of the modified TRAC-BF1 code, analytical evaluation using ICE experimental data and the ITER model with final design report (FDR) data. (author)

  3. Aluminium vacuum vessel/first surface conceptual design for a commercial tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Culbert, M.

    1981-01-01

    The purpose of this investigation was to develop a design concept for a commercial tokamak hybrid reactor (CTHR) vacuum vessel/first surface system which satisfies the engineering requirements for a commercial environment. An important distinction between the design constraints associated with 'pure' fusion and fusion-fission hybrid power reactors is that energy extraction from the first wall is not critical from the point of view of hybrid system economics. This allows the consideration of low temperature structural material for first wall application. The mechanical arrangement consists of a series of internally finned aluminium tube banks running poloidally around the torus. The coolant manifolds are at the top and bottom of the torus. The vessel is divided into sectors, the length of which depends on the spacing between TF coils. The tubes in each sector are welded to tube sheets which are in turn welded to semi-cylindrical manifolds which distribute the coolant uniformly to the tubes. The tubes, which are approx. equal to 2.5 cm in diameter at the manifold location, traverse the torus poloidal periphery and change from a circular cross section to a 2:1 elliptical cross section at the horizontal midplane. The arched tube is designed to be self-supporting between the manifold locations. The vacuum vessel's first surface will be plasma flamed sprayed aluminum applied to the tubular wall. (orig./GG)

  4. Pulse discharge cleaning of the vacuum vessel of HL-1 tokamak

    International Nuclear Information System (INIS)

    Li Guodong; Zhu Yukun; Xiao Zhenggui; Sun Shouqi; Ze Mingrui

    1986-01-01

    The HL-1 Tokamak was test-operated on September 21, 1984. During the period of vacuum conditioning, including 60 hours of baking up to 200 deg C and 7 x 10 4 shots of pulse discharge cleaning, the calculated quantities of carbon and oxygen removed are equivalent to 24 and 6 monolayers, respectively. Then, 124 shots of tokamak discharge were performed with low level plasma parameters. The plasma current and pulse length achieved were 60 kA and 85 ms at the toroidal magnetic field of 15 kG. This paper described the techniques used and the effect on discharge characteristics of bakeout and pulse discharge cleaning of the vacuum vessel

  5. Structural analysis of TFTR vacuum vessel bellows and bellows cover sections

    International Nuclear Information System (INIS)

    Driesen, G.

    1975-10-01

    A structural evaluation of the bellows and bellows cover sections was undertaken in order to confirm the structural integrity of these TFTR vacuum vessel components in the prescribed operating environment. The evaluations investigate component stability, stress, and deflection behavior. The products of this investigation appearing in this report include; (1) Structural verification of the vacuum bellows as currently defined in an operating environment of one atmosphere external pressure and 93 0 C (200 0 F) uniform temperature. (2) The establishment of a structurally adequate design configuration for the bellows cover section. (3) The presentation of a parametric study which indicates the effects of varying some bellows cover section parameters in order to obtain acceptable variations of this design configuration. (4) A verification of bellows and bellows cover section integrity to preclude a fatigue type failure for reactor startup and shutdown cyclic life in the design environment

  6. Thermal performance of the MFTF magnets

    International Nuclear Information System (INIS)

    VanSant, J.H.

    1983-01-01

    A yin-yang pair of liquid-helium (LHe) cooled, superconducting magnets were tested last year at the Lawrence Livermore National Laboratory (LLNL) as part of a series of tests with the Mirror Fusion Test Facility (MFTF). These tests were performed to determine the success of engineering design used in major systems of the MFTF and to provide a technical base for rescoping from a single-mirror facility to the large tandem-mirror configuration (MFTF-B) now under construction. The magnets were cooled, operated at their design current and magnetic field, and warmed to atmospheric temperature. In this report, we describe their thermal behavior during these tests

  7. 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

  8. 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

  9. 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.

  10. Possible incorporation of a dee-shaped vacuum vessel in Doublet III

    International Nuclear Information System (INIS)

    Davis, L.; Rawls, J.M.

    1979-11-01

    The design of Doublet III allows relative straightforward incorporation of any of a number of possible dee-configuration vacuum vessels that can serve as relevant size tests of reactor regime devices. Configurations simulating those of JET, ETF and INTOR with plasma areas larger than TFTR can be attained with significant physics parameter results. Such modifications to Doublet III could be incorporated into planned upgrade activites with operations beginning in 1984, early enough to influence the designs of ETF and INTOR and test the scaling laws, poloidal coil system, and impurity control systems proposed for these ignition devices

  11. 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

  12. 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.

  13. Mechanical impacts of poloidal eddy currents on the continuous vacuum vessel of a tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo.

    1996-11-01

    Poloidal eddy currents are induced on the continuous torus vacuum vessel by changes of the toroidal field during the machine start-up (toroidal field coil charge), shut-down (toroidal field coil discharge) and plasma disruption (plasma diamagnetism change). Analytic forms for the eddy currents flowing on the vessel, consequent pressures and forces acting on it are presented in this report. The results are applied to typical operation modes of the KT-2 tokamak. Stress analysis for two typical operation modes of toroidal field damping during a machine shut-gown and plasma energy quench during a plasma disruption were carried out using 3D FEM code (ANSYS 5.2). (author). 5 tabs., 22 figs., 9 refs

  14. Assessment of stability characteristics of MFTF coils

    International Nuclear Information System (INIS)

    1979-03-01

    Certain aspects of the MFTF (Mirror Fusion Test Facility) conductor performance were investigated. Recovery analysis of the MFTF conductor was studied using GA's stability code. The maximum length of uncooled, unsoldered composite core which can recover from a thermal excursion was determined analytically. A maximum credible mechanical disturbance in terms of energy deposition, conductor motion and length, and time duration, was postulated. 5 references, 4 figures

  15. 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)

  16. Conditioning of SST-1 Tokamak Vacuum Vessel by Baking and Glow Discharge Cleaning

    International Nuclear Information System (INIS)

    Khan, Ziauddin; George, Siju; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; Pathan, Firozkhan S.; Paravastu, Yuvakiran; Raval, Dilip C.; Babu, Gattu Ramesh; Khan, Mohammed Shoaib; Pradhan, Subrata

    2016-01-01

    Highlights: • SST-1 Tokamak was successfully commissioned. • Vacuum vessel was pumped down to 4.5 × 10"–"8 mbar after baking and continuous GDC. • GDC reduced the water vapour by additional 57% while oxygen was reduced by 50%. • Under this condition, an initial plasma breakdown with current of 40 kA for 75 ms was achieved. - Abstract: Steady-state Superconducting Tokamak (SST-1) vacuum vessel (VV) adopts moderate baking at 110 ± 10 °C and the limiters baking at 250 ± 10 °C for ∼ 200 h followed by glow discharge cleaning in hydrogen (GDC-H) with 0.15 A/m"2 current density towards its conditioning prior to plasma discharge experiment. The baking in SST-1 reduces the water (H_2O) vapor by 95% and oxygen (O_2) by 60% whereas the GDC reduces the water vapor by an additional 57% and oxygen by another 50% as measured with residual gas analyzer. The minimum breakdown voltage for H-GDC in SST-1 tokamak was experimentally observed to 300 V at 8 mbar cm. As a result of these adherences, SST-1 VV achieves an ultimate of 4.5 × 10"−"8 mbar with two turbo-molecular pumps with effective pumping speed of 3250 l/s. In the last campaign, SST-1 has achieved successful plasma breakdown, impurity burn through and a plasma current of ∼ 40 kA for 75 ms.

  17. 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.

  18. 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

  19. 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)

  20. 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.

  1. Manufacturing device for vacuum vessel of thermonuclear reactor and manufacturing method therefor

    International Nuclear Information System (INIS)

    Yanagi, Hiroshi; Shibui, Masanao; Uchida, Takaho

    1998-01-01

    The present invention provides a method of manufacturing a vacuum vessel of a thermonuclear reactor with no welding deformation. Namely, there are disposed a manufacturing device comprises a welding machine equipped with a plurality of welding torches which can conduct synchronizing welding and a torch positioning mechanism for positioning the plurality of welding torches each at an optional distance. Then, both ends of a splice plate can be welded by the plurality of welding torches under synchronization. Accordingly, joining portions of sectors of a vacuum vessel can be welded in the site with no deviation of beveling at joining portions between an outer wall and an inner wall with the splice plate due to welding deformation. In addition, the welding machine is mounted on a travelling type clamping mechanism stand or a travelling type clamping mechanism. With such a constitution, since the peripheries of the joining portions on the inner wall are clamped with each other by the travelling type clamping mechanism, no angular distortion is caused in any welded portion of the outer wall. (I.S.)

  2. An Approach for Selection of Flow Regime and Models for Conservative Evaluation of a Vessel Integrity Monitoring System for Water-Cooled Vacuum Vessels

    International Nuclear Information System (INIS)

    Pointer, W. David; Ruggles, Arthur E.

    2003-01-01

    Thin-walled vacuum containment vessels cooled by circulating water jackets are often utilized in research and industrial applications where isolation of equipment or experiments from the influences of the surrounding environment is desirable. The development of leaks in these vessels can result in costly downtime for the facility. A Vessel Integrity Monitoring System (VIMS) is developed to detect leak formation and estimate the size of the leak to allow evaluation of the risk associated with continued operation. A wide range of leak configurations and fluid flow phenomena are considered in the evaluation of the rate at which a tracer gas dissolved in the cooling jacket water is transported into the vacuum vessel. A methodology is presented that uses basic fluid flow models and careful evaluation of their ranges of applicability to provide a conservative estimate of the transport rates for the tracer gas and hence the time required for the VIMS to detect a leak of a given size

  3. The design study of the JT-60SU device. No. 4. The vacuum vessel and cryostat of JT-60SU

    International Nuclear Information System (INIS)

    Neyatani, Yuzuru; Ushigusa, Kenkichi; Tobita, Kenji

    1997-03-01

    The vacuum vessel and the cryostat for the JT-60 Super Upgrade (JT-60SU) have been designed. Two types of the complex materials for the vacuum vessel were chosen on the basis of the avoidance of tritium occlusion and the low irradiation, i.e. (1) SUS316 covered by tungsten plate (30mm thickness) as a γ-ray shielding, (2) Ti-6Al-4V alloy covered by SUS430 plate (1mm thickness) as a tritium protector. Selecting the double skin type of vacuum vessel with toroidally continued structure gave the basic design of the vacuum vessel satisfying the design criteria of the vessel strength for the electromagnetic force, heat load and the property of radiation shielding. The characteristics of the SUS316 covered by tungsten plate type is that as the tungsten can shield the γ-ray, the dose rate inside the vacuum vessel during the maintenance can reduce effectively. The advantage of the Ti-6Al-4V alloy covered by SUS430 plate type vacuum vessel is the quick reduction of the radioactive isotope because of no production of the isotopes with long half-life periods. Channel type and vertical type of the divertor were designed. The sector type of toroidally separated structure was selected for the remote handling. The material of the armor plate was not determined because no material endure the high heat load on the divertor. The cryostat composing the dome and the tank was designed. The electromagnetic force by the eddy current, generated at the plasma start up phase and at the quench of CS super-conducting coil, were small compared to the force produced by the stress limit. (author)

  4. 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.

  5. Calculation of the electromagnetic forces on the ASDEX upgrade vacuum vessel on disruption of the plasma current

    International Nuclear Information System (INIS)

    Preis, H.

    1986-01-01

    This study investigates the magnetic field diffusion through the vacuum vessel of the ASDEX Upgrade tokamak that occurs on sudden disruption of the plasma current. Eddy currents are thereby produced in the vessel wall. Their time behaviour and distribution are determined. Furthermore, the vessel is permeated by various magnetic fields which, together with the eddy currents, exert magnetic forces in the vessel wall. These are also calculated. These numerical analyses are performed for two of the modes of operation envisaged for ASDEX Upgrade: the so-called limiter and single-null magnetic field configurations. (orig.)

  6. Location and repair of air leaks in the ATF vacuum vessel

    International Nuclear Information System (INIS)

    Schwenterly, S.W.; Gabbard, W.A.; Schaich, C.R.; Yarber, J.L.

    1989-01-01

    On the basis of partial pressure rate-of-rise and base pressure measurements, it was determined that the Advanced Toroidal Facility (ATF) vacuum vessel had an air leak in the low 10 -4 mbarx ell/s range. Pinpointing this leak by conventional helium leak-checking procedures was not possible, because large portions of the outside of the vessel are covered by the helical field coils and a structural shell. Various alternative leak detection schemes that were considered are summarized and their advantages and disadvantages noted. In the method ultimately employed, gum-rubber patches of various sizes ranging from 12.7 by 12.7 cm to 20.3 by 30.5 cm were positioned on the inside surfaces of the vessel and evacuated by the leak detector (LD). After roughly 5% of the surface was inspected in this way, a leak of >10 -5 mbar xL/s was discovered and localized to an area of 5 by 5 cm. Dye penetrant applied to this area disclosed three pinholes. Two small slag pockets were discovered while these points were being ground out. After these were rewelded, no further leakage could be found in the repaired area. Global leak rates measured after the machine was reevacuated indicated that this leak was about 30% of the overall leak rate. 1 ref., 5 figs., 1 tab

  7. Location and repair of air leaks in the ATF [Advanced Toroidal Facility] vacuum vessel

    International Nuclear Information System (INIS)

    Schwenterly, S.W.; Gabbard, W.A.; Schaich, C.R.; Yarber, J.L.

    1989-01-01

    On the basis of partial pressure rate-of-rise and base pressure measurements, it was determined that the Advanced Toroidal Facility (ATF) vacuum vessel had an air leak in the low 10 -4 mbar-ell/s range. Pinpointing this leak by conventional helium leak-checking procedures was not possible, because large portions of the outside of the vessel are covered by the helcial field coils and a structural shell. Various alternative leak-detection schemes that were considered are summarized and their advantages and disadvantages noted. In the method ultimately employed, gun-rubber patches of various sizes ranging from 12.7 by 12.7 cm to 20.3 by 30.5 cm were positioned on the inside surfaces of the vessel and evacuated by the leak detector (LD). After roughly 5% of the surface was inspected in this way, a leak of > 10 -5 mbar-ell/s was discovered and localized to an area of 5 by 5 cm. Dye penetrant applied to this area disclosed three pinholes. Two small slag pockets were discovered while these points were being ground out. After these were rewelded, no furthered leakage could be found in the repaired area. Global leak rates measured after the machine was reevacuated indicated that this leak was about 30% of the overall leak rate. 1 ref., 5 figs., 1 tab

  8. Testing of the MFTF magnets

    International Nuclear Information System (INIS)

    Kozman, T.A.; Chang, Y.; Dalder, E.N.C.

    1982-01-01

    This paper describes the cooldown and testing of the first yin-yang magnet for the Mirror Fusion Test Facility. The introduction describes the superconducting magnet; the rest of the paper explains the tests prior to and including magnet cooldown and final acceptance testing. The MFTF (originally MX) was proposed in 1976 and the project was funded for construction start in October 1977. Construction of the first large superconducting magnet set was completed in May 1981 and testing started shortly thereafter. The acceptance test procedures were reviewed in May 1981 and the cooldown and final acceptance test were done by the end of February 1982. During this acceptance testing the magnet achieved its full design current and field

  9. Estimation of neutral-beam-induced field reversal in MFTF by an approximate scaling law

    International Nuclear Information System (INIS)

    Shearer, J.W.

    1980-01-01

    Scaling rules are derived for field-reversed plasmas whose dimensions are common multiples of the ion gyroradius in the vacuum field. These rules are then applied to the tandem MFTF configuration, and it is shown that field reversal appears to be possible for neutral beam currents of the order of 150 amperes, provided that the electron temperature is at least 500 eV

  10. Baking system for ports of experimental advanced super-conducting tokamak vacuum vessel and thermal stress analysis

    International Nuclear Information System (INIS)

    Cheng Yali; Bao Liman; Song Yuntao; Yao Damao

    2006-01-01

    The baking system of Experimental Advanced Super-Conducting Toakamk (EAST) vacuum vessel is necessary to obtain the baking temperature of 150 degree C. In order to define suitable alloy heaters and achieve their reasonable layouts, thermal analysis was carried out with ANSYS code. The analysis results indicate that the temperature distribution and thermal stress of most parts of EAST vacuum vessel ports are uniform, satisfied for the requirement, and are safe based on ASME criterion. Feasible idea on reducing the stress focus is also considered. (authors)

  11. 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

  12. Kinematic analysis on rail development into vacuum vessel for ITER blanket maintenance

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Shibanuma, Kiyoshi

    2006-01-01

    The vehicle manipulator system for blanket maintenance is used as a main driving mechanism for rail development, and three driving mechanisms d1, d2 (or d2') and d3 are used as cycle sequence of the repeated operations for rail development. This repeated operation can develop the articulated rail into the vacuum vessel. The rail development scenario, kinematic analysis model for rail development without any driving mechanisms in the rail joints, equations defined the angular between two rail links, identification of rail link at repeated operation, numerical analysis results on rail deployment under the forced position control of l i+1 , new rail development scenario using two driving mechanisms d1 and d2''under one cycle sequence of the repeated operations, and rail development test are reported. (S.Y.)

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. Thermal loads on the TJ-II Vacuum Vessel under Neutral Beam Injection

    International Nuclear Information System (INIS)

    Guasp, J.; Fuentes, C.; Liniers, M.

    1996-01-01

    In this study a numerical analysis of power loads on the complex 3D structure of the TJ-II Vacuum Vessel, moderated with reasonable accuracy, under NBI, is done. To do this it has been necessary to modify deeply the DENSB code for power loads in order to include the TJ-II VV wall parts as targets and as beam scrapers, allowing the possibility of self-shadowing. After a short description of the primitive version of the DENSB code (paragraph 2) and of the visualisation code MOVIE(paragraph 3), the DENSB upgrading are described (paragraphs 4,5) and finally the results are presented (paragraph 6). These code modifications and the improving on the visualization tools provide more realistic load evaluations, both with and without plasma, validating former results and showing clearly the VV zones that will need new protections. (Author)

  19. Impacts of lost fast ions on the TJ-II Vacuum Vessel during NBI

    International Nuclear Information System (INIS)

    Guasp, J.

    1995-01-01

    The possible deposition patterns, on the Vacuum Vessel, of lost fast ions during the balanced tangential NBI in TJ-II helical axis Stellarator are analysed theoretically, establishing the relation between those impact points, the plasma exit and birth positions and the magnetic configuration characteristics. It is shown that direct losses are the most important, mainly those produced by the beam injected with the same direction that the magnetic field, increasing with beam energy and plasma density but with impacts remaining fixed on well defined zones, a periodically distributed along the Hard Core cover plates, producing high loads at high densities. The remaining losses, except for the shine through ones that predominate at low density, are periodically distributed, with smooth maxima and produce very low loads. No overlapping between the different kind of losses or beams is observed. (Author) 6 refs

  20. Impacts of the CX neutrals on the Vacuum Vessel of TJ-II during NBI

    International Nuclear Information System (INIS)

    Guasp, J.

    1995-09-01

    A numerical analysis of the impact patterns on the Vacuum Vessel produced by CX neutrals during the tangential balanced NBI in TJ-II Helical Axis Stellerator has been done. The results show periodical distribution with smooth maxima and mild loads, concentrated prefentlyon the HC plates. A certain preference of these neutral to emerge downwards from the plasma appears, as consequence of a similar trend for the trapped particles. The differences between the impacts produced by the beam paralel to the magnetic field and the opposite one are small, once more as a consequence of the loss of memory of trapped particles to initial direction. The dependence of loads with plasma density and beam energy follows the trend of CX losses, decreasing strongly with increasing density and decreasing, more smoothly, with energy

  1. Impacts of the Shine Through neutrals on the Vacuum Vessel of TJ-II during NBI

    International Nuclear Information System (INIS)

    Guasp, J.; Liniers, M.

    1995-01-01

    A numerical analysis of the impact patterns on the Vacuum Vessel produced by Shine Through neutrals during the tangential balanced NBI in TJ-II Helical Axis Stellarator has been done. The results show two main concentrations. The first one the circular part of the Hard Core, near the zone of closest approach of the beam. The second one, the most important, around the beam exit, on the border between the plate of the HC cover and the sector wall. As expected, the Shine Through loads decrease strongly with plasma density, predominating at low density at NBI start, decreasing quickly when density increases and increasing slightly with the beam energy. No overlapping with lost fast ions impacts is observed, that, in addition, show an opposite behaviour with density. (Author) 3 refs

  2. Impacts of the Shine Through neutrals on the Vacuum Vessel of TJ-II during NBI

    International Nuclear Information System (INIS)

    Guasp, J.; Liniers, M.

    1995-09-01

    A numerical analysis of the impact patterns on the Vacuum Vessel produced by Shine through neutrals during the tangential balanced NBI in TJ-II Helical Axis Stellarator has been done. The results show two main concentrations. The first one the circular part of the Hard Core, near the zone of closest approach of the beam. The second one, the most important, around the beam exit, on the border between the plate of the HC cover and the sector wall. As expected, the Shine through loads decrease strongly with plasma density, predominating at low density at NBI start, decreasing quickly when density increases and increasing slightly with the beam energy. No overlapping with lost fast ions impacts is observed, that, in addition, show an opposite behaviour with density

  3. Impacts of the CX neutrals on the Vacuum Vessel of TJ-II during NBI

    International Nuclear Information System (INIS)

    Guasp, J.

    1995-01-01

    A numerical analysis of the impact patterns on the Vacuum Vessel produced by CX neutrals during the tangential balanced NBI in TJ-II Helical Axis Stellarator has been done. The results show periodical distributions with smooth maxima and mild loads, concentrated preferential on the HC plates. A certain preference of these neutral to emerge down wards from the plasma appears, as a consequence of a similar trend for the trapped particles. The differences between the impacts produced by the beam parallel to the magnetic field and the opposite one are small, once more as a consequence of the loss of memory of trapped particles to initial direction. The dependence of loads with plasma density and beam energy follows the trend of CX losses, decreasing strongly with increasing density and decreasing, more smoothly, with energy. (Author) 3 refs

  4. Impacts of lost fast ions on the TJ-II Vacuum vessel during NBI

    International Nuclear Information System (INIS)

    Guasp, J.

    1995-09-01

    The possible deposition patterns, on the Vacuum Vessel, of lost fast ions during the balanced tangential NBI in TJ-II helical axis Stellarator are analysed theoretically, establishing the relation between those impact points, the plasma exit and birth positions and positions and the magnetic configuration characteristics. It is shown that direct losses are the most important, mainly those produced by the beam injected with the same direction that the magnetic field, increasing with beam energy and plasma density but with impacts remaining fixed on well defined zones, a periodically distributed along the Hard Core cover plates, producing high loads at high densities. The remaining losses, except for the shine through ones that predominate at low density, are periodically distributed, with smooth maxima and produce very low loads. No overlapping between the different kind of losses or beams is observed

  5. 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.

  6. 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

  7. 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.

  8. 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

  9. Conditioning of SST-1 Tokamak Vacuum Vessel by Baking and Glow Discharge Cleaning

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Ziauddin, E-mail: ziauddin@ipr.res.in; George, Siju; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; Pathan, Firozkhan S.; Paravastu, Yuvakiran; Raval, Dilip C.; Babu, Gattu Ramesh; Khan, Mohammed Shoaib; Pradhan, Subrata

    2016-02-15

    Highlights: • SST-1 Tokamak was successfully commissioned. • Vacuum vessel was pumped down to 4.5 × 10{sup –8} mbar after baking and continuous GDC. • GDC reduced the water vapour by additional 57% while oxygen was reduced by 50%. • Under this condition, an initial plasma breakdown with current of 40 kA for 75 ms was achieved. - Abstract: Steady-state Superconducting Tokamak (SST-1) vacuum vessel (VV) adopts moderate baking at 110 ± 10 °C and the limiters baking at 250 ± 10 °C for ∼ 200 h followed by glow discharge cleaning in hydrogen (GDC-H) with 0.15 A/m{sup 2} current density towards its conditioning prior to plasma discharge experiment. The baking in SST-1 reduces the water (H{sub 2}O) vapor by 95% and oxygen (O{sub 2}) by 60% whereas the GDC reduces the water vapor by an additional 57% and oxygen by another 50% as measured with residual gas analyzer. The minimum breakdown voltage for H-GDC in SST-1 tokamak was experimentally observed to 300 V at 8 mbar cm. As a result of these adherences, SST-1 VV achieves an ultimate of 4.5 × 10{sup −8} mbar with two turbo-molecular pumps with effective pumping speed of 3250 l/s. In the last campaign, SST-1 has achieved successful plasma breakdown, impurity burn through and a plasma current of ∼ 40 kA for 75 ms.

  10. 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

  11. Advanced cutting, welding and inspection methods for vacuum vessel assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Jones, L. E-mail: jonesl@ipp.mgg.de; Alfile, J.-P.; Aubert, Ph.; Punshon, C.; Daenner, W.; Kujanpaeae, V.; Maisonnier, D.; Serre, M.; Schreck, G.; Wykes, M

    2000-11-01

    ITER requires a 316 l stainless steel, double-skinned vacuum vessel (VV), each shell being 60 mm thick. EFDA (European Fusion Development Agreement) is investigating methods to be used for performing welding and NDT during VV assembly and also cutting and re-welding for remote sector replacement, including the development of an Intersector Welding Robot (IWR) [Jones et al. This conference]. To reduce the welding time, distortions and residual stresses of conventional welding, previous work concentrated on CO{sub 2} laser welding and cutting processes [Jones et al. Proc. Symp. Fusion Technol., Marseilles, 1998]. NdYAG laser now provides the focus for welding of the rearside root and for completing the weld for overhead positions with multipass filling. Electron beam (E-beam) welding with local vacuum offers a single-pass for most of the weld depth except for overhead positions. Plasma cutting has shown the capability to contain the backside dross and preliminary work with NdYAG laser cutting has shown good results. Automated ultrasonic inspection of assembly welds will be improved by the use of a phased array probe system that can focus the beam for accurate flaw location and sizing. This paper describes the recent results of process investigations in this R and D programme, involving five European sites and forming part of the overall VV/blanket research effort [W. Daenner et al. This conference].

  12. Overview of the MFTF electrical systems

    International Nuclear Information System (INIS)

    Lindquist, W.B.; Eckard, R.D.; Holdsworth, T.; Mooney, L.J.; Moyer, D.R.; Peterson, R.L.; Shimer, D.W.; Wyman, R.H.; VanNess, H.W.

    1979-01-01

    The Mirror Fusion Test Facility, scheduled for completion in October 1981, will contain a complex, state-of-the-art array of electrical and electronics equipment valued at over 60 M$. Three injector systems will be employed to initiate and sustain the MFTF deuterium plasma. A plasma streaming system and a startup neutron beam system will be used to establish a target plasma. A sustaining neutral beam system will be used to fuel and sustain the MFTF plasma for 0.5 s. Additional power supply systems required on MFTF include two magnet power supplies with quench protection circuitry for powering the superconducting YIN/YANG magnet pair and eight 10 KHz power supplies for powering the Ti gettering system. Due to the complexity, physical size, and multiple systems of MFTF, a distributed, hierarchial, computer control and instrumentation system will be used. Color graphic, touch-panel, control consoles will provide the man-machine interface. The MFTF will have the capability of conducting an experiment every five minutes

  13. Improvement of initial vacuum condition along 2008-2010 KSTAR campaign by vessel baking

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwang Pyo, E-mail: kpkim@nfri.er.ke [National Fusion Research Institute, Gwahagno 113, Daejeon 305-333 (Korea, Republic of); Hong, S.H.; Jung, N.Y.; Kim, S.T.; Kim, H.T.; Lee, K.S.; Kim, K.M.; Bang, E.N.; Chang, Y.B.; Kim, H.K.; Chu, Y.; Kim, Y.O.; Park, S.H.; Woo, I.S.; Hong, J.S.; Kim, S.W.; Park, K.R.; Na, H.K.; Yang, H.L.; Kim, Y.S. [National Fusion Research Institute, Gwahagno 113, Daejeon 305-333 (Korea, Republic of)

    2011-10-15

    Korea Superconducting Tokamak Advanced Research (KSTAR) is upgraded for its KSTAR 3rd campaign for new target mission to produce the D-shaped plasma with a target plasma current of 500 kA and/or pulse length of 5 s. New Plasma Facing Components (PFCs) are installed which leads to the increase of the surface area of the vessel by a factor of about 5. The vacuum conditioning such as the vessel baking has been performed in order to remove various kinds of impurities including H{sub 2}O, carbon and oxygen for the plasma. The total outgassing rate in the KSTAR 1st campaign was measured as 1.5 x 10{sup -4} mbar l s{sup -1} which is increased by a factor of 3 (6.49 x 10{sup -4} mbar l s{sup -1}) in the KSTAR 3rd campaign. Nevertheless, the outgassing rates per unit area have been decreased from 9.31 x 10{sup -5} mbar l m{sup -2} s{sup -1} to 1.22 x 10{sup -5} mbar l m{sup -2} s{sup -1} due to the upgrade of baking system and series of baking operation.

  14. Improvement of initial vacuum condition along 2008-2010 KSTAR campaign by vessel baking

    International Nuclear Information System (INIS)

    Kim, Kwang Pyo; Hong, S.H.; Jung, N.Y.; Kim, S.T.; Kim, H.T.; Lee, K.S.; Kim, K.M.; Bang, E.N.; Chang, Y.B.; Kim, H.K.; Chu, Y.; Kim, Y.O.; Park, S.H.; Woo, I.S.; Hong, J.S.; Kim, S.W.; Park, K.R.; Na, H.K.; Yang, H.L.; Kim, Y.S.

    2011-01-01

    Korea Superconducting Tokamak Advanced Research (KSTAR) is upgraded for its KSTAR 3rd campaign for new target mission to produce the D-shaped plasma with a target plasma current of 500 kA and/or pulse length of 5 s. New Plasma Facing Components (PFCs) are installed which leads to the increase of the surface area of the vessel by a factor of about 5. The vacuum conditioning such as the vessel baking has been performed in order to remove various kinds of impurities including H 2 O, carbon and oxygen for the plasma. The total outgassing rate in the KSTAR 1st campaign was measured as 1.5 x 10 -4 mbar l s -1 which is increased by a factor of 3 (6.49 x 10 -4 mbar l s -1 ) in the KSTAR 3rd campaign. Nevertheless, the outgassing rates per unit area have been decreased from 9.31 x 10 -5 mbar l m -2 s -1 to 1.22 x 10 -5 mbar l m -2 s -1 due to the upgrade of baking system and series of baking operation.

  15. DSTAR: A comprehensive tokamak resistive disruption model for vacuum vessel components

    International Nuclear Information System (INIS)

    Merrill, B.J.; Jardin, S.C.

    1987-01-01

    A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces than can occur during major tokamak plasma disruptions. A disruption analysis has been performed for the TFCX fusion device. The limiters and inboard first wall were assumed to be clad with beryllium. Disruption simulations were performed with and without these structures present, to determine their electromagnetic influence. The results with structure show that the ablated wall material is transported poloidally, as well as radially, in the plasma causing the outermost regions of the plasma to cool. The plasma moves downward and deforms while maintaining contact with the lower limiter. This motion maintains the peak impurity radiant source directly above the exposed surface. For the disruption simulation without the vacuum vessel included, the plasma moves radially along the lower limiter until it contacts the inboard wall, causing ablation of this surface as well. The conclusion is drawn that disruption simulations that do not include both the thermal and electromagnetic response of the vaccum vessel will not result in an accurate prediction. (orig.)

  16. 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

  17. Dutch supplier rewarded for manufacture of the two vacuum vessels for the ATLAS end-cap toroids

    CERN Multimedia

    Maximilien Brice

    2003-01-01

    The ATLAS collaboration has presented an award for outstanding supplier performance to Dutch firm Schelde Exotech. Based on a design by Rutherford Appleton Laboratory, UK, Schelde Exotech manufactured under a NIKHEF contract the two 500 m3 large vacuum vessels for the cryostats of the ATLAS end-cap toroids. These 11-metre diameter castellated aluminium vessels with stainless-steel bore tube are essentially made up of 40-mm-thick plates for the shells, 75-mm-thick plates for the endplates, and 150-mm-thick bars for the flanges. Because of transport constraints, the vessels were made in halves, temporarily sealed and vacuum tested at the works, then transported to CERN for final assembly and acceptance tests. Both vessels were vacuum-tight and the meticulous and clean way of working ensured that a high vacuum was obtained within a few days of pumping. The delivery to CERN was completed in July 2002. Representatives of Schelde Exotech are seen here receiving their award in the ATLAS assembly hall. In the backgro...

  18. Baking of the vacuum vessel prototype of the Spanish stellarator with a control system based on neural network

    International Nuclear Information System (INIS)

    Botija, J.; Alonso, J.; Blaumoser, M.

    1995-01-01

    To bake uniformly, up to 150 C, the vacuum vessel of the Spanish Stellarator TJ-II represents a difficult task to be demonstrated. In order to study the temperature distribution in the vessel, a prototype of this vacuum vessel, mounted in a stainless steel structure, has been heated by means of electrical panels and eddy currents. The induction heating system is provided applying 498 A/11.7 V at 50 Hz to the toroidal field coil located in the middle of the vessel prototype. Practically, this system only heats adequately the rings and poorly the so called groove of the vacuum vessel. On the contrary, the electrical heaters, with a power density of 0.5 W/cm 2 , heat the external part of the sectors and ports. The high density of temperature sensors ensures the uniformity of the heating process during the long heating cycles, making advisable a fault-tolerant control system based on Artificial Neural Networks (ANNs) that implements the control loop to regulate and protect both heating systems. This paper deals with the results of this experiment

  19. Industrialization and production of neutral beam ion sources for MFTF

    International Nuclear Information System (INIS)

    Lynch, W.S.

    1981-01-01

    The existing LLNL designs of the 20 and 80kV deuterium fueled Neutral Beam Ion Source Modules (NBSM) have been industrialized and are being produced successfully for the MFTF. Industrialization includes value engineering, production engineering, cost reduction, fixturing, facilitation and procurement of components. Production assembly, inspection and testing is being performed in a large electronics manufacturing plant. Decades of experience in high voltage, high vacuum power tubes is being applied to the procedures and processes. Independent quality and reliability assurance criteria are being utilized. Scheduling of the various engineering, procurement and manufacturing task is performed by the use of a Critical Path Method (CPM) computer code, Innovative, computerized grid alignment methods were also designed and installed specifically for this project. New jointing and cleaning techniques were devised for the NBSMs. Traceability and cost control are also utilized

  20. MFTF-α + T shield design

    International Nuclear Information System (INIS)

    Gohar, Y.

    1985-01-01

    MFTF-α+T is a DT upgrade option of the Tandem Mirror Fusion Test Facility (MFTF-B) to study better plasma performance, and test tritium breeding blankets in an actual fusion reactor environment. The central cell insert, designated DT axicell, has a 2-MW/m 2 neutron wall loading at the first wall for blanket testing. This upgrade is completely shielded to protect the reactor components, the workers, and the general public from the radiation environment during operation and after shutdown. The shield design for this upgrade is the subject of this paper including the design criteria and the tradeoff studies to reduce the shield cost

  1. 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

  2. Kinematic and dynamic analysis of a serial-link robot for inspection process in EAST vacuum vessel

    International Nuclear Information System (INIS)

    Peng Xuebing; Yuan Jianjun; Zhang Weijun; Yang Yang; Song Yuntao

    2012-01-01

    Highlights: ► A serial-link robot FIVIR is proposed for inspection of EAST PFCs between plasma shots. ► The FIVIR is a function modular design and has specially designed curvilinear mechanism for axes 4–6. ► The D-H coordinate systems, forward and inverse kinematic model can be easily established and solved for the FIVIR. ► The FIVIR can fulfill the required workspace and has a good dynamic performance in the inspection process. - Abstract: The present paper introduces a serial-link robot which is named flexible in-vessel inspection robot (FIVIR) and developed for Experimental Advanced Superconducting Tokamak (EAST). The task of the robot is to carry process tools, such as viewing camera and leakage detector, to inspect the components installed inside of EAST vacuum vessel. The FIVIR can help to understand the physical phenomena which could be happened in the vacuum vessel during plasma operation and could be one part of EAST remote handling system if needed. The FIVIR was designed with the consideration of having easy control and a good mechanics property which drives it resulted in function modular design. The workspace simulation and kinematic analysis are given in this paper. The dynamic behavior of the FIVIR is studied by multi-body system simulation using ADAMS software. The study result shows the FIVIR has ascendant kinematic and dynamic performance and can fulfill the design requirement for inspection process in EAST vacuum vessel.

  3. Kinematic and dynamic analysis of a serial-link robot for inspection process in EAST vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Peng Xuebing, E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Shushanhu Road 350, Hefei, Anhui 230031 (China); Yuan Jianjun; Zhang Weijun [Research Institute of Robotics, Mechanical Engineering School, Shanghai Jiao Tong University, No.800, Dong Chuan Road, Min Hang District, Shanghai 200240 (China); Yang Yang; Song Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Shushanhu Road 350, Hefei, Anhui 230031 (China)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer A serial-link robot FIVIR is proposed for inspection of EAST PFCs between plasma shots. Black-Right-Pointing-Pointer The FIVIR is a function modular design and has specially designed curvilinear mechanism for axes 4-6. Black-Right-Pointing-Pointer The D-H coordinate systems, forward and inverse kinematic model can be easily established and solved for the FIVIR. Black-Right-Pointing-Pointer The FIVIR can fulfill the required workspace and has a good dynamic performance in the inspection process. - Abstract: The present paper introduces a serial-link robot which is named flexible in-vessel inspection robot (FIVIR) and developed for Experimental Advanced Superconducting Tokamak (EAST). The task of the robot is to carry process tools, such as viewing camera and leakage detector, to inspect the components installed inside of EAST vacuum vessel. The FIVIR can help to understand the physical phenomena which could be happened in the vacuum vessel during plasma operation and could be one part of EAST remote handling system if needed. The FIVIR was designed with the consideration of having easy control and a good mechanics property which drives it resulted in function modular design. The workspace simulation and kinematic analysis are given in this paper. The dynamic behavior of the FIVIR is studied by multi-body system simulation using ADAMS software. The study result shows the FIVIR has ascendant kinematic and dynamic performance and can fulfill the design requirement for inspection process in EAST vacuum vessel.

  4. 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.

  5. 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

  6. Rationalization and utilization of double-wall vacuum vessel for tokamak fusion facility

    International Nuclear Information System (INIS)

    Nakahira, Masataka

    2005-09-01

    Vacuum Vessel (VV) of ITER is difficult to apply a non-destructive in-service inspection (ISI) and then new safety concept is needed. Present fabrication standards are not applicable to the VV, because the access is limited to the backside of closure weld of double wall. Fabrication tolerance of VV is ± 5mm even the structure is huge as high as 10m. This accuracy requires a rational method on the estimation of welding deformation. In this report, an inherent safety feature of the tokamak is proved closing up a special characteristic of termination of fusion reaction due to tiny water leak. A rational concept not to require ISI without sacrificing safety is shown based on this result. A partial penetration T-welded joint is proposed to establish a rational fabrication method of double wall. Strength and susceptibility to crevice corrosion is evaluated for this joint and feasibility is confirmed. A rational method of estimation of welding deformation for large and complex structure is proposed and the efficiency is shown by comparing analysis and experimental results of full-scale test. (author)

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. Study of radiation heat transfer between PFC and vacuum vessel during SST-1 baking

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhuri, Paritosh E-mail: paritosh@ipr.res.in; Chenna Reddy, D.; Santra, P.; Khirwadkar, S.; Ravi Pragash, N.; Saxena, Y.C

    2003-01-01

    Steady-state superconducting tokamak (SST-1) is a medium size tokamak with superconducting magnetic field coils. Plasma facing components (PFC) of SST-1 are placed inside the vacuum vessel (VV) of the tokamak and are designed to be compatible for steady-state operation. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m{sup 2}. In addition to remove high heat fluxes, the PFC are also designed to be compatible for baking at high temperature. Since it is difficult to calculate the radiation heat loads between PFC and VV in a 3-D irregular geometry, a simplified model of concentric cylinders has been chosen for the purpose of estimation of the power requirements and the thermal responses of PFC and VV during their bakeout phases. Thermal responses of the PFC and VV have been analysed and the analytical results have been compared with 2-D finite element analysis using ANSYS. The radiation losses between PFC and VV also have been evaluated on the actual model containing all PFC inside the VV.

  12. 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.

  13. Electromagnetic and structural analyses of the vacuum vessel and plasma facing components for EAST

    International Nuclear Information System (INIS)

    Xu, Weiwei; Liu, Xufeng; Song, Yuntao; Li, Jun; Lu, Mingxuan

    2013-01-01

    Highlights: • The electromagnetic and structural responses of VV and PFCs for EAST are analyzed. • A detailed finite element model of the VV including PFCs is established. • The two most dangerous scenarios, major disruptions and downward VDEs are considered. • The distribution patterns of eddy currents, EMFs and torques on PFCs are analyzed. -- Abstract: During plasma disruptions, time-varying eddy currents are induced in the vacuum vessel (VV) and Plasma Facing Components (PFCs) of EAST. Additionally, halo currents flow partly through these structures during the vertical displacement events (VDEs). Under the high magnetic field circumstances, the resulting electromagnetic forces (EMFs) and torques are large. In this paper, eddy currents and EMFs on EAST VV, PFCs and their supports are calculated by analytical and numerical methods. ANSYS software is employed to evaluate eddy currents on VV, PFCs and their structural responses. To learn the electromagnetic and structural response of the whole structure more accurately, a detailed finite element model is established. The two most dangerous scenarios, major disruptions and downward VDEs, are examined. It is found that distribution patterns of eddy currents for various PFCs differ greatly, therefore resulting in different EMFs and torques. It can be seen that for certain PFCs the transient reaction force are severe. Results obtained here may set up a preliminary foundation for the future dynamic response research of EAST VV and PFCs which will provide a theoretical basis for the future engineering design of tokamak devices

  14. 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)

  15. 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.

  16. 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

  17. 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.

  18. Preliminary electromagnetic, thermal and mechanical design for first wall and vacuum vessel of FAST

    Energy Technology Data Exchange (ETDEWEB)

    Lucca, F., E-mail: Flavio.Lucca@LTCalcoli.it [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Bertolini, C. [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Crescenzi, F.; Crisanti, F. [C.R. ENEA Frascati – UT FUS, Via E. Fermi 45, IT-00044 Frascati, RM (Italy); Di Gironimo, G. [CREATE, Università di Napoli Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Labate, C. [CREATE, Università di Napoli Parthenope, Via Acton 38, 80133 Napoli (Italy); Manzoni, M.; Marconi, M.; Pagani, I. [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Ramogida, G. [C.R. ENEA Frascati – UT FUS, Via E. Fermi 45, IT-00044 Frascati, RM (Italy); Renno, F. [CREATE, Università di Napoli Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Roccella, M. [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Roccella, S. [C.R. ENEA Frascati – UT FUS, Via E. Fermi 45, IT-00044 Frascati, RM (Italy); Viganò, F. [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy)

    2015-10-15

    The fusion advanced study torus (FAST), with its compact design, high toroidal field and plasma current, faces many of the problems met by ITER, and at the same time anticipates much of the DEMO relevant physics and technology. The conceptual design of the first wall (FW) and the vacuum vessel (VV) has been defined on the basis of FAST operative conditions and of “Snow Flakes” (SF) magnetic topology, which is also relevant for DEMO. The EM loads are one of the most critical load components for the FW and the VV during plasma disruptions and a first dimensioning of these components for such loads is mandatory. During this first phase of R&D activities the conceptual design of the FW and VV have been assessed estimating, by means of FE simulations, the EM loads due to a typical vertical disruption event (VDE) in FAST. EM loads were then transferred on a FE mechanical model of the FAST structures and the mechanical response of the FW and VV design for the analyzed VDE event was assessed. The results indicate that design criteria are not fully satisfied by the current drawing of the VV and FW components. The most critical regions have been individuated and the effect of some geometrical and material changes has been checked in order to improve the structure.

  19. TSC [Tokamak Simulation Code] disruption scenarios and CIT [Compact Ignition Tokamak] vacuum vessel force evolution

    International Nuclear Information System (INIS)

    Sayer, R.O.; Peng, Y.K.M.; Strickler, D.J.; Jardin, S.C.

    1990-01-01

    The Tokamak Simulation Code and the TWIR postprocessor code have been used to develop credible plasma disruption scenarios for the Compact Ignition Tokamak (CIT) in order to predict the evolution of forces on CIT conducting structures and to provide results required for detailed structural design analysis. The extreme values of net radial and vertical vacuum vessel (VV) forces were found to be F R =-12.0 MN/rad and F Z =-3.0 MN/rad, respectively, for the CIT 2.1-m, 11-MA design. Net VV force evolution was found to be altered significantly by two mechanisms not noted previously. The first, due to poloidal VV currents arising from increased plasma paramagnetism during thermal quench, reduces the magnitude of the extreme F R by 15-50% and modifies the distribution of forces substantially. The second effect is that slower plasma current decay rates give more severe net vertical VV loads because the current decay occurs when the plasma has moved farther from midplane than is the case for faster decay rates. 7 refs., 9 figs., 1 tab

  20. 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

  1. 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

  2. Fundamental study of a water jet injected into a vacuum vessel of fusion reactor under the ingress of coolant event

    International Nuclear Information System (INIS)

    Takase, Kazuyuki; Kunugi, Tomoaki; Seki, Yasushi; Kurihara, Ryouichi; Ueda, Shuzou

    1996-01-01

    As one of some transient sequences for the thermofluid safety in ITER, pressure rise and boiling heat transfer characteristics in a Tokamak vacuum vessel during an ingress of coolant event (ICE) are being investigated experimentally by using the preliminary ICE apparatus. The pressure rise rates in the vacuum vessel and the wall temperature distributions on the target plate were measured quantitatively and clarified at first. In addition, a two-phase flow under the ICE conditions was analyzed numerically for predicting the experimental results using one-dimensional transport equations and the drift-flux model. The experimental results were compared with the numerical results. It was found that the pressurization behavior during the ICE conditions could be estimated qualitatively by the present numerical analyses. 5 refs., 5 figs

  3. 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)

  4. 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)

  5. 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

  6. 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.

  7. Verification of radiation heat transfer analysis in KSTAR PFC and vacuum vessel during baking

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, S.Y. [Chungnam National University, 79 Daehak-ro, Yuseong-gu, Daejeon 34167 (Korea, Republic of); Kim, Y.J., E-mail: k43689@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahang-ro, Yuseong-gu, Daejeon 34133 (Korea, Republic of); Kim, S.T.; Jung, N.Y.; Im, D.S.; Gong, J.D.; Lee, J.M.; Park, K.R.; Oh, Y.K. [National Fusion Research Institute, 169-148 Gwahang-ro, Yuseong-gu, Daejeon 34133 (Korea, Republic of)

    2016-11-01

    Highlights: • Thermal network is used to analyze heat transfer from PFC to VV. • Three heat transfer rate equations are derived based on the thermal network. • The equations is verified using Experimental data and design documents. • Most of the heat lost in tokamak is transferred to experimental room air. • The heat loss to the air is 101 kW of the total heat loss of 154 kW in tokamak. - Abstract: KSTAR PFC (Plasma Facing Component) and VV (Vacuum Vessel) were not arrived at the target temperatures in bake-out phase, which are 300 °C and 110 °C, respectively. The purpose of this study is to find out the reason why they have not been reached the target temperature. A thermal network analysis is used to investigate the radiation heat transfer from PFC to VV, and the thermal network is drawn up based on the actual KSTAR tokamak. The analysis model consists of three equations, and is solved using the EES (Engineering Equation Solver). The heat transfer rates obtained with the analysis model is verified using the experimental data at the KSTAR bake-out phase. The analyzed radiation heat transfer rates from PFC to VV agree quite well with those of experiment throughout the bake-out phase. Heat loss from PFC to experimental room air via flange of VV is also calculated and compared, which is found be the main reason of temperature gap between the target temperature and actually attained temperature of KSTAR PFC.

  8. Verification of radiation heat transfer analysis in KSTAR PFC and vacuum vessel during baking

    International Nuclear Information System (INIS)

    Yoo, S.Y.; Kim, Y.J.; Kim, S.T.; Jung, N.Y.; Im, D.S.; Gong, J.D.; Lee, J.M.; Park, K.R.; Oh, Y.K.

    2016-01-01

    Highlights: • Thermal network is used to analyze heat transfer from PFC to VV. • Three heat transfer rate equations are derived based on the thermal network. • The equations is verified using Experimental data and design documents. • Most of the heat lost in tokamak is transferred to experimental room air. • The heat loss to the air is 101 kW of the total heat loss of 154 kW in tokamak. - Abstract: KSTAR PFC (Plasma Facing Component) and VV (Vacuum Vessel) were not arrived at the target temperatures in bake-out phase, which are 300 °C and 110 °C, respectively. The purpose of this study is to find out the reason why they have not been reached the target temperature. A thermal network analysis is used to investigate the radiation heat transfer from PFC to VV, and the thermal network is drawn up based on the actual KSTAR tokamak. The analysis model consists of three equations, and is solved using the EES (Engineering Equation Solver). The heat transfer rates obtained with the analysis model is verified using the experimental data at the KSTAR bake-out phase. The analyzed radiation heat transfer rates from PFC to VV agree quite well with those of experiment throughout the bake-out phase. Heat loss from PFC to experimental room air via flange of VV is also calculated and compared, which is found be the main reason of temperature gap between the target temperature and actually attained temperature of KSTAR PFC.

  9. 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

  10. 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)

  11. High power Nd:YAG laser welding in manufacturing of vacuum vessel of fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jokinen, Tommi E-mail: tommi.jokinen@vtt.fi; Kujanpaeae, Veli E-mail: veli.kujanpaa@lut.fi

    2003-09-01

    Laser welding has shown many advantages over traditional welding methods in numerous applications. The advantages are mainly based on very precise and powerful heat source of laser light, which change the phenomena of welding process when compared with traditional welding methods. According to the phenomena of the laser welding, penetration is deeper and thus welding speed is higher. Because of the precise power source and high-welding speed, the heat input to the workpiece is small and distortions are reduced. Also, the shape of laser weld is less critical for distortions than traditional welds. For welding thick sections, the usability of lasers is not so practical than with thin sheets, because with power levels of present Nd:YAG lasers depth of penetration is limited up to about 10 mm by single-pass welding. One way to overcome this limitation is to use multi-pass laser welding, in which narrow gap and filler wire is applied. By this process, thick sections can be welded with smaller heat input and then smaller distortions and the process seems to be very effective comparing 'traditional' welding methods, not only according to the narrower gap. Another way to increase penetration and fill the groove is by using the so-called hybrid process, in which laser and GMAW (gas metal arc welding) are combined. In this paper, 20-mm thick austenitic stainless steel was welded using narrow gap configuration with a multi-pass technique. Two welding procedures were used: Nd:YAG laser welding with filler wire and with addition of GMAW, the hybrid process. In the welding experiments, it was noticed that both processes are feasible for welding thicker sections with good quality and with minimal distortions. Thus, these processes should be considered when the evaluation of the welding process is done for joining vacuum vessel sectors of ITER.

  12. 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)

  13. 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

  14. 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.

  15. 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.

  16. MFTF test coil construction and performance

    International Nuclear Information System (INIS)

    Cornish, D.N.; Zbasnik, J.P.; Leber, R.L.; Hirzel, D.G.; Johnston, J.E.; Rosdahl, A.R.

    1978-01-01

    A solenoid coil, 105 cm inside the 167 cm outside diameter, has been constructed and tested to study the performance of the stabilized Nb--Ti conductor to be used in the Mirror Fusion Test Facility (MFTF) being built at Lawrence Livermore Laboratory. The insulation system of the test coil is identical to that envisioned for MFTF. Cold-weld joints were made in the conductor at the start and finish of each layer; heaters were fitted to some of these joints and also to the conductor at various locations in the winding. This paper gives details of the construction of the coil and the results of the tests carried out to determine its propagation and recovery characteristics

  17. Date base management system for the MFTF

    International Nuclear Information System (INIS)

    Choy, J.H.; Wade, J.A.

    1979-01-01

    The data base management system (DBMS) for the Mirror Fusion Test Facility (MFTF) is described as relational in nature and distributed across the nine computers of the supervisory control and diagnostics system. This paper deals with a reentrant runtime package of routines that are used to access data items, the data structures to support the runtime package, and some of the utilities in support of the DBMS

  18. 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

  19. 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

  20. 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

  1. Activation of the JET vacuum vessel: a comparison of calculated with measured gamma-radiation fluxes and dose rates

    International Nuclear Information System (INIS)

    Jarvis, O.N.; Sadler, G.; Avery, A.; Verschuur, K.A.

    1988-01-01

    The gamma-radiation dose-rates inside the JET vacuum vessel due to induced radioactivity were measured at intervals throughout the 1986 period of operation, and the decay gamma energy spectrum was measured during the subsequent lengthy shutdown. The dose-rates were found to be in good agreement with values calculated using the neutron yield records compiled from the time-resolved neutron yield monitor responses for individual discharges. This result provides strong support for the reliability of the neutron yield monitor calibration. (author)

  2. Influence of INCONEL 625 composition on the activation characteristics of the vacuum vessel of experimental fusion tokamaks

    International Nuclear Information System (INIS)

    Cambi, G.; Cepraga, D.G.; Boeriu, S.; Maganzani, I.

    1995-01-01

    The radioactive inventory, the decay heat and the contact dose rate of permanent components such as the vacuum vessel of two experimental fusion tokamaks, the compact IGNITOR-ULT and the ITER-EDA fusion machines, are evaluated by using the ENEA-Bologna integrated methodology. The vacuum vessel material considered is the INCONEL 625. The neutron flux is calculated using the VITAMIN-C 171-group library, based on EFF-2 data and the 1-D transport code XSDRNPM in the S 8 -P 3 approximation. The ANITA-2 code, using updated cross sections and decay data libraries based on EAF-3 and IRDF90 evaluation files is used for activation calculations. The fusion neutron source has been normalised to a neutron first wall load of 2 MW/m 2 and 1 MW/m 2 for IGNITOR-ULT and ITER, respectively. The material irradiation have been described by multistep time histories, resulting in the designed total fluence. Variations in the composition of INCONEL 625 have been assessed and their impact on the activation characteristics are discussed, also from the point of view of waste disposal. (orig.)

  3. 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.)

  4. 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)

  5. Supervisory control software for MFTF neutral beams

    International Nuclear Information System (INIS)

    Woodruff, J.P.

    1981-01-01

    We describe the software structures that control the operation of MFTF Sustaining Neutral Beam Power Supplies (SNBPS). These components of the Supervisory Control and Diagnostics System (SCDS) comprise ten distinct tasks that exist in the SCDS system environment. The codes total about 16,000 lines of commented Pascal code and occupy 240 kbytes of memory. The controls have been running since March 1981, and at this writing are being integrated to the Local Control System and to the power supply Pulse Power Module Controller

  6. Central cell confinement in MFTF-B

    International Nuclear Information System (INIS)

    Jong, R.A.

    1981-01-01

    The point code TANDEM has been used to survey the range of plasma parameters which can be attained in MFTF-B. The code solves for the electron and ion densities and temperatures in the central cell, yin-yang, barrier, and A-cell regions as well as the plasma potential in each region. In these studies, the A-cell sloshing ion beams were fixed while the neutral beams in the yin-yang and central cell, the gas feed in the central cell, and the applied ECRH power β, central cell ion density and temperature, and the confining potential are discussed

  7. Assessing the feasibility of a high-temperature, helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

    International Nuclear Information System (INIS)

    Barnard, H.S.; Hartwig, Z.S.; Olynyk, G.M.; Payne, J.E.

    2012-01-01

    The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B 0 = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m −2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ∼1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to

  8. Man-machine interface for the MFTF

    International Nuclear Information System (INIS)

    Speckert, G.C.

    1979-01-01

    In any complex system, the interesting problems occur at the interface of dissimilar subsystems. Control of the Mirror Fusion Test Facility (MFTF) begins with the US Congress, which controls the dollars, which control the people, who control the nine top-level minicomputers, which control the 65 microprocessors, which control the hardware that controls the physics experiment. There are many interesting boundaries across which control must pass, and the one that this paper addresses is the man-machine one. For the MFTF, the man-machine interface consists of a system of seven control consoles, each allowing one operator to communicate with one minicomputer. These consoles are arranged in a hierarchical manner, and both hardware and software were designed in a top-down fashion. This paper describes the requirements and the design of the console system as a whole, as well as the design and operation of the hardware and software of each console, and examines the possible form of a future man-machine interface

  9. Man-machine interface for the MFTF

    Energy Technology Data Exchange (ETDEWEB)

    Speckert, G.C.

    1979-11-09

    In any complex system, the interesting problems occur at the interface of dissimilar subsystems. Control of the Mirror Fusion Test Facility (MFTF) begins with the US Congress, which controls the dollars, which control the people, who control the nine top-level minicomputers, which control the 65 microprocessors, which control the hardware that controls the physics experiment. There are many interesting boundaries across which control must pass, and the one that this paper addresses is the man-machine one. For the MFTF, the man-machine interface consists of a system of seven control consoles, each allowing one operator to communicate with one minicomputer. These consoles are arranged in a hierarchical manner, and both hardware and software were designed in a top-down fashion. This paper describes the requirements and the design of the console system as a whole, as well as the design and operation of the hardware and software of each console, and examines the possible form of a future man-machine interface.

  10. Magnetic shielding tests for MFTF-B neutral beamlines

    International Nuclear Information System (INIS)

    Kerns, J.; Fabyan, J.; Wood, R.; Koger, P.

    1983-01-01

    A test program to determine the effectiveness of various magnetic shielding designs for MFTF-B beamlines was established at Lawrence Livermore National Laboratory (LLNL). The proposed one-tenth-scale shielding-design models were tested in a uniform field produced by a Helmholtz coil pair. A similar technique was used for the MFTF source-injector assemblies, and the model test results were confirmed during the Technology Demonstration in 1982. The results of these tests on shielding designs for MFTF-B had an impact on the beamline design for MFTF-B. The iron-core magnet and finger assembly originally proposed were replaced by a simple, air-core, race-track-coil, bending magnet. Only the source injector needs to be magnetically shielded from the fields of approximately 400 gauss

  11. Protective interior wall and attaching means for a fusion reactor vacuum vessel

    International Nuclear Information System (INIS)

    Phelps, R.D.; Upham, G.A.; Anderson, P.M.

    1988-01-01

    A protective wall for the interior surface of a fusion reactor vessel wall is described comprising: an array of plates, each plate of the array including a main body section, a pair of edge sections bent at an angle with respect to the main body section, and a pair of flange-like end sections each having protruding sections with cut-aways therein, the protruding sections of the flange-like end sections extending in a direction substantially parallel to the main body section; and means operatively associated with the protruding sections of the flange-like end sections of the plates for mounting the array of plates to an associated vessel wall to be protected

  12. 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.

  13. 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

  14. 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.

  15. Application of ion scattering spectrometers for the observation of process of cleaning of surfaces of materials for vacuum vessel

    International Nuclear Information System (INIS)

    Akashi, Ken-ya; Miyahara, Akira; Sagara, Akio.

    1978-01-01

    The impurity gas emitted from the surfaces of vacuum vessels was investigated by using the shadowing effect of the covering atoms. The ion scattering spectrometer used for the experiment consists of an ion source, a test sample, an energy analyzer and an ion detector. The evacuation system comprises a turbomolecular pump, a Ti-sublimation pump and an ion pump. The achieved final gas pressure is 5 x 10 -10 Torr. The ion beam intensity to a sample is 10 micro ampere/cm 2 , and the ion energy is about 1 to 1.5 keV. The quantity of oxygen on the surface of a sample molybdenum was measured in the process of evacuation. The concentration of surface oxygen decreased with the gas pressure of the system. It was found that residual oxygen was observed after the sputter etching with Ar ion impact on the surface. The reason of this residual oxygen was considered. (Kato, T.)

  16. Magnetic analysis including the field due to vacuum vessel eddy currents in the Hitachi Tokamak (HT-2)

    International Nuclear Information System (INIS)

    Abe, Mitsushi; Takeuchi, Kazuhiro; Fukumoto, Hideshi; Otsuka, Michio

    1989-01-01

    A magnetic analysis to determine plasma surface position is applied to the magnetic data of the Hitachi Tokamak (HT-2). The analysis takes account of toroidal eddy currents on the vacuum vessel wall. Magnetic probes in HT-2 are placed on both sides of the wall (plasma side and outside), making it possible to determine magnitudes of eddy currents which flow in the toroidal direction. The magnitudes of the coil currents and eddy currents are determined so as to reproduce the measured magnetic fields, and to reconstruct flux surfaces and plasma surface are reconstructed. Taking into account the eddy currents, the determination errors of the plasma surface position are reduced by up to 1/2.3 during start-up and terminating phases, compared with the case without eddy currents. (author)

  17. Structural design and analysis for the ISX-C/ATF tokamak of the vacuum vessel, coil joints, and supports

    International Nuclear Information System (INIS)

    Mayhall, J.A.; Cain, W.D.; Hammonds, C.J.; Johnson, R.L.; Gray, W.H.

    1981-01-01

    The ISX-C/ATF is being designed as a test bed for advanced toroidal concepts. Because of numerous design concepts being evaluated, a flexible, easily changeable structural-design math-model was needed to afford quick evalution of the structural feasibility of the many proposed concepts. To satisfy this need, the NASTRAN Automated Multi-Stage Substructures technique was used to build a quick-changeable math model. This technique was especially needed because all the coils, first wall and diagnostic devices are to be supported by the vacuum vessel, requiring the entire structure to be analyzed as a system. Without the use of the substructuring technique, the required man hours and computer core would have made timely design analysis impossible. To illustrate the technique, the detailed design analysis of the concept Torsatron (with helical coils and T.F. coils) is presented

  18. Application of structural mechanics methods to the design of large tandem mirror fusion devices (MFTF-B)

    International Nuclear Information System (INIS)

    Karpenko, V.N.; Ng, D.S.

    1985-01-01

    The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory requires state-of-the-art structural-mechanics methods to deal with access constraints for plasma heating and diagnostics, alignment requirements, and load complexity and variety. Large interactive structures required an integrated analytical approach to achieve a resonable level of overall system optimization. The Tandem Magnet Generator (TMG) creates a magnet configuration for the EFFI calculation of electromagnetic-field forces that, coupled with other loads, form the input loading to magnetic and vessel finite-element models. The anlytical results provide the data base for detailed design of magnet, vessel, foundation, and interaction effects. (orig.)

  19. Heat flux to the helium cryogenic system elements in the case of incidental vacuum vessel ventilation with atmospheric air

    CERN Multimedia

    CERN. Geneva

    2016-01-01

    The selection process for size in safety equipment for cold vessels or process pipes in cryogenic systems should take into consideration the incidental ventilation of the vacuum vessel with atmospheric air. In this case, a significant heat input toward the cold elements of the system can be expected. A number of experimental investigations have been done for the elements at liquid helium temperature which have been covered with 10 layers of MLI. The typical values of the heat flux were measured in a range of 3.7 to 5.0 kW/m2 of the element surface. The helium temperature parts are typically surrounded by thermal shields that are kept in a temperature range of 50-80K. On the external side, the thermal shields are covered with 30-40 layers of MLI while on the internal side, the shields are bare. The theoretical calculations of heat flux to the thermal shield, with respect to the possibility of air condensation and freezing on the bare side of the thermal shield, show that the heat flux to the thermal shield can...

  20. Start-up neutral-beam power supply system for MFTF

    International Nuclear Information System (INIS)

    Mooney, L.J.

    1979-01-01

    This paper describes some of the design features and considerations of the MFTF start-up neutral-beam power supplies. In particular, we emphasize features of the system that will ensure MFTF compatibility and achieve the required reliability/availability for the MFTF to be successful

  1. Computer language evaluation for MFTF SCDS

    International Nuclear Information System (INIS)

    Anderson, R.E.; McGoldrick, P.R.; Wyman, R.H.

    1979-01-01

    The computer languages available for the systems and application implementation on the Supervisory Control and Diagnostics System (SCDS) for the Mirror Fusion Test Facility (MFTF) were surveyed and evaluated. Four language processors, CAL (Common Assembly Language), Extended FORTRAN, CORAL 66, and Sequential Pascal (SPASCAL, a subset of Concurrent Pascal [CPASCAL]) are commercially available for the Interdata 7/32 and 8/32 computers that constitute the SCDS. Of these, the Sequential Pascal available from Kansas State University appears best for the job in terms of minimizing the implementation time, debugging time, and maintenance time. This improvement in programming productivity is due to the availability of a high-level, block-structured language that includes many compile-time and run-time checks to detect errors. In addition, the advanced data-types in language allow easy description of the program variables. 1 table

  2. MFTF plasma diagnostics data acquisition system

    International Nuclear Information System (INIS)

    Davis, G.E.; Coffield, F.E.

    1979-01-01

    The initial goal of the Data Acquisition System (DAS) is to control 11 instruments chosen as the startup diagnostic set and to collect, process, and display the data that these instruments produce. These instruments are described in a paper by Stan Thomas, et. al. entitled ''MFTF Plasma Diagnostics System.'' The DAS must be modular and flexible enough to allow upgrades in the quantity of data taken by an instrument, and also to allow new instruments to be added to the system. This is particularly necessary to support a research project where needs and requirements may change rapidly as a result of experimental findings. Typically, the startup configuration of the diagnostic instruments will contain only a fraction of the planned detectors, and produce approximately one half the data that the expanded version is designed to generate. Expansion of the system will occur in fiscal year 1982

  3. Thermal control for the MFTF magnet

    International Nuclear Information System (INIS)

    Vansant, J.H.; Russ, R.M.

    1980-01-01

    The external dimensions of the Yin-Yang magnet of the Mirror Fusion Test Facility will be 7.8 by 8.5 by 8.5 m, and it will weigh approximately 300 tons. More than 8000 liters of circulating liquid helium will be required to maintain the nearly 50 km of superconductor at below 5.0 K while the latter carries almost 6000 A in a magnetic field of up to nearly 7.7 T. This paper describes several features of the thermal control plans for the Yin-Yang: (1) the proposed cooldown and warmup schedules for the MFTF and the procedure for regenerating external cooling surfaces (2) the design of an external quench resistor based on an estimate of the superconductor's maximum temperature and (3) the use of a computer model of liquid helium circulation in choosing pipe size for the liquid helium lines

  4. Display-management system for MFTF

    International Nuclear Information System (INIS)

    Nelson, D.O.

    1981-01-01

    The Mirror Fusion Test Facility (MFTF) is controlled by 65 local control microcomputers which are supervised by a local network of nine 32-bit minicomputers. Associated with seven of the nine computers are state-of-the-art graphics devices, each with extensive local processing capability. These devices provide the means for an operator to interact with the control software running on the minicomputers. It is critical that the information the operator views accurately reflects the current state of the experiment. This information is integrated into dynamically changing pictures called displays. The primary organizational component of the display system is the software-addressable segment. The segments created by the display creation software are managed by display managers associated with each graphics device. Each display manager uses sophisticated storage management mechanisms to keep the proper segments resident in the local graphics device storage

  5. MFTF supervisory control and diagnostics system hardware

    International Nuclear Information System (INIS)

    Butner, D.N.

    1979-01-01

    The Supervisory Control and Diagnostics System (SCDS) for the Mirror Fusion Test Facility (MFTF) is a multiprocessor minicomputer system designed so that for most single-point failures, the hardware may be quickly reconfigured to provide continued operation of the experiment. The system is made up of nine Perkin-Elmer computers - a mixture of 8/32's and 7/32's. Each computer has ports on a shared memory system consisting of two independent shared memory modules. Each processor can signal other processors through hardware external to the shared memory. The system communicates with the Local Control and Instrumentation System, which consists of approximately 65 microprocessors. Each of the six system processors has facilities for communicating with a group of microprocessors; the groups consist of from four to 24 microprocessors. There are hardware switches so that if an SCDS processor communicating with a group of microprocessors fails, another SCDS processor takes over the communication

  6. 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

  7. Behaviour of tritium in the vacuum vessel of JT-60U

    International Nuclear Information System (INIS)

    Kobayashi, K.; Miya, N.; Ikeda, Y.; Torikai, Y.; Saito, M.; Alimov, V.

    2015-01-01

    The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D 2 (92.8 %) - T 2 (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily

  8. Behaviour of tritium in the vacuum vessel of JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, K.; Miya, N.; Ikeda, Y. [JT-60 Safety Assessment Group, JAEA, Mukoyama (Japan); Torikai, Y. [Hydrogen Isotope Research Center, University of Toyama, Gofuku (Japan); Saito, M.; Alimov, V. [ITER Project Management Group, JAEA, Mukoyama (Japan)

    2015-03-15

    The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D{sub 2} (92.8 %) - T{sub 2} (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily.

  9. 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

  10. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    Kleefeldt, K.

    1985-01-01

    The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m 2 . Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

  11. 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.

  12. Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

    International Nuclear Information System (INIS)

    Di Gironimo, G.; Carfora, D.; Esposito, G.; Lanzotti, A.; Marzullo, D.; Siuko, M.

    2015-01-01

    Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed

  13. ITER-FEAT vacuum vessel and blanket design features and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.; Cardella, A.; Elio, F.; Onozuka, M.; Daenner, W.; Koizumi, K.; Krylov, V.A.

    2001-01-01

    A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R and D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R and D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal. (author)

  14. 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.)

  15. ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme

    Science.gov (United States)

    Ioki, K.; Dänner, W.; Koizumi, K.; Krylov, V. A.; Cardella, A.; Elio, F.; Onozuka, M.; ITER Joint Central Team; ITER Home Teams

    2001-03-01

    A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R&D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.

  16. Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland); Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Esposito, G.; Lanzotti, A.; Marzullo, D. [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Siuko, M. [VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland)

    2015-05-15

    Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed.

  17. Low-level-signal data acquisition for the MFTF superconducting-magnet system

    International Nuclear Information System (INIS)

    Montoya, C.R.

    1981-01-01

    Acquisition of low level signals from sensors mounted on the superconducting yin-yang magnet in the Mirror Fusion Test Facility (MFTF) imposes very strict requirements on the magnet signal conditioning and data acquisition system. Of the various types of sensors required, thermocouples, strain gages, and voltage taps produce very low level outputs. These low level outputs must be accurately measured in the harsh environment of slowly varying magnetic fields, cryogenic temperatures, high vacuum, pulse power and 60 Hz electrical noise, possible neutron radiation, and high common mode voltage resulting from superconducting magnet quench. Successful measurements require careful attention to grounding, shielding, signal handling and processing in the data acquisition system. The magnet instrumentation system provides a means of effectively measuring both low level signals and high level signals from all types of sensors

  18. Low level signal data acquisition for the MFTF-B superconducting magnet system

    International Nuclear Information System (INIS)

    Montoya, C.R.

    1984-01-01

    Acquisition of low level signals from sensors mounted on the superconducting magnets in the Tandem Mirror Fusion Test Facility (MFTF-B) impose very strict requirements on the magnet signal conditioning and data acquisition system. Of the various types of sensors required, thermocouples and strain gages produce very low level outputs. These low level outputs must be accurately measured in the harsh environment of slowly varying magnetic fields, cryogenic temperatures, high vacuum, 80 kV pulse power, 60 Hz, 17 MHz and 28, 35, and 56 GHz electrical noise and possible neutron radiation. Successful measurements require careful attention to grounding, shielding, signal handling and processing in the data acquisition system. The magnet instrumentation system provides a means of effectively measuring both low level signals and high level signals from all types of sensors. Various methods involved in the design and implementation of the system for signal conditioning and data gathering will be presented

  19. Drift orbits in the TMX and MFTF-B tandem mirrors

    International Nuclear Information System (INIS)

    Byers, J.A.

    1982-01-01

    Drift orbits for the TMX and MFTF-B tandem-mirror designs are followed by using a long-thin expansion of the drift equations. Unexpected asymmetries in the field-line curvatures in the yin-yang end-mirror traps, caused by the transition coils between the solenoid and the yin-yang, result in an elliptical distortion of the drift surface with a/b=1.5 at most, a perhaps tolerable deviation from omnigenity. Yushmanov-trapped particles are no worse than the bulk hot particles. Finite-beta plasma fields, coupled to the asymmetric curvature, produce sizeable banana orbits with widths comparable to the plasma radius, but these orbits are possible for only a few of the particles. Details of the transition through resonance in the solenoid are shown, including the banana shapes of the drift surfaces and the disruption of the surface in the stochastic regime. The orbits in the original design for the A-cell of MFTF-B are the most extreme; in the vacuum fields they all have an extended peanut shape that finally closes only at about 3m. This shape is strongly non-omnigenous and suggests a hollow plasma-density profile. Finite-beta B vectorxnablaB drifts can help to minimize the radial extent of these orbits, but the strength of the vacuum curvatures makes omnigenity only marginally possible. Including B vectorxnablaphi drifts makes omnigenity even more unlikely for the ions, for which the B vectorxnablaB and B vectorxnablaphi drifts are of opposite sign, and conversely helps to omnigenize the drift surfaces of the ECRH 200-keV electrons. It is argued that not every class of particles can have good, i.e. near-omnigenous drifts, regardless of the ability of phi(r) to adjust to limit the radial extent of the orbits. This lack of omnigenity leaves one with no theoretical base for describing the MHD equilibrium in the original designs, but a new magnetic field design for MFTF-B A-cell has apparently completely restored omnigenous orbits. (author)

  20. Liquid helium cooling of the MFTF superconducting magnets

    International Nuclear Information System (INIS)

    VanSant, J.H.; Zbasnik, J.P.

    1986-09-01

    During acceptance testing of the Mirror Fusion Test Facility (MFTF), we measured these tests: liquid helium heat loads and flow rates in selected magnets. We used the data from these tests to estimate helium vapor quality in the magnets so that we could determine if adequate conductor cooling conditions had occurred. We compared the measured quality and flow with estimates from a theoretical model developed for the MFTF magnets. The comparison is reasonably good, considering influences that can greatly affect these values. This paper describes the methods employed in making the measurements and developing the theoretical estimates. It also describes the helium system that maintained the magnets at required operating conditions

  1. Axicell MFTF-B superconducting-magnet system

    International Nuclear Information System (INIS)

    Wang, S.T.; Bulmer, R.; Hanson, C.; Hinkle, R.; Kozman, T.; Shimer, D.; Tatro, R.; VanSant, J.; Wohlwend, J.

    1982-01-01

    The Axicell MFTF-B magnet system will provide the field environment necessary for tandem mirror plasma physics investigation with thermal barriers. The performance of the device will stimulate DT to achieve energy break-even plasma conditions. Operation will be with deuterium only. There will be 24 superconducting coils consisting of 2 sets of yin-yang pairs, 14 central-cell solenoids, 2 sets of axicell mirror-coil pairs, and 2 transition coils between the axicell mirror coil-pairs and the yin-yang coils. This paper describes the progress in the design and construction of MFTF-B Superconducting-Magnet System

  2. 1000 kW ICRH amplifiers for MFTF-B

    International Nuclear Information System (INIS)

    Boksberger, U.

    1986-01-01

    For the startup of the MFTF-B ICRH heating will be applied. Two commercial amplifiers derived from standard broadcast transmitters provide 1000 kW RF power each into a matching system for any VSWR as high as 1.5. Emphasis is put on the specific environment of magnetic fields and seismic loads as well as to the particular RF power control requirements and remote operation. Also addressed is the amplifier's performance into a typical load. The load variations due to the MFTF-B plasma coupling were calculated by TRW

  3. Protection of the MFTF accel power supplies

    International Nuclear Information System (INIS)

    Wilson, J.H.; Wood, J.C.

    1979-01-01

    The MFTF experiment's Sustaining Neutral Beam Power Supply System (SNBPSS) includes twenty-four 95 kV, 80 A accel dc power supplies (ADCPS). Each power supply includes a relatively high-impedance (20 percent) rectifier transformer and a step voltage regulator with a 50-100 percent voltage range. With this combination, the fault current for some postulated faults may be lower than the supply's full load current at maximum voltage. A design has been developed which uses protective relays and current-limiting fuses coordinated to detect phase and ground faults, DC faults, incorrect voltage conditions, rectifier faults, power factor correction capacitor faults, and overloads. This unusual solution ensures fast tripping on potentially destructive high-current faults and long-time delays at lower currents to allow 30 second pulse operation. The ADCPS meets the LLL specification that all major assemblies be self-protecting, that is, able to sustain external faults without damage to minimize damage due to internal faults

  4. Weld defects analysis of 60 mm thick SS316L mock-ups of TIG and EB welds by ultrasonic inspection for fusion reactor vacuum vessel applications

    International Nuclear Information System (INIS)

    Buddu, Ramesh Kumar; Shaikh, Shamsuddin; Raole, P.M.; Sarkar, B.

    2015-01-01

    The present paper reports the weld quality inspections carried with 60 mm thick AISI welds of SS316L. The high thickness steel plates requirement is due to the specific applications in case of advanced fusion reactor structural components like vacuum vessel and others. Different kind welds are proposed for the thick plate joints like Tungsten Inert Gas (TIG) welding, Electron beam welding as per stringent conditions (like very low distortions and residual stresses) for the vacuum vessel fabrication. Mock-ups of laboratory scale welds are fabricated by TIG (multi-pass) and EB (double pass) process techniques and different weld quality inspections are carried by different NDT tests. The welds are examined with Liquid penetrant examination to check sub surface cracks/discontinuities towards the defects observation

  5. Proposal to negotiate an amendment to an existing contract for the supply of additional vacuum vessels for the short-straight sections of the LHC

    CERN Document Server

    2005-01-01

    This document concerns the proposal to negotiate an amendment to an existing contract for the supply of additional vacuum vessels for the short-straight sections of the LHC. For the reasons explained in this document, the Finance Committee is invited to approve an amendment to an existing contract with SLOVENSKE ENERGETICKE STROJARNE (SK) for the supply of 25 additional vacuum vessels for the short-straight sections for the LHC for an amount of 140 000 euros (216 999 Swiss francs), subject to revision for inflation, bringing the total to a maximum amount of 6 176 855 euros (9 574 097 Swiss francs), subject to revision for inflation. The rate of exchange used is that stipulated in the tender.

  6. Vacuum Technology

    Energy Technology Data Exchange (ETDEWEB)

    Biltoft, P J

    2004-10-15

    The environmental condition called vacuum is created any time the pressure of a gas is reduced compared to atmospheric pressure. On earth we typically create a vacuum by connecting a pump capable of moving gas to a relatively leak free vessel. Through operation of the gas pump the number of gas molecules per unit volume is decreased within the vessel. As soon as one creates a vacuum natural forces (in this case entropy) work to restore equilibrium pressure; the practical effect of this is that gas molecules attempt to enter the evacuated space by any means possible. It is useful to think of vacuum in terms of a gas at a pressure below atmospheric pressure. In even the best vacuum vessels ever created there are approximately 3,500,000 molecules of gas per cubic meter of volume remaining inside the vessel. The lowest pressure environment known is in interstellar space where there are approximately four molecules of gas per cubic meter. Researchers are currently developing vacuum technology components (pumps, gauges, valves, etc.) using micro electro mechanical systems (MEMS) technology. Miniature vacuum components and systems will open the possibility for significant savings in energy cost and will open the doors to advances in electronics, manufacturing and semiconductor fabrication. In conclusion, an understanding of the basic principles of vacuum technology as presented in this summary is essential for the successful execution of all projects that involve vacuum technology. Using the principles described above, a practitioner of vacuum technology can design a vacuum system that will achieve the project requirements.

  7. MFTF-B PACE tests and final cost report

    International Nuclear Information System (INIS)

    Krause, K.H.; Kozman, T.A.; Smith, J.L.; Horan, R.J.

    1986-10-01

    The Mirror Fusion Test Facility (MFTF-B) construction project was successfully completed in February 1986, with the conclusion of the Plant and Capital Equipment (PACE) Tests. This series of tests, starting in September 1985 and running through February 1986, demonstrated the overall machine capabilities and special facilities accomplishments for the Mirror Fusion Test Facility Project

  8. Axicell design for the end plugs of MFTF-B

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Karpenko, V.N.

    1982-01-01

    Certain changes in the end-plug design in the Mirror Fusion Test Facility (MFTF-B) are described. The Laboratory (LLNL) proposes to implement these changes as soon as possible in order to construct the machine in an axicell configuration. The present physics and technology goals as well as the project cost and schedule will not be affected by these changes

  9. Mechanical considerations for MFTF-B plasma-diagnostic system

    International Nuclear Information System (INIS)

    Thomas, S.R. Jr.; Wells, C.W.

    1981-01-01

    The reconfiguration of MFTF to a tandem mirror machine with thermal barriers has caused a significant expansion in the physical scope of plasma diagnostics. From a mechanical perspective, it complicates the plasma access, system interfaces, growth and environmental considerations. Conceptual designs characterize the general scope of the design and fabrication which remains to be done

  10. MFTF-α+T end plug magnet design

    International Nuclear Information System (INIS)

    Srivastava, V.C.; O'Toole, J.A.

    1983-01-01

    The conceptual design of the end-plug magnets for MFTF-α+T is described. MFTF-α+ T is a near-term upgrade of MFTF-B, which features new end plugs to improve performance. The Fusion Engineering Design Center has performed the engineering design of MFTF-α+T under the overall direction of Lawrence Livermore National Laboratory. Each end plug consists of two Yin-Yang pairs, each with approx.2.5:1 mirror ratio and approx.5-T peak field on axis; two transition coils; and a recircularizing solenoid. This paper describes the end-plug magnet system functional requirements and presents a conceptual design that meets them. The peak field at the windings of the end-plug coils is approx.6-T. These coils are designed using the NbTi MFTF-B conductor and cooled by a 4.2K liquid helium bath. All the end-plug magnets are designed to operate in the cryostable mode with adequate quench protection for safety. Shielding requirements are stated and a summary of heat loads is provided. Field and force calculations are discussed. The field on axis is shown to meet the functional requirements. Force resultants are reported in terms of winding running loads and resultant coil forces are also given. The magnet structural support is described. A trade study to determine the optimum end-cell coil internal nuclear shield thickness and the resulting coil size based on minimizing the end-cell life cycle cost is summarized

  11. Software protocol design: Communication and control in a multi-task robot machine for ITER vacuum vessel assembly and maintenance

    International Nuclear Information System (INIS)

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

    2015-01-01

    applying these protocols, the software for a multi-task robot machine that is used for ITER vacuum vessel assembly and maintenance has been developed and it is demonstrated that machining tasks of the robot machine, such as milling, drilling, welding etc., can be implemented in both an individual and composite way.

  12. Software protocol design: Communication and control in a multi-task 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); Wang, Yongbo [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland)

    2015-10-15

    applying these protocols, the software for a multi-task robot machine that is used for ITER vacuum vessel assembly and maintenance has been developed and it is demonstrated that machining tasks of the robot machine, such as milling, drilling, welding etc., can be implemented in both an individual and composite way.

  13. Application of structural-mechanics methods to the design of large tandem-mirror fusion devices (MFTF-B). Revision 1

    International Nuclear Information System (INIS)

    Karpenko, V.N.; Ng, D.S.

    1985-01-01

    The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory requires state-of-the-art structural-mechanics methods to deal with access constraints for plasma heating and diagnostics, alignment requirements, and load complexity and variety. Large interactive structures required an integrated analytical approach to achieve a reasonable level of overall system optimization. The Tandem Magnet Generator (TMG) creates a magnet configuration for the EFFI calculation of electromagnetic-field forces that, coupled with other loads, form the input loading to magnet and vessel finite-element models. The analytical results provide the data base for detailed design of magnet, vessel, foundation, and interaction effects. 13 refs

  14. Thermofluid experiments for Fusion Reactor Safety. Visualization of exchange flows through breaches of a vacuum vessel in a fusion reactor under the LOVA condition

    International Nuclear Information System (INIS)

    Fujii, Sadao; Shibazaki, Hiroaki; Takase, Kazuyuki; Kunugi, Tomoaki.

    1997-01-01

    Exchange flow rates through breaches of a vacuum vessel in a fusion reactor under the LOVA (Loss of VAcuum event) conditions were measured quantitatively by using a preliminary LOVA apparatus and exchange flow patterns over the breach were visualized qualitatively by smoke. Velocity distributions in the exchange flows were predicted from the observed flow patterns by using the correlation method in the flow visualization procedures. Mean velocities calculated from the predicted velocity distributions at the outside of the breach were in good agreement with the LOVA experimental results when the exchange flow velocities were low. It was found that the present flow visualization and the image processing system might be an useful procedure to evaluate the exchange flow rates. (author)

  15. Mechanical properties and microstructural investigations of TIG welded 40 mm and 60 mm thick SS 316L samples for fusion reactor vacuum vessel applications

    Energy Technology Data Exchange (ETDEWEB)

    Buddu, Ramesh Kumar, E-mail: brkumar75@gmail.com; Chauhan, N.; Raole, P.M.

    2014-12-15

    Highlights: • Austenitic stainless steels (316L) of 40 mm and 60 mm thickness plates were joined by Tungsten Inert Gas welding (TIG) process which are probable materials for advanced fusion reactor vacuum vessel requirements. • Mechanical properties and detailed microstructure studies have been carried out for welded samples. • Fractography analysis of impact test specimens indicated ductile fracture mode in BM, HAZ and WZ samples. • Presence of delta ferrite phase was observed in the welded zone and ferrite number data was measured for the base and weld metal and was found high in welds. - Abstract: The development of advanced fusion reactors like DEMO will have various challenges in materials and fabrication. The vacuum vessel is important part of the fusion reactor. The double walled design for vacuum vessel with thicker stainless steel material (40–60 mm) has been proposed in the advanced fusion reactors like ITER. Different welding techniques will have to be used for such vacuum vessel development. The required mechanical, structural and other properties of stainless steels have to be maintained in these joining processes of components of various shapes and sizes in the form of plates, ribs, shells, etc. The present paper reports characterization of welding joints of SS316L plates with higher thicknesses like 40 mm and 60 mm, prepared using multi-pass Tungsten Inert Gas (TIG) welding process. The weld quality has been evaluated with non-destructive tests by X-ray radiography and ultrasonic methods. The mechanical properties like tensile, bend tests, Vickers hardness and impact fracture tests have been carried out for the weld samples. Tensile property test results indicate sound weld joints with efficiencies over 100%. Hardening was observed in the weld zone in non-uniform manner. Macro and microstructure studies have been carried out for Base Metal (BM), Heat Affected Zone (HAZ) and Weld Zone (WZ). Scanning Electron Microscopy (SEM) analysis carried

  16. Mechanical properties and microstructural investigations of TIG welded 40 mm and 60 mm thick SS 316L samples for fusion reactor vacuum vessel applications

    International Nuclear Information System (INIS)

    Buddu, Ramesh Kumar; Chauhan, N.; Raole, P.M.

    2014-01-01

    Highlights: • Austenitic stainless steels (316L) of 40 mm and 60 mm thickness plates were joined by Tungsten Inert Gas welding (TIG) process which are probable materials for advanced fusion reactor vacuum vessel requirements. • Mechanical properties and detailed microstructure studies have been carried out for welded samples. • Fractography analysis of impact test specimens indicated ductile fracture mode in BM, HAZ and WZ samples. • Presence of delta ferrite phase was observed in the welded zone and ferrite number data was measured for the base and weld metal and was found high in welds. - Abstract: The development of advanced fusion reactors like DEMO will have various challenges in materials and fabrication. The vacuum vessel is important part of the fusion reactor. The double walled design for vacuum vessel with thicker stainless steel material (40–60 mm) has been proposed in the advanced fusion reactors like ITER. Different welding techniques will have to be used for such vacuum vessel development. The required mechanical, structural and other properties of stainless steels have to be maintained in these joining processes of components of various shapes and sizes in the form of plates, ribs, shells, etc. The present paper reports characterization of welding joints of SS316L plates with higher thicknesses like 40 mm and 60 mm, prepared using multi-pass Tungsten Inert Gas (TIG) welding process. The weld quality has been evaluated with non-destructive tests by X-ray radiography and ultrasonic methods. The mechanical properties like tensile, bend tests, Vickers hardness and impact fracture tests have been carried out for the weld samples. Tensile property test results indicate sound weld joints with efficiencies over 100%. Hardening was observed in the weld zone in non-uniform manner. Macro and microstructure studies have been carried out for Base Metal (BM), Heat Affected Zone (HAZ) and Weld Zone (WZ). Scanning Electron Microscopy (SEM) analysis carried

  17. Control and diagnostic data structures for the MFTF

    International Nuclear Information System (INIS)

    Wade, J.A.; Choy, J.H.

    1979-01-01

    A Data Base Management System (DBMS) is being written as an integral part of the Supervisory Control and Diagnostics System (SCDS) of programs for control of the Mirror Fusion Test Facility (MFTF). The data upon which the DBMS operates consist of control values and evaluative information required for facilities control, along with control values and disgnostic data acquired as a result of each MFTF shot. The user interface to the DBMS essentially consists of two views: a computer program interface called the Program Level Interface (PLI) and a stand-alone interactive program called the Query Level Interface to support terminal-based queries. This paper deals specifically with the data structure capabilities from the viewpoint of the PLI user

  18. Progress on axicell MFTF-B superconducting magnet systems

    International Nuclear Information System (INIS)

    Wang, S.T.; Kozman, T.A.; Hanson, C.L.; Shimer, D.W.; VanSant, J.H.; Zbasnik, J.

    1983-01-01

    Since the entire Mirror Fusion Test Facility (MFTF-B) Magnet System was reconfigured from the original A-cell to an axicell design, much progress has been made on the design, fabrication, and installation planning. The axicell MFTF-B magnet array consists of a total of 26 large superconducting main coils. This paper provides an engineering overview of the progress of these coils. Recent studies on the effects of field errors on the plasma at the recircularizing region (transition coils) show that small field errors will generate large displacements of the field lines. These field errors might enhance radial electron heat transport and deteriorate the plasma confinement. Therefore, 16 superconducting trim coils have been designed to correct the coil misalignments. Progress of the trim coils are reported also

  19. ASME Section VIII Recertification of a 33,000 Gallon Vacuum-jacketed LH2 Storage Vessel for Densified Hydrogen Testing at NASA Kennedy Space Center

    Science.gov (United States)

    Swanger, Adam M.; Notardonato, William U.; Jumper, Kevin M.

    2015-01-01

    The Ground Operations Demonstration Unit for Liquid Hydrogen (GODU-LH2) has been developed at NASA Kennedy Space Center in Florida. GODU-LH2 has three main objectives: zero-loss storage and transfer, liquefaction, and densification of liquid hydrogen. A cryogenic refrigerator has been integrated into an existing, previously certified, 33,000 gallon vacuum-jacketed storage vessel built by Minnesota Valley Engineering in 1991 for the Titan program. The dewar has an inner diameter of 9.5 and a length of 71.5; original design temperature and pressure ranges are -423 F to 100 F and 0 to 95 psig respectively. During densification operations the liquid temperature will be decreased below the normal boiling point by the refrigerator, and consequently the pressure inside the inner vessel will be sub-atmospheric. These new operational conditions rendered the original certification invalid, so an effort was undertaken to recertify the tank to the new pressure and temperature requirements (-12.7 to 95 psig and -433 F to 100 F respectively) per ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. This paper will discuss the unique design, analysis and implementation issues encountered during the vessel recertification process.

  20. Physics conceptual design for the MFTF-B transition coil

    International Nuclear Information System (INIS)

    Baldwin, D.E.; Bulmer, R.H.

    1982-01-01

    The physics constraints related to finite-β equilibria, β limits due to curvature-driven MHD modes, and ion transport in the central cell. These physics constraints had to be satisfied subject to certain non-physics constraints. Principal among these were the geometric and structural features of the existing MFTF-B magnet set and the required access for neutral beams for pumping. These constraints and their origins are discussed

  1. Manufacturing and quality assurance for the MFTF superconductor core

    International Nuclear Information System (INIS)

    Scanlan, R.M.; Johnston, J.E.; Waide, P.A.; Zeitlin, B.A.; Smith, G.B.; Nelson, C.T.

    1979-01-01

    A total of 55,000 m of multifilamentary Nb-Ti superconductor in minimum lengths of 380 m are required for the Mirror Fusion Test Facility. This conductor is a large cross-section monolith and, as such, has presented several new manufacturing challenges. In addition, a monolith requires more stringent quality assurance procedures than braids or cables. This paper describes the manufacturing steps and the quality assurance program which have been developed for the MFTF superconductor core

  2. Model approach for simulating the thermodynamic behavior of the MFTF cryogenic cooling systems - a status report

    International Nuclear Information System (INIS)

    Sutton, S.B.; Stein, W.; Reitter, T.A.; Hindmarsh, A.C.

    1983-01-01

    A numerical model for calculating the thermodynamic behavior of the MFTF-B cryogenic cooling system is described. Nine component types are discussed with governing equations given. The algorithm for solving the coupled set of algebraic and ordinary differential equations is described. The model and its application to the MFTF-B cryogenic cooling system has not been possible due to lack of funding

  3. Analysis of the effect of the Electron-Beam welding sequence for a fixed manufacturing route using finite element simulations applied to ITER vacuum vessel manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Martín-Menéndez, Cristina, E-mail: cristina@natec-ingenieros.com [Numerical Analysis Technologies, S.L. Marqués de San Esteban No. 52, 33206 Gijón (Spain); Rodríguez, Eduardo [Department of Mechanical Engineering, University of Oviedo, Campus de Gijón, 33203 Gijón (Spain); Ottolini, Marco [Ansaldo Nucleare S.p.A., Corso Perrone 25, 16152 Genova (Italy); Caixas, Joan [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 No. 52, 33206 Gijón (Spain)

    2016-03-15

    Highlights: • The simulation methodology employed in this paper is able to adapt inside a complex manufacturing route. • The effect of the sequence is lower in a highly constrained assembly than in a lowly constrained one. • The most relevant influence on the distortions is the jigs design, instead of the welding sequence. • The welding distortion analysis should be used as a guidance to design and improve the manufacturing strategy. - Abstract: The ITER Vacuum Vessel Sectors have very tight tolerances and high density of welding. Therefore, prediction and reduction of welding distortion are critical to allow the final assembly with the other Vacuum Vessel Sectors without the production of a full scale prototype. In this paper, the effect of the welding sequence in the distortions inside a fixed manufacturing route and in a highly constrained assembly is studied in the poloidal segment named inboard (PS1). This is one of the four poloidal segments (PS) assembled for the sector. Moreover, some restrictions and limitations in the welding sequence related to the manufacturing process are explained. The results obtained show that the effect of the sequence is lower in a highly constrained assembly than in a low constrained one. A prototype manufactured by AMW consortium (PS1 mock-up) is used in order to validate the finite element method welding simulation employed. The obtained results confirmed that for Electron-Beam welds, both the welding simulation and the mock-up show a low value of distortions.

  4. Field-reversal experiments in the mirror fusion test facility (MFTF)

    International Nuclear Information System (INIS)

    Shearer, J.W.; Condit, W.C.

    1977-01-01

    Detailed consideration of several aspects of a field-reversal experiment was begun in the Mirror Fusion Test Facility (MFTF): Model calculations have provided some plausible parameters for a field-reversed deuterium plasma in the MFTF, and a buildup calculation indicates that the MFTF neutral-beam system is marginally sufficient to achieve field reversal by neutral injection alone. However, the many uncertainties indicate the need for further research and development on alternate buildup methods. A discussion of experimental objectives is presented and important diagnostics are listed. The range of parameter space accessible with the MFTF magnet design is explored, and we find that with proper aiming of the neutral beams, meaningful experiments can be performed to advance toward these objectives. Finally, it is pointed out that if we achieve enhanced n tau confinement by means of field reversal, then quasi-steady-state operation of MFTF is conceivable

  5. Mirror Fusion vacuum technology developments

    International Nuclear Information System (INIS)

    Batzer, T.H.; Call, W.R.

    1983-01-01

    Magnetic Mirror Fusion experiments, such as MFTF-B+T (Mirror Fusion Test Facility-B, Tritium Upgrade) and foreseeable follow-on devices, have operational and maintenance requirements that have not yet been fully demonstrated. Among those associated with vacuum technology are the very-high continuous-pumping speeds, 10 7 to 10 8 l/s for D 2 , T 2 and, to a lesser extent, He; the early detection of water leaks from the very-high heat-flux neutral-beam dumps and the detection and location of leaks in the superconducting magnets not protected by guard vacuums. Possible solutions to these problems have been identified and considerable progress has been made toward successfully demonstrating their feasibility

  6. Mirror fusion vacuum technology developments

    International Nuclear Information System (INIS)

    Batzer, T.H.; Call, W.R.

    1983-01-01

    Magnetic Mirror Fusion experiments, such as MFTF-B+T (Mirror Fusion Test Facility-B, Tritium Upgrade) and foreseeable follow-on devices, have operational and maintenance requirements that have not yet been fully demonstrated. Among those associated with vacuum technology are the very-high continuous-pumping speeds, 10 7 to 10 8 l/s for D 2 , T 2 and, to a lesser extent, He; the early detection of water leaks from the very-high heat-flux neutral-beam dumps and the detection and location of leaks in the superconducting magnets not protected by guard vacuums. Possible solutions to these problems have been identified and considerable progress has been made toward successfully demonstrating their feasibility

  7. 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.

  8. Maintenance and availability considerations for MFTF-B upgrade

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1983-01-01

    The upgrade of the Mirror Fusion Test Facility (MFTF-B) tandem mirror device incorporates the operation of advanced systems plus the requirement for remote maintenance. To determine if the operating availability goal of this device is achievable, an assessment of component lifetimes was made, along with estimates of device downtime. Key subsystem components were considered from the magnet, heating, impurity control, pumping, and test module systems. Component replacements were grouped into three categories, and a lifetime operating plan, including component replacements, was developed. It was determined that this device could achieve a 10% operating availability

  9. Electrical supply for MFTF-B superconducting magnet system

    International Nuclear Information System (INIS)

    Shimer, D.W.; Owen, E.W.

    1985-01-01

    The MFTF-B magnet system consists of 42 superconducting magnets which must operate continuously for long periods of time. The magnet power supply system is designed to meet the operational requirements of accuracy, flexibility, and reliability. The superconducting magnets require a protection system to protect against critical magnet faults of quench, current lead overtemperature, and overcurrent. The protection system is complex because of the large number of magnets, the strong coupling between magnets, and the high reliability requirement. This paper describes the power circuits and the components used in the design

  10. New kind of user interface for controlling MFTF diagnostics

    International Nuclear Information System (INIS)

    Preckshot, G.G.; Saroyan, R.A.; Mead, J.E.

    1983-01-01

    The Mirror Fusion Test Facility (MFTF) at Lawrence Livermore National Laboratory is faced with the problem of controlling a multitude of plasma diagnostics instruments from a central, multiprocessor computer facility. A 16-bit microprocessor-based workstation allows each physicist entree into the central multiprocessor, which consists of nine Perkin-Elmer 32-bit minicomputers. The workstation provides the user interface to the larger system, with display graphics, windowing, and a physics notebook. Controlling a diagnostic is now equivalent to making entries into a traditional physics notebook

  11. A new kind of user interface for controlling MFTF diagnostics

    International Nuclear Information System (INIS)

    Preckshot, G.; Mead, J.; Saroyan, R.

    1983-01-01

    The Mirror Fusion Test Facility (MFTF) at Lawrence Livermore National Laboratory is faced with the problem of controlling a multitude of plasma diagnostics instruments from a central, multiprocessor computer facility. A 16-bit microprocessor-based workstation allows each physicist entree into the central multiprocessor, which consists of nine Perkin-Elmer 32-bit minicomputers. The workstation provides the user interface to the larger system, with display graphics, windowing, and a physics notebook. Controlling a diagnostic is now equivalent to making entries into a traditional physics notebook

  12. Alternatives for contaminant control during MFTF plasma buildup

    International Nuclear Information System (INIS)

    Khan, J.M.; Valby, L.E.

    1979-01-01

    The MFTF mirror device considers all low-energy species to be contaminants, since their primary effect is to erode the plasma boundary by charge-exchange reactions. Confinement for other than hydrogen isotypes is far from complete and confinement time is hardly more than transit time from the source to the end wall. The brevity of the confinement time makes it all the more necessary to prevent any contamination which might further reduce it. At Livermore, the historical solution to contaminant control has been to evaporate titanium onto cold surfaces. An alternative to this approach and its implications are considered

  13. Design of the electromagnetic fluctuations diagnostic for MFTF-B

    International Nuclear Information System (INIS)

    House, P.A.; Goerz, D.A.; Martin, R.

    1983-01-01

    The Electromagnetic Fluctuations (EMF) diagnostic will be used to monitor ion fluctuations which could be unstable in MFTF-B. Each probe assembly includes a high impedance electrostatic probe to measure potential fluctuations, and a group of nested, single turn loops to measure magnetic fluctuations in three directions. Eventually, more probes and loops will be added to each probe assembly for making more detailed measurements. The sensors must lie physically close to the plasma edge and are radially positionable. Also, probes at separate axial locations can be positioned to connect along the same magnetic field line. These probes are similar in concept to the rf probes used on TMX, but the high thermal load for 30-second shots on MFTF-B requires a water-cooled design along with temperature monitors. Each signal channel has a bandwidth of .001 to 150 MHz and is monitored by up to four different data channels which obtain amplitude and frequency information. This paper describes the EMF diagnostic and presents the detailed mechanical and electrical designs

  14. User interface on networked workstations for MFTF plasma diagnostic instruments

    International Nuclear Information System (INIS)

    Renbarger, V.L.; Balch, T.R.

    1985-01-01

    A network of Sun-2/170 workstations is used to provide an interface to the MFTF-B Plasma Diagnostics System at Lawrence Livermore National Laboratory. The Plasma Diagnostics System (PDS) is responsible for control of MFTF-B plasma diagnostic instrumentation. An EtherNet Local Area Network links the workstations to a central multiprocessing system which furnishes data processing, data storage and control services for PDS. These workstations permit a physicist to command data acquisition, data processing, instrument control, and display of results. The interface is implemented as a metaphorical desktop, which helps the operator form a mental model of how the system works. As on a real desktop, functions are provided by sheets of paper (windows on a CRT screen) called worksheets. The worksheets may be invoked by pop-up menus and may be manipulated with a mouse. These worksheets are actually tasks that communicate with other tasks running in the central computer system. By making entries in the appropriate worksheet, a physicist may specify data acquisition or processing, control a diagnostic, or view a result

  15. Options for axisymmetric operation of MFTF-B

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Devoto, R.S.; Thomassen, K.I.

    1986-01-01

    The flexibility of MFTF-B for axisymmetric experiments has been investigated. Interhcanging the axicell coils and increasing their separation results in an axisymmetric plug cell with 12:1 and 6:1 inner and outer mirror ratios, respectively. For axisymmetric operation, the sloshing-ion neutral beams, ECRH gyrotrons, and the pumping system would be moved to the axicell. Stabilization by E-rings could be explored in this configuration. With the addition of octopole magnets, off-axis multipole stabilization could also be tested. Operating points for octopole and E-ring-stabilized configurations with properties similar to those of the quadrupole MFTF-B, namely T/sub ic/ = 10 - 15 keV and n/sub c/ approx. = 3 x 10 13 cm -3 , have been obtained. Because of the negligible radial transport of central-cell ions, the required neutral-beam power in the central cell has been dramatically reduced. In addition, because MHD stabilization is achieved by off-axis hot electrons in both cases, much lower barrier beta is possible, which aids in reducing the barrier ECRH power. Total ECRH power in the end cell is projected to be approx. =1 MW. Possible operating points for both octopole and E-ring configurations are described along with the stability considerations involved

  16. 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.

  17. Design and fabrication of the MFTF-B magnet system

    International Nuclear Information System (INIS)

    Tatro, R.E.; Kozman, T.A.

    1985-09-01

    The MFTF-B superconducting magnet system consists of 40 NbTi magnets and two Nb 3 Sn magnets. General Dynamics (GD) designed all magnets except for the small trim coils. GD then fabricated 20 NbTi magnets, while LLNL fabricated 20 NbTi magnets and two Nb 3 Sn magnets. The design phase was completed in February 1984 and included the competitive procurement of magnet structural fabrication, superconductor, G-10CR insulation, support struts and bearings, vapor-cooled leads, and thermal shields for all magnets. Fabrication of all magnets was completed in March 1985. At GD, dual assembly lines were necessary during fabrication in order to meet the aggressive LLNL schedule. The entire magnet system has been installed and aligned at LLNL, and Tech Demo tests will be performed during September-November 1985

  18. Data triggered data processing at MFTF-B

    International Nuclear Information System (INIS)

    Jackson, R.J.; Balch, T.R.; Preckshot, G.G.

    1985-01-01

    A primary characteristic of most batch systems is that the input data files must exist before jobs are scheduled. On the Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory we schedule jobs to process experimental data to be collected during a five minute shot cycle. Our data-driven processing system emulates a coarsely granular data flow architecture. Processing jobs are scheduled before the experimental data is collected. Processing jobs ''fire'', or execute, as input data becomes available. Similar to UNIX ''pipes'', data produced by upstream processing nodes may be used as inputs by following nodes. Users, working on our networked SUN workstations, specify data processing templates which define processes and their data dependencies. Data specifications indicate the source of data; actual associations with specific data instantiations are made when the jobs are scheduled. We report here on details of diagnostic data processing and our experiences

  19. Alternative connections for the large MFTF-B solenoids

    International Nuclear Information System (INIS)

    Owen, E.W.; Shimer, D.W.; Wang, S.T.

    1983-01-01

    The MFTF-B central-cell solenoids are a set of twelve closely coupled, large superconducting magnets with similar but not exactly equal currents. Alternative methods of connecting them to their power supplies and dump resistors are investigated. The circuits are evaluated for operating conditions and fault conditions. The factors considered are the voltage to ground during a dump, short circuits, open circuits, quenches, and failure of the protection system to detect a quench. Of particular interest are the current induced in coils that remain superconducting when one or more coils quench. The alternative connections include separate power supplies, combined power supplies, individual dump resistors, series dump resistors and combinations of these. A new circuit that contains coupling resistors is proposed. The coupling resistors do not affect normal fast dumps but reduce the peak induced currents while also reducing the energy rating of the dump resistors. Another novel circuit, the series circuit with diodes, is discussed in detail

  20. Overview of MFTF supervisory control and diagnostics system software

    International Nuclear Information System (INIS)

    Ng, W.C.

    1979-01-01

    The Mirror Fusion Test Facility (MFTF) at the Lawrence Livermore Laboratory (LLL) is currently the largest mirror fusion research project in the world. Its Control and Diagnostics System is handled by a distributed computer network consisting of nine Interdata minicomputer systems and about 65 microprocessors. One of the design requirements is tolerance of single-point failure. If one of the computer systems becomes inoperative, the experiment can still be carried out, although the system responsiveness to operator command may be degraded. In a normal experiment cycle, the researcher can examine the result of the previous experiment, change any control parameter, fire a shot, collect four million bytes of diagnostics data, perform intershot analysis, and have the result presented - all within five minutes. The software approach adopted for the Supervisory Control and Diagnostics System features chief programmer teams and structured programming. Pascal is the standard programming language in this project

  1. Improvement in MFTF data base system response times

    International Nuclear Information System (INIS)

    Lang, N.C.; Nelson, B.C.

    1983-01-01

    The Supervisory Control and Diagnostic System for the Mirror Fusion Test Facility (MFTF) has been designed as an event driven system. To this end we have designed a data base notification facility in which a task can request that it be loaded and started whenever an element in the data base is changed beyond some user defined range. Our initial implementation of the notify facility exhibited marginal response times whenever a data base table with a large number of outstanding notifies was written into. In this paper we discuss the sources of the slow response and describe in detail a new structure for the list of notifies which minimizes search time resulting in significantly faster response

  2. MFTF 230 kV pulsed power substation

    International Nuclear Information System (INIS)

    Wilson, J.H.

    1979-01-01

    The Mirror Fusion Test Facility (MFTF) currently under construction at the Lawrence Livermore Laboratory includes a Sustaining Neutral Beam Power Supply System (SNBPSS) consisting of 24 power-supply sets. The System will operate in long pulses (initially .5 seconds and eventually 30 seconds) at high power (200 MW), which will necessitate a large source of ac power. To meet this requirement, a new 230-kV substation is also being built at LLL. The constraints of cost, equipment protection, short operating lifetime (10 years), and reliability dictated a unique substation design. Its unusual features include provisions for fast fault detection and tripping, a capability for limiting ground fault current, low impedance, and economical design

  3. High field Nb/sub 3/Sn Axicell insert coils for the Mirror Fusion Test Facility-B (MFTF-B) axicell configuration. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Baldi, R.W.; Tatro, R.E.; Scanlan, R.M.; Agarwal, K.L.; Bailey, R.E.; Burgeson, J.E.; Kim, I.K.; Magnuson, G.D.; Mallett, B.D.; Pickering, J.L.

    1984-03-01

    Two 12-tesla superconducting insert coils are being designed by General Dynamics Convair Division for the axicell regions of MFTF-B for Lawrence Livermore National Laboratory. A major challenge of this project is to ensure that combined fabrication and operational strains induced in the conductor are within stringent limitations of the relatively brittle Nb/sub 3/Sn superconductor filaments. These coils are located in the axicell region of MFTF-B. They have a clear-bore diameter of 36.195cm (14.25 inches) and consist of 27 double pancakes (i.e., 54 pancakes per coil) would on an electrically insulated 304LN stainless steel/bobbin helium vessel. Each pancake has 57 turns separated by G-10CR insulation. The complete winding bundle has 4.6 million ampere-turns and uniform current density of 2007 A/cm/sup 2/. In conjunction with the other magnets in the system, they produce a 12-tesla central field and a 12.52-tesla peak field. A multifilamentary Nb/sub 3/Sn conductor was selected to meet these requirements. The conductor consists of a monolithic insert soldered into a copper stabilizer. Sufficient cross-sectional area and work-hardening of the copper stabilizer has been provided for the conductor to self-react the electromagnetic Lorentz force induced hoop stresses with normal operational tensile strains less than 0.07 percent.

  4. High field Nb3Sn Axicell insert coils for the Mirror Fusion Test Facility-B (MFTF-B) axicell configuration. Final report

    International Nuclear Information System (INIS)

    Baldi, R.W.; Tatro, R.E.; Scanlan, R.M.

    1984-03-01

    Two 12-tesla superconducting insert coils are being designed by General Dynamics Convair Division for the axicell regions of MFTF-B for Lawrence Livermore National Laboratory. A major challenge of this project is to ensure that combined fabrication and operational strains induced in the conductor are within stringent limitations of the relatively brittle Nb 3 Sn superconductor filaments. These coils are located in the axicell region of MFTF-B. They have a clear-bore diameter of 36.195cm (14.25 inches) and consist of 27 double pancakes (i.e., 54 pancakes per coil) would on an electrically insulated 304LN stainless steel/bobbin helium vessel. Each pancake has 57 turns separated by G-10CR insulation. The complete winding bundle has 4.6 million ampere-turns and uniform current density of 2007 A/cm 2 . In conjunction with the other magnets in the system, they produce a 12-tesla central field and a 12.52-tesla peak field. A multifilamentary Nb 3 Sn conductor was selected to meet these requirements. The conductor consists of a monolithic insert soldered into a copper stabilizer. Sufficient cross-sectional area and work-hardening of the copper stabilizer has been provided for the conductor to self-react the electromagnetic Lorentz force induced hoop stresses with normal operational tensile strains less than 0.07 percent

  5. Structural analysis interpretation task for the magnet system for Mirror Fusion Test Facility (MFTF)

    International Nuclear Information System (INIS)

    Baldi, R.W.

    1979-11-01

    The primary objective of this study was to develop recommendations to improve and substantiate the structural integrity of the highly stresses small radius region of the MFTF magnet. The specific approach is outlined: (1) Extract detail stress/strain data from General Dynamics Convair Finite-Element Refinement Analysis. (2) Diagram local plate stress distribution and its relationship to the adjacent weldment. (3) Update the parametric fracture mechanics analysis using most recent MFTF related data developed by National Bureau of Standards. (4) Review sequence and assembly as modified by Chicago Bridge and Iron for adaptability to refinements. (5) Investigate the need for fillet radii weldments to reduce stress concentrations at critical corners. (6) Review quality assurance plan for adequacy to insure structural quality in the small radius region. (7) Review instrumentation plan for adequacy of structural diagnostics in small radius region. (8) Participate in planning a small-scale fatigue test program of a typical MFTF weldment

  6. Physics basis for an axicell design for the end plugs of MFTF-B

    International Nuclear Information System (INIS)

    Baldwin, D.E.; Logan, B.G.

    1982-01-01

    The primary motivation for conversion of MFTF-B to an axicell configuration lies in its engineering promise as a reactor geometry based on circular high-magnetic-field coils. In comparing this configuration to the previous A-cell geometry, we find a number of differences that might significantly affect the physics performance. The purpose of the present document is to examine those features and to assess their impact on the performance of the axicell, as compared to the A-cell configuration, for MFTF-B. In so doing, we address only those issues thought to be affected by the change in geometry and refer to the original report Physics Basis for MFTF-B, for discussion of those issues thought not to be affected. In Sec. 1, we summarize these physics issues. In Sec. 2, we describe operating scenarios in the new configuration. In the Appendices, we discuss those physics issues that require more detailed treatment

  7. Review of MFTF yin-yang magnet displacement and magnetic field measurements and calculations

    International Nuclear Information System (INIS)

    Hanson, C.L.; Myall, J.O.; Wohlwend, J.W.

    1983-01-01

    During the recent testing of the MFTF yin-yang magnet, measurements of coil position, structural case strain, and magnetic field were made to verify calculated values. Measurements to detect magnet movement were taken throughout cooldown and during the operation of the magnet. The magnetic field at the mirror points was measured by Hall-effect probes. The magnet position, structural case strain, and magnetic field measurements indicated a reasonably close correlation with calculated values. Information obtained from the yin-yang test has been very useful in setting realistic mechanical alignment values for the new MFTF-B magnet system

  8. Review of MFTF yin-yang magnet displacement and magnetic field measurements and calculations

    International Nuclear Information System (INIS)

    Hanson, C.L.; Myall, J.O.; Wohlwend, J.W.

    1983-01-01

    During the recent testing of the MFTF yin-yang magnet, measurements of coil position, structural case strain, and magnetic field were made to verify calculated values. Measurements to detect magnet movement were taken throughout cooldown and during the operation of the magnet. The magnetic field at the mirror points was measured by Hall-effect probes. The magnet position, structural case strain, and magntic field measurements indicated a reasonably close correlation with calculated values. Information obtained from the yin-yang test has been very useful in setting realistic mechanical alignment values for the new MFTF-B magnet system

  9. Sparking protection for MFTF-B Neutral Beam Power Supplies

    International Nuclear Information System (INIS)

    Cummings, D.B.

    1983-01-01

    This paper describes the upgrade of MFTF-B Neutral Beam Power Supplies for sparking protection. High performance ion sources spark repeatedly so ion source power supplies must be insensitive to sparking. The hot deck houses the series tetrode, arc and filament supplies, and controls. Hot deck shielding has been upgraded and a continuous shield around the arc, filament, gradient grid, and control cables now extends from the hot deck, through the core snubber, to the source. The shield carries accelerating current and connects only to the source. Shielded source cables go through an outer duct which now connects to a ground plane under the hot deck. This hybrid transmission line is a low inductance path for sparks discharging the stray capacitance of the hot deck and isolation transformers, reducing coupling to building steel. Parallel DC current return cables inside the duct lower inductance to reduce inductive turn-off transients. MOVs to ground further limit surges in the remote power supply return. Single point grounding is at the source. No control or rectifier components have been damaged nor are there any known malfunctions due to sparking up to 80 kV output

  10. MFTF-B quasi-optical ECRH transmission system

    International Nuclear Information System (INIS)

    Yugo, J.J.; Shearer, J.W.; Ziolkowski, R.W.

    1983-01-01

    The microwave transmission system for ERCH on MFTF-B will utilize quasi-optical transmission techniques. The system consists of ten gyrotron oscillators: two gyrotrons at 28 GHz, two at 35 GHz, and six at 56 GHz. The 28 and 35 GHz gyrotrons both heat the electrons in the end plug (potential peak) while the 56 GHz sources heat the minimum-B anchor region (potential minimum). Microwaves are launched into a pair of cylindrical mirrors that form a pseudo-cavity which directs the microwaves through the plasma numerous times before they are lost out of the cavity. The cavity allows the microwave beam to reach the resonance zone over a wide range of plasma densities and temperatures. The fundamental electron cyclotron resonance moves to higher axial positions as a result of beta-depression of the magnetic field, doppler shifting of the resonance, and relativistic mass corrections for the electrons. With this system the microwave beam will reach the resonance surface at the correct angle of incidence for any density or temperature without active aiming of the antennas. The cavity also allows the beam to make multiple passes through the plasma to increase the heating efficiency at low temperatures and densities when the single pass absorption is low. In addition, neutral beams and diagnostics have an unobstructed view of the plasma

  11. Sparking protection for MFTF-B neutral beam power supplies

    International Nuclear Information System (INIS)

    Cummings, D.B.

    1983-01-01

    This paper describes the upgrade of MFTF-B Neutral Beam Power Supplies for sparking protection. High performance ion sources spark repeatedly so ion source power supplies must be insensitive to sparking. The hot deck houses the series tetrode, arc and filament supplies, and controls. Hot deck shielding has been upgraded and a continuous shield around the arc, filament, gradient grid, and control cables now extends from the hot deck, through the core snubber, to the source. The shield carries accelerating current and connects only to the source. Shielded source cables go through an outer duct which now connects to a ground plane under the hot deck. This hybrid transmission line is a low inductance path for sparks discharging the stray capacitance of the hot deck and isolation transformers, reducing coupling to building steel. Parallel dc current return cables inside the duct lower inductance to reduce inductive turn-off transients. MOVs to ground further limit surges in the remote power supply return. Single point grounding is at the source. No control or rectifier components have been damaged nor are there any known malfunctions due to sparking up to 80 kV output

  12. Directions for possible upgrades of the Mirror Fusion Test Facility (MFTF)

    International Nuclear Information System (INIS)

    Damm, C.C.; Coensgen, F.H.; Devoto, R.S.; Molvik, A.W.; Porter, G.D.; Shearer, J.W.; Stallard, B.W.

    1977-01-01

    The Mirror Fusion Test Facility (MFTF) may be upgraded by extending the time of plasma sustenance in an approach to steady-state operation and/or by increasing the neutral-beam injection energy. Some parameter bounds for these upgrades are discussed as they relate to a definition of the required neutral-beam development

  13. Confirmatory analysis and detail design of the magnet system for mirror fusion test facility (MFTF)

    International Nuclear Information System (INIS)

    Tatro, R.E.; Baldi, R.W.

    1978-10-01

    This summary covers the six individual reports delivered to the LLL MFTF program staff. They are: (1) literature survey (helium heat transfer), (2) thermodynamic analysis, (3) structural analysis, (4) manufacturing/producibility study, (5) instrumentation plan and (6) quality assurance report

  14. Report on the engineering test of the LBL 30 second neutral beam source for the MFTF-B project

    International Nuclear Information System (INIS)

    Vella, M.C.; Pincosy, P.A.; Hauck, C.A.; Pyle, R.V.

    1984-08-01

    Positive ion based neutral beam development in the US has centered on the long pulse, Advanced Positive Ion Source (APIS). APIS eventually focused on development of 30 second sources for MFTF-B. The Engineering Test was part of competitive testing of the LBL and ORNL long pulse sources carried out for the MFTF-B Project. The test consisted of 500 beam shots with 80 kV, 30 second deuterium, and was carried out on the Neutral Beam Engineering Test Facility (NBETF). This report summarizes the results of LBL testing, in which the LBL APIS demonstrated that it would meet the requirements for MFTF-B 30 second sources. In part as a result of this test, the LBL design was found to be suitable as the baseline for a Common Long Pulse Source design for MFTF-B, TFTR, and Doublet Upgrade

  15. Evaporation under vacuum condition

    International Nuclear Information System (INIS)

    Mizuta, Satoshi; Shibata, Yuki; Yuki, Kazuhisa; Hashizume, Hidetoshi; Toda, Saburo; Takase, Kazuyuki; Akimoto, Hajime

    2000-01-01

    In nuclear fusion reactor design, an event of water coolant ingress into its vacuum vessel is now being considered as one of the most probable accidents. In this report, the evaporation under vacuum condition is evaluated by using the evaporation model we have developed. The results show that shock-wave by the evaporation occurs whose behavior strongly depends on the initial conditions of vacuum. And in the case of lower initial pressure and temperature, the surface temp finally becomes higher than other conditions. (author)

  16. Results of studies performed on the model of the MFTF Supervisory Control and Diagnostics System (SCDS)

    International Nuclear Information System (INIS)

    Wyman, R.H.

    1979-01-01

    The design and implementation of the SCDS is a relatively complex problem involving a nine-computer network coupled with a unique color graphics control console system, 50 local control minicomputers, and the usual array of drives, printers, magnetic tapes, etc. Four million bytes of data are to be collected on each MFTF cycle with a repetition rate of five minutes per shot, and the associated data processing and storing load is a major concern. Crude paper studies were made initially to try to size the various components of the system and various configurations were proposed and analyzed prior to the solicitation for the computer system. However, once the hardware was purchased and a preliminary software design was completed, it became essential and feasible to do an analysis of the system to considerably greater depth in order to identify bottlenecks and other system problems and to verify those parts of the design that met the MFTF requirements

  17. Quench Detection and Magnet Protection Study for MFTF. LLL final review

    International Nuclear Information System (INIS)

    1979-06-01

    The results of a Quench Detection and Magnet Protection Study for MFTF are summarized. The study was directed toward establishing requirements and guidelines for the electronic package used to protect the MFTF superconducting magnets. Two quench detection schemes were analyzed in detail, both of which require a programmable quench detector. Hardware and software recommendations for the quench detector were presented as well as criteria for dumping the magnet energy in the event of a quench. Overall magnet protection requirements were outlined in a detailed Failure Mode Effects and Criticality analysis, (FMECA). Hardware and software packages compatible with the FMECA were recommended, with the hardware consisting of flexible, dedicated intelligent modules specifically designed for magnet protection

  18. Startup experience with the MFTF-B ECRH 100 kV dc power supply

    International Nuclear Information System (INIS)

    Bishop, S.R.; Goodman, R.A.; Wilson, J.H.

    1983-01-01

    One of the 24 Accel dc Power Supplies (ADCPS) originally intended for the Mirror Fusion Test Facility (MFTF-B) Neutral Beam Power Supply (NBPS) System has been converted to provide negative polarity output at 90 kV with a load current of 64 A dc. The load duty cycle is a pulse of 30-seconds duration with a pulse repetition period of five minutes. A new control system has been built which will serve as a prototype for the MFTF-B ADCPS controls, and a test setup was built which will be used to test the ADCPS. The Electron Cyclotron Resonance Heating (ECRH) dc Power Supply (DCPS) has been tested under both no-load and dummy-load conditions, under remote control, without notable problems. Test results indicate that the power supply should be reliable and safe to operate, and will meet the load duty requirements

  19. Startup experience with the MFTF-B ECRH 100 kV dc power supply

    International Nuclear Information System (INIS)

    Bishop, S.R.; Goodman, R.A.; Wilson, J.H.

    1983-01-01

    One of the 24 Accel DC Power Supplies (ADCPS) originally intended for the Mirror Fusion Test Facility (MFTF-B) Neutral Beam Power Supply (NBPS) System has been converted to provide negative polarity output at 90 kV with a load current of 64 A dc. The load duty cycle is a pulse of 30-seconds duration with a pulse repetition period of five minutes. A new control system has been built which will serve as a prototype for the MFTF-B ADCPS controls, and a test setup was built which will be used to test the ADCPS. The Electron Cyclotron Resonance Heating (ECRH) DC Power Supply (DCPS) has been tested under both no-load and dummy-load conditions, under remote control, without notable problems. Test results indicate that the power supply should be reliable and safe to operate, and will meet the load duty requirements

  20. Ion trajectories of the MFTF unshielded 80-keV neutral-beam sources

    International Nuclear Information System (INIS)

    Ling, R.C.; Bulmer, R.H.; Cutler, T.A.; Foote, J.H.; Horvath, J.A.

    1978-01-01

    The trajectories of ions from the Magnetic Fusion Test Facility (MFTF) 80-keV neutral-beam sources are calculated to obtain a preliminary understanding of the ion-beam paths and the magnitude of the power densities. This information will be needed for locating and designing thermal (kinetic-energy) absorbers for the ions. The calculations are made by employing a number of previously written computer codes. The TIBRO code is used to calculate the trajectories of the ions in the fringe magnetic field of the MFTF machine, which can operate with a center-field intensity of up to 2 T. The SAMPP code gives three-dimensional views of the ion beams for better visualization of the ion-beam paths. Also used are the codes MIG, XPICK, and MERGE, which were all previously written for manipulating data

  1. Integration of cooking and vacuum cooling of carrots in a same vessel Integração dos processos de cozimento e resfriamento a vácuo de cenouras em um mesmo tanque

    Directory of Open Access Journals (Sweden)

    Luiz Gustavo Gonçalves Rodrigues

    2012-03-01

    Full Text Available Cooked vegetables are commonly used in the preparation of ready-to-eat foods. The integration of cooking and cooling of carrots and vacuum cooling in a single vessel is described in this paper. The combination of different methods of cooking and vacuum cooling was investigated. Integrated processes of cooking and vacuum cooling in a same vessel enabled obtaining cooked and cooled carrots at the final temperature of 10 ºC, which is adequate for preparing ready-to-eat foods safely. When cooking and cooling steps were performed with the samples immersed in boiling water, the effective weight loss was approximately 3.6%. When the cooking step was performed with the samples in boiling water or steamed, and the vacuum cooling was applied after draining the boiling water, water loss ranged between 15 and 20%, which caused changes in the product texture. This problem can be solved with rehydration using a small amount of sterile cold water. The instrumental textural properties of carrots samples rehydrated at both vacuum and atmospheric conditions were very similar. Therefore, the integrated process of cooking and vacuum cooling of carrots in a single vessel is a feasible alternative for processing such kind of foods.Para a preparação de refeições rápidas é comum o uso de legumes cozidos. A integração dos processos de cozimento e resfriamento de cenouras em um mesmo tanque pelo uso do resfriamento a vácuo é descrito neste artigo. A combinação de diferentes métodos de cozimento e resfriamento a vácuo foi investigada. O processo integrado de cozimento-resfriamento a vácuo em um mesmo tanque permitiu obter cenouras cozidas-resfriadas com temperaturas finais de 10 ºC, o que é adequado à preparação de refeições rápidas com segurança. Quando o processo de cozimento-resfriamento foi realizado com amostras imersas em água de cozimento, a perda efetiva de massa foi de aproximadamente 3,6%. Quando o processo de cozimento-resfriamento foi

  2. Computer circuit analysis of induced currents in the MFTF-B magnet system

    International Nuclear Information System (INIS)

    Magnuson, G.D.; Woods, E.L.

    1981-01-01

    An analysis was made of the induced current behavior of the MFTF-B magnet system. Although the magnet system consists of 22 coils, because of its symmetry we considered only 11 coils in the analysis. Various combinations of the coils were dumped either singly or in groups, with the current behavior in all magnets calculated as a function of time after initiation of the dump

  3. Design of the drift pumping system for MFTF-α+T

    International Nuclear Information System (INIS)

    Metlzer, D.H.

    1983-01-01

    Drift pumping in mirrors is a new concept (less than one year old). If it works, compared to charge-exchange pumping, it will simplify the MFTF-α+T interface and possibly reduce the circulating power required. From an engineering standpoint, it has some very demanding requirements in terms of power and bandwidth. This paper describes a design which satisfies these requirements. It also identifies a number of promising alternatives requiring investigation and verification

  4. Evaluating and tuning system response in the MFTF-B control and diagnostics computers

    International Nuclear Information System (INIS)

    Palasek, R.L.; Butner, D.N.; Minor, E.G.

    1983-01-01

    The software system running on the Supervisory Control and Diagnostics System (SCDS) of MFTF-B is, for the major part, an event driven one. Regular, periodic polling of sensors' outputs takes place only at the local level, in the sensors' corresponding local control microcomputers (LCC's). An LCC reports a sensor's value to the supervisory computer only if there was a significant change. This report is passed as a message, routed among and acted upon by a network of applications and systems tasks within the supervisory computer (SCDS). Commands from the operator's console are similarly routed through a network of tasks, but in the oppostie direction to the experiment's hardware. In a network such as this, response time is partialy determined by system traffic. Because the hardware of MFTF-B will not be connected to the computer system for another two years, we are using the local control computers to simulate the event driven traffic that we expect to see during MFTF-B operation. In this paper we show how we are using the simulator to measure and evaluate response, loading, throughput, and utilization of components within the computer system. Measurement of the system under simulation allows us to identify bottlenecks and verify their unloosening. We also use the traffic simulators to evaluate prototypes of different algorithms for selected tasks, comparing their responses under the spectrum of traffic intensities

  5. Impacts of the Shine Through neutrals on the Vacuum Vessel of TJ-II during NBI; Impactos de los Neutros de Shine Through en la Camara de Vacio del TJ-II durante NBI

    Energy Technology Data Exchange (ETDEWEB)

    Guasp, J.; Liniers, M.

    1995-07-01

    A numerical analysis of the impact patterns on the Vacuum Vessel produced by Shine Through neutrals during the tangential balanced NBI in TJ-II Helical Axis Stellarator has been done. The results show two main concentrations. The first one the circular part of the Hard Core, near the zone of closest approach of the beam. The second one, the most important, around the beam exit, on the border between the plate of the HC cover and the sector wall. As expected, the Shine Through loads decrease strongly with plasma density, predominating at low density at NBI start, decreasing quickly when density increases and increasing slightly with the beam energy. No overlapping with lost fast ions impacts is observed, that, in addition, show an opposite behaviour with density. (Author) 3 refs.

  6. Impacts of lost fast ions on the TJ-II Vacuum Vessel during NBI; Impactos de los iones rapidos en la Camara de Vacio del TJ-II durante NBI

    Energy Technology Data Exchange (ETDEWEB)

    Guasp, J

    1995-07-01

    The possible deposition patterns, on the Vacuum Vessel, of lost fast ions during the balanced tangential NBI in TJ-II helical axis Stellarator are analysed theoretically, establishing the relation between those impact points, the plasma exit and birth positions and the magnetic configuration characteristics. It is shown that direct losses are the most important, mainly those produced by the beam injected with the same direction that the magnetic field, increasing with beam energy and plasma density but with impacts remaining fixed on well defined zones, a periodically distributed along the Hard Core cover plates, producing high loads at high densities. The remaining losses, except for the shine through ones that predominate at low density, are periodically distributed, with smooth maxima and produce very low loads. No overlapping between the different kind of losses or beams is observed. (Author) 6 refs.

  7. D-T axicell magnet system for MFTF-α+T

    International Nuclear Information System (INIS)

    Srivastava, V.C.

    1983-01-01

    The configuration and design of the deuterium-tritium (D-T) axicell superconducting magnets for the Mirror Fusion Test Facility (MFTF-α+T) are described. The MFTF-α+T is an upgrade of the MFTF-B, with new end-plug magnets and a neutron-producing central D-T axicell section. The 4-m long axicell - its length defined by the 12-T peaks in the mirror field - is beam fueled and heated by two beam lines, each with four neutral beam injection ports. Two large superconducting coils (means diameter approx. 3.8 m) located at Z = +-2.40 m, in conjunction with a small copper coil located outside the test volume region, produce the 4.5-T mirror midplane field. This background field is augmented by two copper coils to create the 12-T peak mirror fields at Z = +-2 m. The central region of the axicell accommodates a 1-m-long, replaceable blanket test module. The length (4 m) of the axicell was chosen to provide relatively uniform neutron wall loading over the test module. In many respects, this axicell is less than full scale, but it could be viewed as a short section of a reactor, complete with the support systems and technologies associated with a mirror reactor. The peak field at the superconducting coils is 10.8 T. The coils employ hybrid superconducting winding - Nb 3 Sn conductor in the 8- to 12-T region and NbTi in the 0- to 8-T region. The winding is cryostable and is cooled by a 4.2 K liquid helium bath. The conductor design, the winding design, and the performance analyses for these superconducting coils are described

  8. The local area network for the plasma Diagnostics System of MFTF-B

    International Nuclear Information System (INIS)

    Lau, N.H.; Minor, E.G.

    1983-01-01

    The MFTF-B Plasma Diagnostics System will be implemented in stages, beginning with a start-up set of diagnostics and evolving toward a basic set. The start-up set contains 12 diagnostics which will acquire a total of about 800 Kbytes of data per machine pulse; the basic set contains 23 diagnostics which will acquire a total of about 8 Mbytes of data per pulse. Each diagnostic is controlled by a ''Foundation System'' consisting of a DEC LSI-11/23 microcomputer connected to CAMAC via a 5 Mbits/second serial fiber-optic link and connected to a supervisory computer (Perkin-Elmer 3250) via a 9600 baud RS232 link. The Foundation System is a building block used throughout MFTF-B for control and status monitoring. However, its 9600 baud link to the supervisor presents a bottleneck for the large data transfers required by diagnostics. To overcome this bottleneck the diagnostics Foundation Systems will be connected together with an additional LSI-11/23 called the ''master'' to form a Local Area Network (LAN) for data acquisition. The Diagnostics LAN has a ring architecture with token passing arbitration

  9. Overview of the data acquisition and control system for plasma diagnostics on MFTF-B

    International Nuclear Information System (INIS)

    Wyman, R.H.; Deadrick, F.J.; Lau, N.H.; Nelson, B.C.; Preckshot, G.G.; Throop, A.L.

    1983-01-01

    For MFTF-B, the plasma diagnostics system is expected to grow from a collection of 12 types of diagnostic instruments, initially producing about 1 Megabyte of data per shot, to an expanded set of 22 diagnostics producing about 8 Megabytes of data per shot. To control these diagnostics and acquire and process the data, a system design has been developed which uses an architecture similar to the supervisory/local-control computer system which is used to control other MFTF-B subsystems. This paper presents an overview of the hardware and software that will control and acquire data from the plasma diagnostics system. Data flow paths from the instruments, through processing, and into final archived storage will be described. A discussion of anticipated data rates, including anticipated software overhead at various points of the system, is included, along with the identification of possible bottlenecks. A methodology for processing of the data is described, along with the approach to handle the planned growth in the diagnostic system. Motivations are presented for various design choices which have been made

  10. A user interface on networked workstations for MFTF-B plasma diagnostic instruments

    International Nuclear Information System (INIS)

    Balch, T.R.; Renbarger, V.L.

    1986-01-01

    A network of Sun-2/170 workstations is used to provide an interface to the MFTF-B Plasma Diagnostics System at Lawrence Livermore National Laboratory. The Plasma Diagnostics System (PDS) is responsible for control of MFTF-B plasma diagnostic instrumentation. An EtherNet Local Area Network links the workstations to a central multiprocessing system which furnishes data processing, data storage and control services for PDS. These workstations permit a physicist to command data acquisition, data processing, instrument control, and display of results. The interface is implemented as a metaphorical desktop, which helps the operator form a mental model of how the system works. As on a real desktop, functions are provided by sheets of paper (windows on a CRT screen) called worksheets. The worksheets may be invoked by pop-up menus and may be manipulated with a mouse. These worksheets are actually tasks that communicate with other tasks running in the central computer system. By making entries in the appropriate worksheet, a physicist may specify data acquisition or processing, control a diagnostic, or view a result

  11. Use of spreadsheets for interactive control of MFTF-B plasma diagnostic instruments

    International Nuclear Information System (INIS)

    Preckshot, G.G.; Goldner, A.L.; Kobayashi, A.

    1986-01-01

    The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory has a variety of highly individualized plasma diagnostic instruments attached to the experiment. These instruments are controlled through graphics workstations networked to a central computer system. A distributed spreadsheet-like program runs in both the graphics workstations and in the central computer system. An interface very similar to a commercial spreadsheet program is presented to the user at a workstation. In a commercial spreadsheet program, the user may attach mathematical calculation functions to spreadsheet cells. At MFTF-B, hardware control functions, hardware monitoring functions, and communications functions, as well as mathematical functions, may be attached to cells. Both the user and feedback from instrument hardware may make entries in spreadsheet cells; any entry in a spreadsheet cell may cause reevaluation of the cell's associated functions. The spreadsheet approach makes the addition of a new instrument a matter of designing one or more spreadsheet tables with associated meta-language-defined control and communication function strings. This paper describes the details of the spreadsheets and the implementation experience

  12. Local area network for the plasma diagnostics system of MFTF-B

    International Nuclear Information System (INIS)

    Lau, N.H.; Minor, E.G.

    1983-01-01

    The MFTF-B Plasma Diagnostics System will be implemented in stages, beginning with a start-up set of diagnostics and evolving toward a basic set. The start-up set contains 12 diagnostics which will acquire a total of about 800 Kbytes of data per machine pulse; the basic set contains 23 diagnostics which will acquire a total of about 8 Mbytes of data per pulse. Each diagnostic is controlled by a Foundation System consisting of a DEC LSI-11/23 microcomputer connected to CAMAC via a 5 Mbits/second serial fiber-optic link and connected to a supervisory computer (Perkin-Elmer 3250) via a 9600 baud RS232 link. The Foundation System is a building block used throughout MFTF-B for control and status monitoring. However, its 9600 baud link to the supervisor presents a bottleneck for the large data transfers required by diagnostics. To overcome this bottleneck the diagnostics Foundation Systems will be connected together with an additional LSI-11/23 called the master to form a Local Area Network (LAN) for data acquisition

  13. Design of a magnetic field alignment diagnostic for the MFTF-B magnet system

    International Nuclear Information System (INIS)

    Deadrick, F.J.; House, P.A.; Frye, R.W.

    1985-01-01

    Magnet alignment in tandem mirror fusion machines plays a crucial role in achieving and maintaining plasma confinement. Various visual alignment tools have been described by Post et al. to align the Tara magnet system. We have designed and installed a remotely operated magnetic field alignment (MFA) diagnostic system as a part of the Mirror Fusion Test Facility (MFTF-B). It measures critical magnetic field alignment parameters of the MFTF-B coil set while under full-field operating conditions. The MFA diagnostic employs a pair of low-energy, electron beam guns on a remotely positionable probe to trace and map selected magnetic field lines. An array of precision electrical detector paddles locates the position of the electron beam, and thus the magnetic field line, at several critical points. The measurements provide a means to compute proper compensating currents to correct for mechanical misalignments of the magnets with auxiliary trim coils if necessary. This paper describes both the mechanical and electrical design of the MFA diagnostic hardware

  14. Use of spreadsheets for interactive control of MFTF-B plasma diagnostic instruments

    International Nuclear Information System (INIS)

    Preckshot, G.G.; Goldner, A.; Kobayashi, A.

    1985-01-01

    The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory has a variety of highly individualized plasma diagnostic instruments attached to the experiment. These instruments are controlled through graphics workstations networked to a central computer system. A distributed spreadsheet-like program runs in both the graphics workstations and in the central computer system. An interface very similar to a commercial spreadsheet program is presented to the user at a workstation. In a commercial spreadsheet program, the user may attach mathematical calculation functions to spreadsheet cells. At MFTF-B, hardware control functions, hardware monitoring functions, and communications functions, as well as mathematical functions, may be attached to cells. Both the user and feedback from instrument hardware may make entries in spreadsheet cells; any entry in a spreadsheet cell may cause reevaluation of the cell's associated functions. The spreadsheet approach makes the addition of a new instrument a matter of designing one or more spreadsheet tables with associated meta-language-defined control and communication function strings. We report here details of our spreadsheets and our implementation experience

  15. Mirror Fusion Test Facility-B (MFTF-B) axicell configuration: NbTi magnet system. Manufacturing/producibility final report. Volume 2

    International Nuclear Information System (INIS)

    Ritschel, A.J.; White, W.L.

    1985-05-01

    This Final MFTF-B Manufacturing/Producibility Report covers facilities, tooling plan, manufacturing sequence, schedule and performance, producibility, and lessons learned for the solenoid, axicell, and transition coils, as well as a deactivation plan, conclusions, references, and appendices

  16. Vacuum extraction

    DEFF Research Database (Denmark)

    Maagaard, Mathilde; Oestergaard, Jeanett; Johansen, Marianne

    2012-01-01

    Objectives. To develop and validate an Objective Structured Assessment of Technical Skills (OSATS) scale for vacuum extraction. Design. Two-part study design: Primarily, development of a procedure-specific checklist for vacuum extraction. Hereafter, validation of the developed OSATS scale for vac...

  17. Vacuum mechatronics

    Science.gov (United States)

    Hackwood, Susan; Belinski, Steven E.; Beni, Gerardo

    1989-01-01

    The discipline of vacuum mechatronics is defined as the design and development of vacuum-compatible computer-controlled mechanisms for manipulating, sensing and testing in a vacuum environment. The importance of vacuum mechatronics is growing with an increased application of vacuum in space studies and in manufacturing for material processing, medicine, microelectronics, emission studies, lyophylisation, freeze drying and packaging. The quickly developing field of vacuum mechatronics will also be the driving force for the realization of an advanced era of totally enclosed clean manufacturing cells. High technology manufacturing has increasingly demanding requirements for precision manipulation, in situ process monitoring and contamination-free environments. To remove the contamination problems associated with human workers, the tendency in many manufacturing processes is to move towards total automation. This will become a requirement in the near future for e.g., microelectronics manufacturing. Automation in ultra-clean manufacturing environments is evolving into the concept of self-contained and fully enclosed manufacturing. A Self Contained Automated Robotic Factory (SCARF) is being developed as a flexible research facility for totally enclosed manufacturing. The construction and successful operation of a SCARF will provide a novel, flexible, self-contained, clean, vacuum manufacturing environment. SCARF also requires very high reliability and intelligent control. The trends in vacuum mechatronics and some of the key research issues are reviewed.

  18. Turbomolecular pump vacuum system for the Princeton Large Torus

    International Nuclear Information System (INIS)

    Dylla, H.F.

    1977-10-01

    A turbomolecular pump vacuum system has been designed and installed on the Princeton Large Torus (PLT). Four vertical shaft, oil-bearing, 1500 l/s turbomolecular pumps have been interfaced to the 6400 liter PLT Vacuum vessel to provide a net pumping speed of 3000 l/s for H 2 . The particular requirements and problems of tokamak vacuum systems are enumerated. A vacuum control system is described which protects the vacuum vessel from contamination, and protects the turbomolecular pumps from damage under a variety of possible failure modes. The performance of the vacuum system is presented in terms of pumping speed measurements and residual gas behavior

  19. Vessel Operating Units (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data for vessels that are greater than five net tons and have a current US Coast Guard documentation number. Beginning in1979, the NMFS...

  20. Cosmic vacuum

    International Nuclear Information System (INIS)

    Chernin, Artur D

    2001-01-01

    Recent observational studies of distant supernovae have suggested the existence of cosmic vacuum whose energy density exceeds the total density of all the other energy components in the Universe. The vacuum produces the field of antigravity that causes the cosmological expansion to accelerate. It is this accelerated expansion that has been discovered in the observations. The discovery of cosmic vacuum radically changes our current understanding of the present state of the Universe. It also poses new challenges to both cosmology and fundamental physics. Why is the density of vacuum what it is? Why do the densities of the cosmic energy components differ in exact value but agree in order of magnitude? On the other hand, the discovery made at large cosmological distances of hundreds and thousands Mpc provides new insights into the dynamics of the nearby Universe, the motions of galaxies in the local volume of 10 - 20 Mpc where the cosmological expansion was originally discovered. (reviews of topical problems)

  1. Cosmic vacuum

    Energy Technology Data Exchange (ETDEWEB)

    Chernin, Artur D [P.K. Shternberg State Astronomical Institute at the M.V. Lomonosov Moscow State University, Moscow (Russian Federation)

    2001-11-30

    Recent observational studies of distant supernovae have suggested the existence of cosmic vacuum whose energy density exceeds the total density of all the other energy components in the Universe. The vacuum produces the field of antigravity that causes the cosmological expansion to accelerate. It is this accelerated expansion that has been discovered in the observations. The discovery of cosmic vacuum radically changes our current understanding of the present state of the Universe. It also poses new challenges to both cosmology and fundamental physics. Why is the density of vacuum what it is? Why do the densities of the cosmic energy components differ in exact value but agree in order of magnitude? On the other hand, the discovery made at large cosmological distances of hundreds and thousands Mpc provides new insights into the dynamics of the nearby Universe, the motions of galaxies in the local volume of 10 - 20 Mpc where the cosmological expansion was originally discovered. (reviews of topical problems)

  2. Integrated operations plan for the MFTF-B Mirror Fusion Test Facility. Volume I. Organization plan

    International Nuclear Information System (INIS)

    1981-12-01

    This plan and the accompanying MFTF-B Integrated Operations Plan are submitted in response to UC/LLNL Purchase Order 3883801, dated July 1981. The organization plan also addresses the specific tasks and trade studies directed by the scope of work. The Integrated Operations Plan, which includes a reliability, quality assurance, and safety plan and an integrated logistics plan, comprises the burden of the report. In the first section of this volume, certain underlying assumptions and observations are discussed setting the requirements and limits for organization. Section B presents the recommended structure itself. Section C Device Availability vs Maintenance and Support Efforts and Section D Staffing Levels and Skills provide backup detail and justification. Section E is a trade study on maintenance and support by LLNL staff vs subcontract and Section F is a plan for transitioning from the construction phase into operation. A brief summary of schedules and estimated costs concludes the volume

  3. Currents and voltages in the MFTF coils during the formation of a normal zone

    International Nuclear Information System (INIS)

    Owen, E.W.

    1980-08-01

    Expressions are obtained for the currents and voltages in a pair of inductively coupled superconducting coils under two conditions: formation of a normal zone and during a change in the level of the current in one coil. A dump resistor of low resistance and a detector bridge is connected across each coil. Calculated results are given for the MFTF coils. The circuit equations during formation of a normal zone are nonlinear and time-varying, consequently, only a series solution is possible. The conditions during a change in current are more easily found. After the transient has died away, the voltages in the coil associated with the changing source are all self-inductive, while the voltages in the other coil are all mutually inductive

  4. Noise filtering algorithm for the MFTF-B computer based control system

    International Nuclear Information System (INIS)

    Minor, E.G.

    1983-01-01

    An algorithm to reduce the message traffic in the MFTF-B computer based control system is described. The algorithm filters analog inputs to the control system. Its purpose is to distinguish between changes in the inputs due to noise and changes due to significant variations in the quantity being monitored. Noise is rejected while significant changes are reported to the control system data base, thus keeping the data base updated with a minimum number of messages. The algorithm is memory efficient, requiring only four bytes of storage per analog channel, and computationally simple, requiring only subtraction and comparison. Quantitative analysis of the algorithm is presented for the case of additive Gaussian noise. It is shown that the algorithm is stable and tends toward the mean value of the monitored variable over a wide variety of additive noise distributions

  5. Design lessons from using programmable controllers in the MFTF-B personnel safety and interlocks system

    International Nuclear Information System (INIS)

    Branum, J.D.

    1983-01-01

    Applying programmable controllers in critical applications such as personnel safety and interlocks systems requires special considerations in the design of both hardware and software. All modern programmable controller systems feature extensive internal diagnostic capabilities to protect against problems such as program memory errors; however most, if not all present designs lack an intrinsic capability for detecting and countering failures on the field-side of their I/O modules. Many of the most common styles of I/O modules can also introduce potentially dangerous sneak circuits, even without component failure. This paper presents the most significant lessons learned to date in the design of the MFTF-B Personnel Safety and Interlocks System, which utilizes two non-redundant programmable controllers with over 800 I/O points each. Specific problems recognized during the design process as well as those discovered during initial testing and operation are discussed along with their specific solutions in hardware and software

  6. Design and prototype results of a far-infrared interferometer for MFTF-B

    International Nuclear Information System (INIS)

    Monjes, J.A.; Throop, A.L.; Thomas, S.R.; Peebles, A.; Zu, Qin-Zin.

    1983-01-01

    A Far-Infrared (FIR) Laser Interferometer (FLI), operating at 185 μm wavelength is planned as part of the initial start-up set of plasma diagnostics for the Mirror Fusion Test Facility (MFTF-B). The FLI will consist of a heterodyne, three-chord laser interferometer which will be used initially to measure line-integrated plasma density in the high-density, center cell region of the machine. The conceptual system design and analysis has been completed. There are several unique environmental/physical constraints and performance requirements for this system which have required that technology-evaluation and prototyping experiments be completed to support the design effort and confirm the expected performance parameters. Issues which have been addressed include extensive use of long-path dielectric waveguide, coupling and control of free-space propagation of the beam, and polarization control. The results and conclusions of the design analysis and experimental measurements will be presented

  7. 12-T solenoid-design options for the MFTF-B Upgrade

    International Nuclear Information System (INIS)

    Schultz, J.H.; Diatchenko, N.

    1983-01-01

    The major options for the 12 T magnets examined here are the selection of normal, superconducting or hybrid normal/superconducting magnet systems. The tradeoffs are those between the higher initial cost of superconducting magnet system, the need for thick shielding of superconducting magnets, higher recirculating power in the normal magnets and poorly characterized reliability of lightly shielded normal magnets. The size and shielding tradeoffs among these options are illustrated. The design concepts presented here are evaluated only for the first design iteration of MFTF-B + T, mentioned above. In particular, all concepts now being considered have made topological improvements in the center cell, so that neutral beam power is no longer a strong function of choke coil size. This function was strongly favorable to the use of normal magnets over superconducting magnets and its absence will be discussed qualitatively in the cost comparisons

  8. Plasma potential formation and measurement in TMX-U and MFTF-B

    International Nuclear Information System (INIS)

    Grubb, D.P.

    1984-01-01

    Tandem mirrors control the axial variation of the plasma potential to create electrostatic plugs that improve the axial confinement of central cell ions and, in a thermal barrier tandem mirror, control the electron axial heat flow. Measurements of the spatial and temporal variations of the plasma potential are, therefore, important to the understanding of confinement in a tandem mirror. In this paper we discuss potential formation in a thermal barrier tandem mirror and examine the diagnostics and data obtained on the TMX-U device, including measurements of the thermal barrier potential profile using a diagnostic neutral beam and charged particle energy-spectroscopy. We then describe the heavy ion beam probe and other new plasma potential diagnostics that are under development for TMX-U and MFTF-B and examine problem areas where additional diagnostic development is desirable

  9. Structural analysis of the magnet system for Mirror Fusion Test Facility (MFTF). Addendum I

    International Nuclear Information System (INIS)

    Loss, K.R.; Wohlwend, J.W.

    1979-09-01

    The stress analysis refinement of the MFTF magnet system using GDSAP (General Dynamics Structural Analysis Program) and NASTRAN finite element computer models has been completed. The objective of this analysis was to calculate a more refined case and jacket stress distribution. The GDSAP model was refined in the minor radius area to yield a more detailed prediction of the stress distributions in critical areas identified by the previous analysis. Modifications in the case plate thickness (from 3.0 inches to 3.2 inches) and in the conductor pack load distribution and stiffness were included. The GDSAP model was converted to an identical NASTRAN model to determine the influence on stress results using higher order elements

  10. Vacuum vessel for a thermonuclear device

    International Nuclear Information System (INIS)

    Yamamoto, Masahiro; Shimizu, Masatsugu; Takatsu, Hideyuki; Kuriyama, Masaaki.

    1980-01-01

    Purpose: To avoid stress resulted from the deformation differences in the poloidal direction near the joinings between bellows and thick rings. Constitution: Thick rings and bellows are disposed alternately in the toroidal direction. Electric conductors in electric contaction with the thick rings are provided in the joinings between the thick rings and the bellows while projected from the thick rings to the bellows, thereby causing most of the poloidal components in the saddle pattern current in the bellows to flow through the conductors. The value of the saddle pattern current can be made to such a value as equalizing the poloidal deformation in the bellows caused by the electromagnetic force of the saddle-pattern current to that in the thick rings by designing the electric conductors to a suitable thickness. (Furukawa, Y.)

  11. Design and fabrication of the superconducting-magnet system for the Mirror Fusion Test Facility (MFTF-B)

    International Nuclear Information System (INIS)

    Tatro, R.E.; Wohlwend, J.W.; Kozman, T.A.

    1982-01-01

    The superconducting magnet system for the Mirror Fusion Test Facility (MFTF-B) consists of 24 magnets; i.e. two pairs of C-shaped Yin-Yang coils, four C-shaped transition coils, four solenoidal axicell coils, and a 12-solenoid central cell. General Dynamics Convair Division has designed all the coils and is responsible for fabricating 20 coils. The two Yin-Yang pairs (four coils) are being fabricated by the Lawrence Livermore National Laboratory. Since MFTF-B is not a magnet development program, but rather a major physics experiment critical to the mirror fusion program, the basic philosophy has been to use proven materials and analytical techniques wherever possible. The transition and axicell coils are currently being analyzed and designed, while fabrication is under way on the solenoid magnets

  12. Ion cyclotron resonance heating (ICRH) start-up antenna for the mirror fusion test facility (MFTF-B)

    International Nuclear Information System (INIS)

    McCarville, T.M.; Romesser, T.E.

    1985-01-01

    The purpose of the ICRH start-up antenna on MFTF-B is to heat the plasma and control the ion distribution as the density increases during start-up. The antenna, consisting of two center fed half turn loops phased 180 0 apart, has been designed for 1 MW of input power, with a goal of coupling 400 kW into the ions. To vary the heating frequency relative to the local ion cyclotron frequency, the antenna is tunable over a range from 7.5 to 12.5 MHz. The thermal requirements common to low duty cycle ICRH antennas are especially severe for the MFTF-B antenna. The stress requirements are also unique, deriving from the possibility of seismic activity or JxB forces if the magnets unexpectedly quench. Considerable attention has been paid to contact control at high current bolt-up joints, and arranging geometries so as to minimize the possibility of voltage breakdown

  13. Design of the ZTH vacuum liner

    International Nuclear Information System (INIS)

    Prince, P.P.; Dike, R.S.

    1987-01-01

    The current status of the ZTh vacuum liner design is covered by this report. ZTH will be the first experiment to be installed in the CPRF (Confinement Physics Research Facility) at the Los Alamos National Laboratory and is scheduled to be operational at the rated current of 4 MA in 1992. The vacuum vessel has a 2.4m major radius and a 40 cm minor radius. Operating parameters which drive the vacuum vessel mechanical design include a 300 C bakeout temperature, an armour support system capable of withstanding 25 kV, a high toroidal resistance, 1250 kPa magnetic loading, a 10 minute cycle time, and high positional accuracy with respect to the conducting shell. The vacuum vessel design features which satisfy the operating parameters are defined

  14. Test of an undulated vacuum chamber for the ISR

    CERN Multimedia

    1975-01-01

    This picture shows mechanical tests of an undulated vacuum chamber for downstream arms of ISR intersections. This chamber, made of 0.3 mm thick inconel, had inner dimensions of 150 mm by 50 mm. The deflection under vacuum is measured by dial gauges. On the left one sees the large vessel where vacuum chambers were tested at pressures above atmospheric pressure.

  15. Vacuum gauges

    International Nuclear Information System (INIS)

    Power, B.D.; Priestland, C.R.D.

    1978-01-01

    This invention relates to vacuum gauges, particularly of the type known as Penning gauges, which are cold cathode ionisation gauges, in which a magnetic field is used to lengthen the electron path and thereby increase the number of ions produced. (author)

  16. Expanded Fermilab pressure vessel directory program

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect.

  17. Expanded Fermilab pressure vessel directory program

    International Nuclear Information System (INIS)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect

  18. Selection of vacuum seals for EBT-P

    International Nuclear Information System (INIS)

    Dillow, C.F.; Adlon, G.L.; Stubblefield, V.E.

    1981-01-01

    Elmo Bumpy Torus Proof of Principle (EBT-P) is a magnetic fusion RandD facility being constructed by McDonnell Douglas Astronautics Company - St. Louis Division of Oak Ridge National Laboratory (ORNL). EBT-P is a truly unique fusion device requiring careful consideration in selecting both the primary vacuum seals on the toroidal vessel and the secondary vacuum seals in components such as vacuum pumps and and valves. The vacuum seal environment is described and the considerations in vacuum seal selection fully discussed. Methods for protecting vacuum seals in pumps and valves from the microwave environments are also presented

  19. Deflated-Victims of vacuum

    International Nuclear Information System (INIS)

    Sanders, Roy E.

    2007-01-01

    Atmospheric pressure combined with a partial vacuum within chemical plant or refinery tanks can result in some ego-deflating moments. This article will review three catastrophic vessel failures in detail and touch on several other incidents. A 4000-gal acid tank was destroyed by a siphoning action; a well maintained tank truck was destroyed during a routine delivery; and a large, brand new refinery mega-vessel collapsed as the steam within it condensed. Seasoned engineers are aware of the frail nature of tanks and provide safeguards or procedures to limit damages. The purpose of this paper is to ensure this new generation of chemical plant/refinery employees understand the problems of the past and take the necessary precautions to guard against tank damages created by partial vacuum conditions

  20. Bake-Out Mobile Controls for Large Vacuum Systems

    CERN Document Server

    Blanchard, S; Gomes, P; Pereira, H; Kopylov, L; Merker, S; Mikheev, M

    2014-01-01

    Large vacuum systems at CERN (Large Hadron Collider - LHC, Low Energy Ion Rings - LEIR...) require bake-out to achieve ultra-high vacuum specifications. The bake-out cycle is used to decrease the outgassing rate of the vacuum vessel and to activate the Non-Evaporable Getter (NEG) thin film. Bake-out control is a Proportional-Integral-Derivative (PID) regulation with complex recipes, interlocks and troubleshooting management and remote control. It is based on mobile Programmable Logic Controller (PLC) cabinets, fieldbus network and Supervisory Control and Data Acquisition (SCADA) application. The CERN vacuum installations include more than 7 km of baked vessels; using mobile cabinets reduces considerably the cost of the control system. The cabinets are installed close to the vacuum vessels during the time of the bake-out cycle. Mobile cabinets can be used in any of the CERN vacuum facilities. Remote control is provided through a fieldbus network and a SCADA application

  1. A computer model of the MFTF-B neutral beam accel dc power supply

    International Nuclear Information System (INIS)

    Wilson, J.H.

    1983-01-01

    Using the SCEPTRE circuit modeling code, a computer model was developed for the MFTF Neutral Beam Power Supply System (NBPSS) Accel DC Power Supply (ADCPS). The ADCPS provides 90 kV, 88 A, to the Accel Modulator. Because of the complex behavior of the power supply, use of the computer model is necessary to adequately understand the power supply's behavior over a wide range of load conditions and faults. The model developed includes all the circuit components and parameters, and some of the stray values. The model has been well validated for transients with times on the order of milliseconds, and with one exception, for steady-state operation. When using a circuit modeling code for a system with a wide range of time constants, it can become impossible to obtain good solutions for all time ranges at once. The present model concentrates on the millisecond-range transients because the compensating capacitor bank tends to isolate the power supply from the load for faster transients. Attempts to include stray circuit elements with time constants in the microsecond and shorter range have had little success because of hugh increases in computing time that result. The model has been successfully extended to include the accel modulator

  2. Safety procedures for the MFTF sustaining-neutral-beam power supply

    International Nuclear Information System (INIS)

    Wilson, J.H.

    1981-01-01

    The MFTF SNBPSS comprises a number of sources of potentially hazardous electrical energy in a small physical area. Power is handled at 80 kV dc, 80 A; 70 V dc, 4000 A; 25 V dc, 5500 A; 3 kV dc, 10 A; and 2 kV dc, 10 A. Power for these systems is furnished from two separate 480 V distribution systems and a 13.8 kV distribution system. A defense in depth approach is used; interlocks are provided in the hardware to make it difficult to gain access to an energized circuit, and the operating procedure includes precautions which would protect personnel even if no interlocks were working. The complexity of the system implies a complex operating procedure, and this potential complexity is controlled by presenting the procedure in a modular form using 37 separate checklists for specific operations. The checklists are presented in flowchart form, so contingencies can be handled at the lowest possible level without compromising safety

  3. Computer model of the MFTF-B neutral beam Accel dc power supply

    International Nuclear Information System (INIS)

    Wilson, J.H.

    1983-01-01

    Using the SCEPTRE circuit modeling code, a computer model was developed for the MFTF Neutral Beam Power Supply System (NBPSS) Accel dc Power Supply (ADCPS). The ADCPS provides 90 kV, 88 A, to the Accel Modulator. Because of the complex behavior of the power supply, use of the computer model is necessary to adequately understand the power supply's behavior over a wide range of load conditions and faults. The model developed includes all the circuit components and parameters, and some of the stray values. The model has been well validated for transients with times on the order of milliseconds, and with one exception, for steady-state operation. When using a circuit modeling code for a system with a wide range of time constants, it can become impossible to obtain good solutions for all time ranges at once. The present model concentrates on the millisecond-range transients because the compensating capacitor bank tends to isolate the power supply from the load for faster transients. Attempts to include stray circuit elements with time constants in the microsecond and shorter range have had little success because of huge increases in computing time that result. The model has been successfully extended to include the accel modulator

  4. 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

  5. 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

  6. Research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, P.S.

    The role of the research vessels as a tool for marine research and exploration is very important. Technical requirements of a suitable vessel and the laboratories needed on board are discussed. The history and the research work carried out...

  7. Integrated operations plan for the MFTF-B Mirror Fusion Test Facility. Volume II. Integrated operations plan

    Energy Technology Data Exchange (ETDEWEB)

    1981-12-01

    This document defines an integrated plan for the operation of the Lawrence Livermore National Laboratory (LLNL) Mirror Fusion Test Facility (MFTF-B). The plan fulfills and further delineates LLNL policies and provides for accomplishing the functions required by the program. This plan specifies the management, operations, maintenance, and engineering support responsibilities. It covers phasing into sustained operations as well as the sustained operations themselves. Administrative and Plant Engineering support, which are now being performed satisfactorily, are not part of this plan unless there are unique needs.

  8. Integrated operations plan for the MFTF-B Mirror Fusion Test Facility. Volume II. Integrated operations plan

    International Nuclear Information System (INIS)

    1981-12-01

    This document defines an integrated plan for the operation of the Lawrence Livermore National Laboratory (LLNL) Mirror Fusion Test Facility (MFTF-B). The plan fulfills and further delineates LLNL policies and provides for accomplishing the functions required by the program. This plan specifies the management, operations, maintenance, and engineering support responsibilities. It covers phasing into sustained operations as well as the sustained operations themselves. Administrative and Plant Engineering support, which are now being performed satisfactorily, are not part of this plan unless there are unique needs

  9. Advanced light source vacuum policy and vacuum guidelines for beamlines and experiment endstations

    International Nuclear Information System (INIS)

    Hussain, Z.

    1995-08-01

    The purpose of this document is to: (1) Explain the ALS vacuum policy and specifications for beamlines and experiment endstations. (2) Provide guidelines related to ALS vacuum policy to assist in designing beamlines which are in accordance with ALS vacuum policy. This document supersedes LSBL-116. The Advanced Light Source is a third generation synchrotron radiation source whose beam lifetime depends on the quality of the vacuum in the storage ring and the connecting beamlines. The storage ring and most of the beamlines share a common vacuum and are operated under ultra-high-vacuum (UHV) conditions. All endstations and beamline equipment must be operated so as to avoid contamination of beamline components, and must include proper safeguards to protect the storage ring vacuum from an accidental break in the beamline or endstation vacuum systems. The primary gas load during operation is due to thermal desorption and electron/photon induced desorption of contaminants from the interior of the vacuum vessel and its components. The desorption rates are considerably higher for hydrocarbon contamination, thus considerable emphasis is placed on eliminating these sources of contaminants. All vacuum components in a beamline and endstation must meet the ALS vacuum specifications. The vacuum design of both beamlines and endstations must be approved by the ALS Beamline Review Committee (BRC) before vacuum connections to the storage ring are made. The vacuum design is first checked during the Beamline Design Review (BDR) held before construction of the beamline equipment begins. Any deviation from the ALS vacuum specifications must be approved by the BRC prior to installation of the equipment on the ALS floor. Any modification that is incorporated into a vacuum assembly without the written approval of the BRC is done at the user's risk and may lead to rejection of the whole assembly

  10. 20 CFR 655.520 - Special provisions regarding automated vessels.

    Science.gov (United States)

    2010-04-01

    ... longshore work consisting of the use of automated self-unloading conveyor belt or vacuum-actuated systems on... use of automated self-unloading conveyor belt or vacuum-actuated equipment at the particular port...-unloading conveyor belt or vacuum-actuated systems on a vessel is the prevailing practice at a particular...

  11. Vacuum system of SST-1 Tokamak

    International Nuclear Information System (INIS)

    Khan, Ziauddin; Pathan, Firozkhan; George, Siju; Semwal, Pratibha; Dhanani, Kalpesh; Paravastu, Yuvakiran; Thankey, Prashant; Ramesh, Gattu; Himabindu, Manthena; Pradhan, Subrata

    2013-01-01

    Highlights: ► Air leaks developed during ongoing SST-1 cooldown campaign were detected online using RGA. ► The presence of N 2 and O 2 gases with the ratio of their partial pressures with ∼3.81:1 confirmed the air leaks. ► Baking of SST-1 was done efficiently by flowing hot N 2 gas in C-channels welded on inner surfaces without any problem. ► In-house fabricated demountable bull nose couplers were demonstrated for high temperature and pressure applications. ► Cryopumping effect was observed when liquid helium cooled superconducting magnets reached below 63 K. -- Abstract: Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN 2 cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10 −4 mbar and 1.0 × 10 −5 mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10 −6 mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10 −5 mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ∼3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10 −5 mbar is achieved inside the cryostat. Baking of the vacuum vessel up to 110 °C with ±10

  12. Vacuum system of SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Ziauddin, E-mail: ziauddin@ipr.res.in [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382 428 (India); Pathan, Firozkhan; George, Siju; Semwal, Pratibha; Dhanani, Kalpesh; Paravastu, Yuvakiran; Thankey, Prashant; Ramesh, Gattu; Himabindu, Manthena; Pradhan, Subrata [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382 428 (India)

    2013-10-15

    Highlights: ► Air leaks developed during ongoing SST-1 cooldown campaign were detected online using RGA. ► The presence of N{sub 2} and O{sub 2} gases with the ratio of their partial pressures with ∼3.81:1 confirmed the air leaks. ► Baking of SST-1 was done efficiently by flowing hot N{sub 2} gas in C-channels welded on inner surfaces without any problem. ► In-house fabricated demountable bull nose couplers were demonstrated for high temperature and pressure applications. ► Cryopumping effect was observed when liquid helium cooled superconducting magnets reached below 63 K. -- Abstract: Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN{sub 2} cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10{sup −4} mbar and 1.0 × 10{sup −5} mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10{sup −6} mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10{sup −5} mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ∼3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10{sup −5} mbar is achieved inside the cryostat. Baking of the

  13. Vacuum exhaust duct used for thermonuclear device

    International Nuclear Information System (INIS)

    Tachikawa, Nobuo; Kondo, Mitsuaki; Honda, Tsutomu.

    1990-01-01

    The present invention concerns a vacuum exhaust duct used for a thermonuclear device. A cylindrical metal liners is lined with a gap to the inside of a vacuum exhaust duct main body. Bellows are connected to both ends of the metal liners and the end of the bellows is welded to the vacuum exhaust duct main body. Futher, a heater is mounted to the metal liner on the side of the vacuum exhaust duct main body, and the metal liner is heated by the heater to conduct baking for the vacuum exhaust duct main body. Accordingly, since there is no requirement for elevating the temperature of the vacuum exhaust duct upon conducting baking, the vacuum exhaust duct scarcely suffers substantial deformation due to heat expansion. Further, there is also no substantial deformation for the bellows disposed between the outer circumference of the vacuum vessel and a portion of a vacuum exhaust duct, so that the durability of the bellows is greatly improved. (I.S.)

  14. Leybold vacuum handbook

    CERN Document Server

    Diels, K; Diels, Kurt

    1966-01-01

    Leybold Vacuum Handbook presents a collection of data sets that are essential for numerical calculation of vacuum plants and vacuum processes. The title first covers vacuum physics, which includes gas kinetics, flow phenomena, vacuum gauges, and vapor removal. Next, the selection presents data on vacuum, high vacuum process technology, and gas desorption and gettering. The text also deals with materials, vapor pressure, boiling and melting points, and gas permeability. The book will be of great interest to engineers and technicians that deals with vacuum related technologies.

  15. Design of the ZTH vacuum liner

    International Nuclear Information System (INIS)

    Prince, P.P.; Dike, R.S.

    1987-01-01

    The current status of the ZTH vacuum liner design is covered by this report. ZTH will be the first experiment to be installed in the CPRF (Confinement Physics Research Facility) at the Los Alamos National Laboratory and is scheduled to be operational at the rated current of 4 MA in 1992. The vacuum vessel has a 2.4 m major radius and a 40 cm minor radius. Operating parameters which drive the vacuum vessel mechanical design include a 300 C bakeout temperature, an armour support system capable of withstanding 25 kV, a high toroidal resistance, 1250 kPa magnetic loading, a 10 minute cycle time, and high positional accuracy with respect to the conducting shell. The vacuum vessel design features which satisfy the operating parameters are defined. The liner is constructed of Inconel 625 and has a geometry which alternates sections of thin walled bellows with rigid ribs. These composite sections span between pairs of the 16 diagnostic stations to complete the torus. The thin bellows sections maximize the liner toroidal resistance and the ribs provide support and positional accuracy for the armour in relation to the conducting shell. Heat transfer from the vessel is controlled by a blanket wrap of ceramic fiber insulation and the heat flux is dissipated to a water cooling jacket in the conducting shell

  16. Metallurgy of steels for PWR pressure vessels

    International Nuclear Information System (INIS)

    Kepka, M.; Mocek, J.; Barackova, L.

    1980-01-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure. (B.S.)

  17. Metallurgy of steels for PWR pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kepka, M; Mocek, J; Barackova, L [Skoda, Plzen (Czechoslovakia)

    1980-09-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure.

  18. Filling the vacuum at LHCb

    CERN Multimedia

    Katarina Anthony

    2013-01-01

    Last month, the Vacuum, Surfaces and Coatings (VSC) group was tasked with an unusually delicate operation in the LHCb experiment cavern: removing the LHC beam pipe while keeping the sensitive Vertex Locator vacuum vessel (VELO) completely isolated from the action.   The VSC group seal off the VELO beam pipe with a flange. Image: Gloria Corti. LHCb’s VELO detector is one of the crown jewels of the experiment. With detector elements surrounded by a vacuum, it gets as close as 5 cm from the beam. Fantastic for physics, but difficult for all-important access. “Because of the sensitivity of the VELO detector and its proximity to the beam, the collaboration decided not to bake (see box) its portion of the beam pipe,” says Giulia Lanza (TE-VSC-LBV), the expert in charge of the beam vacuum operation. “Our group was therefore asked to remove the rest of the LHC beam pipe while keeping the VELO portion of the pipe completely isolated. This work...

  19. MEA vacuum system

    International Nuclear Information System (INIS)

    Stroo, R.; Schwebke, H.; Heine, E.

    1984-01-01

    This report describes construction and operation of the MEA vacuum system of NIKHEF (Netherlands). First, the klystron vacuum system, beam transport system, diode pump and a triode pump are described. Next, the isolation valve and the fast valves of the vacuum system are considered. Measuring instruments, vacuum system commands and messages of failures are treated in the last chapter. (G.J.P.)

  20. Vacuum system for ISABELLE

    International Nuclear Information System (INIS)

    Hobson, J.P.

    1975-01-01

    An analysis is presented of the proposed vacuum system for the planned ISABELLE storage rings with respect to acceptability and practicality from the vacuum viewport. A comparison is made between the proposed vacuum system and the vacuum system at the CERN ISR, and some comments on various design and operational parameters are made

  1. Vacuum Bellows, Vacuum Piping, Cryogenic Break, and Copper Joint Failure Rate Estimates for ITER Design Use

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2010-06-01

    The ITER international project design teams are working to produce an engineering design in preparation for construction of the International Thermonuclear Experimental Reactor (ITER) tokamak. During the course of this work, questions have arisen in regard to safety barriers and equipment reliability as important facets of system design. The vacuum system designers have asked several questions about the reliability of vacuum bellows and vacuum piping. The vessel design team has asked about the reliability of electrical breaks and copper-copper joints used in cryogenic piping. Research into operating experiences of similar equipment has been performed to determine representative failure rates for these components. The following chapters give the research results and the findings for vacuum system bellows, power plant stainless steel piping (amended to represent vacuum system piping), cryogenic system electrical insulating breaks, and copper joints.

  2. Vacuum Systems Consensus Guideline for Department of Energy Accelerator Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Casey,R.; Haas, E.; Hseuh, H-C.; Kane, S.; Lessard, E.; Sharma, S.; Collins, J.; Toter, W. F.; Olis, D. R.; Pushka, D. R.; Ladd, P.; Jobe, R. K.

    2008-09-09

    Vacuum vessels, including evacuated chambers and insulated jacketed dewars, can pose a potential hazard to equipment and personnel from collapse, rupture due to back-fill pressurization, or implosion due to vacuum window failure. It is therefore important to design and operate vacuum systems in accordance with applicable and sound engineering principles. 10 CFR 851 defines requirements for pressure systems that also apply to vacuum vessels subject to back-fill pressurization. Such vacuum vessels are potentially subject to the requirements of the American Society of Mechanical Engineers (ASME) Pressure Vessel Code Section VIII (hereafter referred to as the 'Code'). However, the scope of the Code excludes vessels with internal or external operating pressure that do not exceed 15 pounds per square inch gauge (psig). Therefore, the requirements of the Code do not apply to vacuum systems provided that adequate pressure relief assures that the maximum internal pressure within the vacuum vessel is limited to less than 15 psig from all credible pressure sources, including failure scenarios. Vacuum vessels that cannot be protected from pressurization exceeding 15 psig are subject to the requirements of the Code. 10 CFR 851, Appendix A, Part 4, Pressure Safety, Section C addresses vacuum system requirements for such cases as follows: (c) When national consensus codes are not applicable (because of pressure range, vessel geometry, use of special materials, etc.), contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local code. Measures must include the following: (1) Design drawings, sketches, and calculations must be reviewed and approved by a qualified independent design professional (i.e., professional engineer). Documented organizational peer review is acceptable. (2) Qualified personnel must be used to perform examinations

  3. Design and test of-80 kV snubber core assemblies for MFTF sustaining-neutral-beam power supplies

    International Nuclear Information System (INIS)

    Bishop, S.R.; Mayhall, D.J.; Wilson, J.H.; De Vore, K.R.; Ross, R.I.; Sears, R.G.

    1981-01-01

    Core snubbers, located near the neutral beam source ends of the Mirror Fusion Test Facility (MFTF) Sustaining Neutral Beam Power Supply System (SNBPSS) source cables, protect the neutral beam source extractor grid wires from overheating and sputtering during internal sparkdowns. The snubbers work by producing an induced counter-emf which limits the fault current and by absorbing the capacitive energy stored on the 80 kV source cables and power supplies. A computer program STACAL was used in snubber magnetic design to choose appropriate tape wound cores to provide 400 Ω resistance and 25 J energy absorption. The cores are mounted horizontally in a dielectric structure. The central source cable bundle passes through the snubber and terminates on three copper buses. Multilam receptacles on the buses connect to the source module jumper cables. Corona rings and shields limit electric field stresses to allow close clearances between snubbers

  4. 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

    2016-11-15

    Highlights: • Eddy current distribution in the SST-1 vacuum vessel. • Circuit model analysis of eddy current. • A comparison of the field lines with and without the plasma column in identical conditions. • The influence of eddy current in magnetic NULL dynamics. - Abstract: Eddy current distribution in the vacuum vessel of the Steady state superconducting (SST-1) tokamak has been determined from the experimental data obtained using an array of internal voltage loops (flux loop) installed inside the vacuum vessel. A simple circuit model has been employed. The model takes into account the geometric and constructional features of SST-1 vacuum vessel. SST-1 vacuum vessel is a modified ‘D’ shaped vessel having major axis of 1.285 m and minor axis of 0.81 m and has been manufactured from non-magnetic stainless steel. The Plasma facing components installed inside the vacuum vessel are graphite blocks mounted on Copper Chromium Zirconium (CuCrZr) heat sink plates on inconel supports. During discharge of the central solenoid, eddy currents get generated in the vacuum vessel and passive supports on it. These eddy currents influence the early magnetic NULL dynamics and plasma break-down and start-up characteristics. The computed results obtained from the model have been benchmarked against experimental data obtained in large number of SST-1 plasma shots. The results are in good agreement. Once bench marked, the calculated eddy current based on flux loop signal and circuit equation model has been extended to the reconstruction of the overall B- field contours of SST-1 tokamak in the vessel region. A comparison of the field lines with and without the plasma column in identical conditions of the central solenoid and equilibrium field profiles has also been done with an aim to quantify the diagnostics responses in vacuum shots.

  5. Engineering study of the neutral beam and rf heating systems for DIII-D, MFTF-B, JET, JT-60 and TFTR

    International Nuclear Information System (INIS)

    Lindquist, W.B.; Staten, S.H.

    1987-01-01

    An engineering study was performed on the rf and neutral beam heating systems implemented for DIII-D, MFTF-B, JET, JT-60 and TFTR. Areas covered include: methodology used to implement the systems, technology, cost, schedule, performance, problems encountered and lessons learned. Systems are compared and contrasted in the areas studied. Summary statements were made on common problems and lessons learned. 3 refs., 6 tabs

  6. The JET high temperature in-vessel inspection system

    International Nuclear Information System (INIS)

    Businaro, T.; Cusack, R.; Calbiati, L.; Raimondi, T.

    1989-01-01

    The JET In-vessel Inspection System (IVIS) has been enhanced for operation under the following nominal conditions: vacuum vessel at 350 degC; vacuum vessel evacuated (∼10 -9 mbar); radiation dose during D-T phase 10 rads. The target resolution of the pictures is 2 mm at 5 m distance and tests on radiation resistance of the IVIS system are being carried out. Since June 1988, the new system is installed in the JET machine and the first inspections of the intire vessel at 250 degC have been satisfactory done. (author). 3 refs.; 6 figs.; 1 tab

  7. In-vessel maintenance remote manipulator system

    International Nuclear Information System (INIS)

    Jimenez, E.

    1978-01-01

    The radiation environment within the Tokamak Fusion Test Reactor (TFTR) vacuum vessel necessitates the development of a Remote Manipulator System (RMS) to perform required periodic inspection and maintenance tasks. The RMS must be able to perform dexterous operations and handle loads that exceed human capabilities. The limited size of the access ports on the TFTR vacuum vessel and the performance profile, defined by the various handling requirements, present unique design constraints. The design approach and formulation of a RMS configuration which satisfies TFTR requirements is presented herein

  8. Vacuum-assisted delivery

    Science.gov (United States)

    ... medlineplus.gov/ency/patientinstructions/000514.htm Vacuum-assisted delivery To use the sharing features on this page, ... through the birth canal. When is Vacuum-assisted Delivery Needed? Even after your cervix is fully dilated ( ...

  9. Relaxed plasma-vacuum systems

    International Nuclear Information System (INIS)

    Spies, G.O.; Lortz, D.; Kaiser, R.

    2001-01-01

    Taylor's theory of relaxed toroidal plasmas (states of lowest energy with fixed total magnetic helicity) is extended to include a vacuum between the plasma and the wall. In the extended variational problem, one prescribes, in addition to the helicity and the magnetic fluxes whose conservation follows from the perfect conductivity of the wall, the fluxes whose conservation follows from the assumption that the plasma-vacuum interface is also perfectly conducting (if the wall is a magnetic surface, then one has the toroidal and the poloidal flux in the vacuum). Vanishing of the first energy variation implies a pressureless free-boundary magnetohydrostatic equilibrium with a Beltrami magnetic field in the plasma, and in general with a surface current in the interface. Positivity of the second variation implies that the equilibrium is stable according to ideal magnetohydrodynamics, that it is a relaxed state according to Taylor's theory if the interface is replaced by a wall, and that the surface current is nonzero (at least if there are no closed magnetic field lines in the interface). The plane slab, with suitable boundary conditions to simulate a genuine torus, is investigated in detail. The relaxed state has the same double symmetry as the vessel if, and only if, the prescribed helicity is in an interval that depends on the prescribed fluxes. This interval is determined in the limit of a thin slab

  10. Vacuum leak test technique of JT-60

    International Nuclear Information System (INIS)

    Kaminaga, Atsushi; Arai, Takashi; Kodama, Kozo; Sasaki, Noboru; Saidoh, Masahiro

    1998-01-01

    Since a vacuum vessel of JT-60 is very large (167 m 3 ) and is combined with many components, such as magnetic coils, neutral beam injection systems and RF heating systems, etc., the position of leak testing exceeds 700. The two kind of techniques for vacuum leak test used in JT-60 has been described. Firstly the probe helium gas can be fed remotely in the three-dimensionally sectioned 54 regions of the JT-60 torus. The leak test was very rapidly performed by using this method. Secondly the helium detector system has been modified by the additional installation of the cryopump, which reduced the background level of the deuterium gas. The sensitivity of vacuum leak test with the cryopump was two orders of magnitude larger than that of without it. The examples of the performed vacuum leak test are stated. The vacuum leaks during experiments were 9 times. They were caused by thermal strain and plasma discharge. The vacuum leaks just after maintenance are 36 times which mainly caused by mis-installation. (author)

  11. Reconstruction of vacuum magnetic flux in QUEST

    International Nuclear Information System (INIS)

    Ishiguro, Masaki; Hanada, Kazuaki; Nakamura, Kazuo

    2010-01-01

    It is important to determine the best method for reconstructing the magnetic flux when eddy currents are significantly induced during magnetic measurement in spherical tokamaks (STs). Four methods for this reconstruction are investigated, and the calculated magnetic fluxes are compared to those measured in the cavity of a vacuum vessel. The results show that the best method is the one that uses currents from virtual coils for reconstruction. In this method, the placement of the virtual coils is optimized with numerical simulations using the Akaike information criterion (AIC), which indicates the goodness of fit of models used to fit measured data. The virtual coils are set on a line 15 cm outside the vacuum vessel. (author)

  12. Indian Vacuum Society: The Indian Vacuum Society

    Science.gov (United States)

    Saha, T. K.

    2008-03-01

    The Indian Vacuum Society (IVS) was established in 1970. It has over 800 members including many from Industry and R & D Institutions spread throughout India. The society has an active chapter at Kolkata. The society was formed with the main aim to promote, encourage and develop the growth of Vacuum Science, Techniques and Applications in India. In order to achieve this aim it has conducted a number of short term courses at graduate and technician levels on vacuum science and technology on topics ranging from low vacuum to ultrahigh vacuum So far it has conducted 39 such courses at different parts of the country and imparted training to more than 1200 persons in the field. Some of these courses were in-plant training courses conducted on the premises of the establishment and designed to take care of the special needs of the establishment. IVS also regularly conducts national and international seminars and symposia on vacuum science and technology with special emphasis on some theme related to applications of vacuum. A large number of delegates from all over India take part in the deliberations of such seminars and symposia and present their work. IVS also arranges technical visits to different industries and research institutes. The society also helped in the UNESCO sponsored post-graduate level courses in vacuum science, technology and applications conducted by Mumbai University. The society has also designed a certificate and diploma course for graduate level students studying vacuum science and technology and has submitted a syllabus to the academic council of the University of Mumbai for their approval, we hope that some colleges affiliated to the university will start this course from the coming academic year. IVS extended its support in standardizing many of the vacuum instruments and played a vital role in helping to set up a Regional Testing Centre along with BARC. As part of the development of vacuum education, the society arranges the participation of

  13. Containment vessel

    International Nuclear Information System (INIS)

    Zbirohowski-Koscia, K.F.; Roberts, A.C.

    1980-01-01

    A concrete containment vessel for nuclear reactors is disclosed that is spherical and that has prestressing tendons disposed in first, second and third sets, the tendons of each set being all substantially concentric and centred around a respective one of the three orthogonal axes of the sphere; the tendons of the first set being anchored at each end at a first anchor rib running around a circumference of the vessel, the tendons of the second set being anchored at each end at a second anchor rib running around a circumference of the sphere and disposed at 90 0 to the first rib, and the tendons of the third set being anchored some to the first rib and the remainder to the second rib. (author)

  14. Design of the ZTH vacuum liner

    International Nuclear Information System (INIS)

    Prince, P.P.

    1987-01-01

    The current status of the ZTH vacuum liner design is covered by this report. ZTH will be the first experiment to be installed in the CPRF (Confinement Physics Research Facility) at the Los Alamos National Laboratory and is scheduled to be operational at 2 MA in 1990 and at the rated current of 4 MA in 1992. The vacuum vessel has a 2.4m major radius and a 40 cm minor radius. The vacuum vessel design features which satisfy the operating parameters are defined. The liner is constructed of Inconel 625 and has a geometry which alternates sections of thin walled bellows with rigid ribs. These composite sections span between pairs of the 16 diagnostic stations to complete the torus. The thin bellows sections maximize the liner toroidal resistance and the ribs provide support and positional accuracy for the armour in relation to the conducting shell. Heat transfer from the vessel is controlled by a blanket wrap of ceramic fiber insulation and the heat flux is dissipated to a water cooling jacket in the conducting shell. 7 figs., 1 tab

  15. Magnetically induced vacuum decay

    International Nuclear Information System (INIS)

    Xue Shesheng

    2003-01-01

    We study the fermionic vacuum energy of vacua with and without application of an external magnetic field. The energetic difference of two vacua leads to the vacuum decaying and the vacuum energy being released. In the context of quantum field theories, we discuss why and how the vacuum energy can be released by spontaneous photon emission and/or paramagnetically screening the external magnetic field. In addition, we quantitatively compute the vacuum energy released, the paramagnetic screening effect, and the rate and spectrum of spontaneous photon emission. The possibilities of experimentally detecting such an effect of vacuum-energy release and that this effect accounts for the anomalous x-ray pulsar are discussed

  16. Flexibility of MFTF-B for thermal-barrier modifications and axisymmetric upgrades

    International Nuclear Information System (INIS)

    Thomassen, K.I.

    1981-01-01

    Flexibility in MFTB-B will be achieved partly by using the margins in particle and energy control designed into the machine and partly by making modest changes based on results obtained in TMX Upgrade. This latter flexibility is permitted by the schedule for vessel construction and component fabrication. The changes we might expect were determined by an examination of the processes involved in creating a thermal barrier and by speculating on the range of outcomes from TMX Upgrade experiments

  17. Gravitation and vacuum field

    International Nuclear Information System (INIS)

    Tevikyan, R.V.

    1986-01-01

    This paper presents equations that describe particles with spins s = 0, 1/2, 1 completely and which also describe 2s + 2 limiting fields as E → ∞. It is shown that the ordinary Hilbert-Einstein action for the gravitation field must be augmented by the action for the Bose vacuum field. This means that one must introduce in the gravitational equations a cosmological term proportional to the square of the strength of the Bose vacuum field. It is shown that the theory of gravitation describes three realities: matter, field, and vacuum field. A new form of matter--the vacuum field--is introduced into field theory

  18. Vacuum system and cleaning techniques in the FTU machines

    International Nuclear Information System (INIS)

    Alessandrini, C.; Apicella, M.L.; Ferro, C.

    1988-01-01

    FTU (Frascati Tokamak Upgrade) is a high magnetic field (8T) tokamak under construction at the Frascati Energy Research Center (ENEA). Its vacuum systems has been already manifactured and is presently being assembled. It consist of an all metallic fully welded vessel, pumped by six turbomolecular pumps. The vacuum system has been dimensioned to allow a base pressure lower than 2.6 x 10 -6 Pa. The paper reports the design philosophy of the vacuum system. The results of the cleaning techniques performed on a 1:1 scale toroidal sector of FTU are also presented and discussed

  19. Insulation vacuum and beam vacuum overpressure release

    CERN Document Server

    Parma, V

    2009-01-01

    There is evidence that the incident of 19th September caused a high pressure build-up inside the cryostat insulation vacuum which the existing overpressure devices could not contain. As a result, high longitudinal forces acting on the insulation vacuum barriers developed and broke the floor and the floor fixations of the SSS with vacuum barriers. The consequent large longitudinal displacements of the SSS damaged chains of adjacent dipole cryo-magnets. Estimates of the helium mass flow and the pressure build- up experienced in the incident are presented together with the pressure build-up for an even more hazardous event, the Maximum Credible Incident (MCI). The strategy of limiting the maximum pressure by the installation of addition pressure relieve devices is presented and discussed. Both beam vacuum lines were ruptured during the incident in sector 3-4 giving rise to both mechanical damage and pollution of the system. The sequence, causes and effects of this damage will be briefly reviewed. We will then an...

  20. Modern vacuum physics

    CERN Document Server

    Chambers, Austin

    2005-01-01

    Modern Vacuum Physics presents the principles and practices of vacuum science and technology along with a number of applications in research and industrial production. The first half of the book builds a foundation in gases and vapors under rarefied conditions, The second half presents examples of the analysis of representative systems and describes some of the exciting developments in which vacuum plays an important role. The final chapter addresses practical matters, such as materials, components, and leak detection. Throughout the book, the author''s explanations are presented in terms of first principles and basic physics, augmented by illustrative worked examples and numerous figures.

  1. Advanced toroidal facility vaccuum vessel stress analyses

    International Nuclear Information System (INIS)

    Hammonds, C.J.; Mayhall, J.A.

    1987-01-01

    The complex geometry of the Advance Toroidal Facility (ATF) vacuum vessel required special analysis techniques in investigating the structural behavior of the design. The response of a large-scale finite element model was found for transportation and operational loading. Several computer codes and systems, including the National Magnetic Fusion Energy Computer Center Cray machines, were implemented in accomplishing these analyses. The work combined complex methods that taxed the limits of both the codes and the computer systems involved. Using MSC/NASTRAN cyclic-symmetry solutions permitted using only 1/12 of the vessel geometry to mathematically analyze the entire vessel. This allowed the greater detail and accuracy demanded by the complex geometry of the vessel. Critical buckling-pressure analyses were performed with the same model. The development, results, and problems encountered in performing these analyses are described. 5 refs., 3 figs

  2. Vacuum technologies developed for at-400A Type B transportation and storage package

    International Nuclear Information System (INIS)

    Franklin, K.W.; Cockrell, G.D.

    1995-01-01

    The AT-400A TYPE B transportation and storage container will be used at Pantex Plant for the transportation and interim storage of plutonium pits. The AT-400A was designed by a joint effort between Sandia National Labs, Los Alamos National Labs, Lawrence Livermore National Laboratory, and Mason and Hanger-Silas Mason Co., Inc. In order to meet the requirements for transportation and storage, five different vacuum technologies had to be developed. The goals of the various vacuum technologies were to verify the plutonium pit was sealed, perform the assembly verification leak check in accordance with ANSI N-14.5 and to provide a final inert gas backfill in the containment vessel. This paper will discuss the following five vacuum technologies: (1) Pit Leak Testing, (2) Containment Vessel Purge and Backfill with tracer gas, (3) Containment Vessel Leak Testing, (4) Containment Vessel Purge and Final Backfill, and (5) Leak Testing of the Containment Vessel Gas Transfer tube

  3. Vacuum boundary modifications of the RFX-mod machine

    Energy Technology Data Exchange (ETDEWEB)

    Patel, Nisarg, E-mail: nisarg.patel@igi.cnr.it [University of Padova,Via 8 Febbraio 2, Padova 35122 (Italy); Consorzio RFX, Corso StatiUniti 4, Padova 35127 (Italy); Dalla Palma, Mauro; Dal Bello, Samuele; Grando, Luca; Peruzzo, Simone [Consorzio RFX, Corso StatiUniti 4, Padova 35127 (Italy); Sonato, Piergiorgio [University of Padova,Via 8 Febbraio 2, Padova 35122 (Italy); Consorzio RFX, Corso StatiUniti 4, Padova 35127 (Italy)

    2016-11-01

    Highlights: • Different vacuum sealing solutions are designed for cuts of Toroidal Support Structure. • New supporting rings are designed for In-TSS components. • Identified integration of the sub-assemblies. • Thermo-mechanical behaviour of components are verified against standard design rules. - Abstract: The results produced by experimental campaigns of RFX-mod shows the need to improve passive MHD control and minimise braking torque on plasma. These improvements require major mechanical changes on the present components of the machine including first wall, vacuum vessel, and toroidal support structure (TSS). The vacuum vessel will be removed and the first wall tiles will be directly supported by the passive stabilising shell, so increasing the poloidal cross section by 28 mm radially and bringing the passive stabilising shell as close as possible to the plasma boundary. This paper presents the mechanical design modifications of the torus assembly. Composite rings are designed to support the passive stabilising shell on the TSS. The vacuum boundary is shifted at the TSS by developing different joint solutions compatible with the stringent requirements of the present components: ceramic-metal brazed rings at the two poloidal joints, fully welded solution at the external equatorial joint, and resistive weld plate at the internal equatorial joint. The vacuum vessel portholes for pumping, fuelling, diagnostics, and sensor cable routing are redesigned to be integrated onto the TSS. The design is supported by thermo-mechanical analyses and verifications carried out applying load combinations corresponding to the expected load cases.

  4. Helical type vacuum container

    International Nuclear Information System (INIS)

    Owada, Kimio.

    1989-01-01

    Helical type vacuum containers in the prior art lack in considerations for thermal expansion stresses to helical coils, and there is a possibility of coil ruptures. The object of the present invention is to avoid the rupture of helical coils wound around the outer surface of a vacuum container against heat expansion if any. That is, bellows or heat expansion absorbing means are disposed to a cross section of a helical type vacuum container. With such a constitution, thermal expansion of helical coils per se due to temperature elevation of the coils during electric supply can be absorbed by expansion of the bellows or absorption of the heat expansion absorbing means. Further, this can be attained by arranging shear pins in the direction perpendicular to the bellows axis so that the bellows are not distorted when the helical coils are wound around the helical type vacuum container. (I.S.)

  5. Vacuum considerations: summary

    International Nuclear Information System (INIS)

    Blechschmidt, D.; Halama, H.J.

    1978-01-01

    A summary is given of the efforts of a vacuum systems study group of the workshop on a Heavy Ion Demonstration Experiment (HIDE) for heavy ion fusion. An inadequate knowledge of cross-sections prevents a more concrete vacuum system design. Experiments leading to trustworthy numbers for charge exchange, stripping and capture cross-sections are badly needed and should start as soon as possible. In linacs, beam loss will be almost directly proportional to the pressure inside the tanks. The tanks should, therefore, be built in such a way that they can be baked-out in situ to improve their vacuum, especially if the cross-sections turn out to be higher than anticipated. Using standard UHV techniques and existing pumps, an even lower pressure can be achieved. The vacuum system design for circular machines will be very difficult, and in some cases, beyond the present state-of-the-art

  6. Handbook of vacuum technology

    CERN Document Server

    2016-01-01

    This comprehensive, standard work has been updated to remain an important resource for all those needing detailed knowledge of the theory and applications of vacuum technology. With many numerical examples and illustrations to visualize the theoretical issues.

  7. Cold Vacuum Drying Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Located near the K-Basins (see K-Basins link) in Hanford's 100 Area is a facility called the Cold Vacuum Drying Facility (CVDF).Between 2000 and 2004, workers at the...

  8. Vacuum mechatronics first international workshop

    Energy Technology Data Exchange (ETDEWEB)

    Belinski, S.E.; Shirazi, M.; Hackwood, S.; Beni, G. (eds.) (California Univ., Santa Barbara, CA (USA))

    1989-01-01

    This report contains papers on the following topics: proposed epitaxial thin film growth in the ultra-vacuum of space; particle monitoring and control in vacuum processing equipment; electrostatic dust collector for use in vacuum systems; materials evaluation of an electrically noisy vacuum slip ring assembly; an overview of lubrication and associated materials for vacuum service; the usage of lubricants in a vacuum environment; guidelines and practical applications for lubrication in vacuum; recent development in leak detector and calibrator designs; the durability of ballscrews for ultrahigh vacuum; vacuum-compatible robot for self-contained manufacturing systems; the design, fabrication, and assembly of an advanced vacuum robotics system for space payload calibration; design criteria for mechanisms used in space; and concepts and requirements for semiconductor multiprocess integration in vacuum. These papers are indexed separately elsewhere.

  9. TFTR diagnostic vacuum controller

    International Nuclear Information System (INIS)

    Olsen, D.; Persons, R.

    1981-01-01

    The TFTR diagnostic vacuum controller (DVC) provides in conjunction with the Central Instrumentation Control and Data Acquisition System (CICADA), control and monitoring for the pumps, valves and gauges associated with each individual diagnostic vacuum system. There will be approximately 50 systems on TFTR. Two standard versions of the controller (A and B) wil be provided in order to meet the requirements of two diagnostic manifold arrangements. All pump and valve sequencing, as well as protection features, will be implemented by the controller

  10. Ultra high vacuum technology

    CERN Multimedia

    CERN. Geneva

    2001-01-01

    A short introduction for some basic facts and equations. Subsquently, discussion about: Building blocks of an ultrahigh vacuum system - Various types of pumps required to reach uhv and methods to reduce these effects - Outgassing phenomena induced by the presence of a particle beam and the most common methods to reduce these effects It will be given some practical examples from existing CERN accelerators and discuss the novel features of the future LHC vacuum system.

  11. A Planck Vacuum Cosmology

    Directory of Open Access Journals (Sweden)

    Daywitt W. C.

    2009-04-01

    Full Text Available Both the big-bang and the quasi-steady-state cosmologies originate in some type of Planck state. This paper presents a new cosmological theory based on the Planck- vacuum negative-energy state, a state consisting of a degenerate collection of negative- energy Planck particles. A heuristic look at the Einstein field equation provides a con- vincing argument that such a vacuum state could provide a theoretical explanation for the visible universe.

  12. Vessel Operator System

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Operator cards are required for any operator of a charter/party boat and or a commercial vessel (including carrier and processor vessels) issued a vessel permit from...

  13. Design of large aperture, low mass vacuum windows

    International Nuclear Information System (INIS)

    Leonhardt, W.J.; Mapes, M.

    1993-01-01

    Large vacuum vessels are employed downstream of fixed targets in High Energy Physics experiments to provide a long path for particles to traverse without interacting with air molecules. These vessels generally have a large aperture opening known as a vacuum window which employs a thin membrane to preserve the vacuum environment yet allows the particles to pass through with a minimal effect on them. Several large windows have been built using a composite of Kevlar/Mylar including circular windows to a diameter of 96.5 cm and rectangular windows up to 193 cm x 86 cm. This paper describes the design, fabrication, testing and operating experience with these windows and relates the actual performance to theoretical predictions

  14. Design of large aperture, low mass vacuum windows

    International Nuclear Information System (INIS)

    Leonhardt, W.J.; Mapes, M.

    1993-01-01

    Large vacuum vessels are employed downstream of fixed targets in High Energy Physics experiments to provide a long path for particles to transverse without interacting with air molecules. These vessels generally have a large aperture opening known as a open-quotes vacuum windowclose quotes which employs a thin membrane to preserve the vacuum environment yet allows the particles to pass through with a minimal effect on them. Several large windows have been built using a composite of Kevlar/Mylar including circular windows to a diameter of 96.5 cm and rectangular windows up to 193 cm x 86 cm. This paper describes the design, fabrication, testing and operating experience with these windows and relates the actual performance to theoretical predictions

  15. Application of local vacuum slide sealing electron beam welding procedure

    International Nuclear Information System (INIS)

    Sato, Shozo; Takano, Genta; Minami, Masaharu; Enami, Koji; Uchikawa, Takashi; Kuri, Shuhei

    1982-01-01

    Electron beam welding process is efficient and is superior in workmanship and its application to the welding of large plate structures is eagerly awaited. However, since electron beam welding is generally performed with the object of welding entirely put in a vacuum chamber, high welding cost becomes a problem. In response to this demand, two kinds of local vacuum slide sealing type electron beam welding machines have been developed. These welding machines are designed to perform welding with only the neighborhood of the weld line put in vacuum, one of which is for longitudinal joints and the other for circumferential joints. The welding machine for circumferential joints has been put to practical use for the welding of nucear fusion reactor vacuum vessels (outside diameter 3.5 m, inside diameter 1.7 m), showing that it is applicable to the welding of large structures. (author)

  16. Vacuum arc anode phenomena

    International Nuclear Information System (INIS)

    Miller, H.C.

    1976-01-01

    A brief review of anode phenomena in vacuum arcs is presented. Discussed in succession are: the transition of the arc into the anode spot mode; the temperature of the anode before, during and after the anode spot forms; and anode ions. Characteristically the anode spot has a temperature of the order of the atmospheric boiling point of the anode material and is a copious source of vapor and energetic ions. The dominant mechanism controlling the transition of the vacuum arc into the anode spot mode appears to depend upon the electrode geometry, the electrode material, and the current waveform of the particular vacuum arc being considered. Either magnetic constriction in the gap plasma or gross anode melting can trigger the transition; indeed, a combination of the two is a common cause of anode spot formation

  17. Visual tritium imaging of In-Vessel surfaces

    International Nuclear Information System (INIS)

    Gentile, C. A.; Zweben, S. J.; Skinner, C. H.; Young, K. M.; Langish, S. W.; Nishi, M. F.; Shu, W. M.; Parker, J.; Isobe, K.

    2000-01-01

    A imaging detector has been developed for the purpose of providing a non-destructive, real time method of determining tritium concentrations on the surface of internal TFTR vacuum vessel components. The detector employs a green phosphor screen (P31, zinc sulfide: copper) with a wave length peak of 530 nm, a charge-coupled device (CCD) camera linked to a computer, and a detection chamber for inserting components recovered from the vacuum vessel. This detector is capable of determining tritium concentrations on the surfaces. The detector provides a method of imaging tritium deposition on the surfaces in a fairly rapid fashion

  18. Simulation of Diffusive Lithium Evaporation Onto the NSTX Vessel Walls

    International Nuclear Information System (INIS)

    Stotler, D.P.; Skinner, C.H.; Blanchard, W.R.; Krstic, P.S.; Kugel, H.W.; Schneider, H.; Zakharov, L.E.

    2010-01-01

    A model for simulating the diffusive evaporation of lithium into a helium filled NSTX vacuum vessel is described and validated against an initial set of deposition experiments. The DEGAS 2 based model consists of a three-dimensional representation of the vacuum vessel, the elastic scattering process, and a kinetic description of the evaporated atoms. Additional assumptions are required to account for deuterium out-gassing during the validation experiments. The model agrees with the data over a range of pressures to within the estimated uncertainties. Suggestions are made for more discriminating experiments that will lead to an improved model.

  19. Visual tritium imaging of in-vessel surfaces

    International Nuclear Information System (INIS)

    Gentile, C.A.; Zweben, S.J.; Skinner, C.H.; Young, K.M.; Langish, S.W.; Nishi, M.F.; Shu, W.M.; Parker, J.; Isobe, K.

    2000-01-01

    An imaging detector has been developed for the purpose of providing a non-destructive, real time method of determining tritium concentrations on the surface of internal TFTR vacuum vessel components. The detector employs a green phosphor screen (P31, zinc sulfide: copper) with a wave length peak of 530 nm, a charge-coupled device (CCD) camera linked to a computer, and a detection chamber for inserting components recovered from the vacuum vessel. This detector is capable of determining tritium concentrations on the surfaces. The detector provides a method of imaging tritium deposition on the surfaces in a fairly rapid fashion

  20. Vacuum Arc Ion Sources

    CERN Document Server

    Brown, I.

    2013-12-16

    The vacuum arc ion source has evolved into a more or less standard laboratory tool for the production of high-current beams of metal ions, and is now used in a number of different embodiments at many laboratories around the world. Applications include primarily ion implantation for material surface modification research, and good performance has been obtained for the injection of high-current beams of heavy-metal ions, in particular uranium, into particle accelerators. As the use of the source has grown, so also have the operational characteristics been improved in a variety of different ways. Here we review the principles, design, and performance of vacuum arc ion sources.

  1. Vacuum fusion of uranium

    International Nuclear Information System (INIS)

    Stohr, J.A.

    1957-01-01

    After having outlined that vacuum fusion and moulding of uranium and of its alloys have some technical and economic benefits (vacuum operations avoid uranium oxidation and result in some purification; precision moulding avoids machining, chip production and chemical reprocessing of these chips; direct production of the desired shape is possible by precision moulding), this report presents the uranium fusion unit (its low pressure enclosure and pumping device, the crucible-mould assembly, and the MF supply device). The author describes the different steps of cast production, and briefly comments the obtained results

  2. Baryogenesis in false vacuum

    Energy Technology Data Exchange (ETDEWEB)

    Hamada, Yuta [KEK Theory Center, IPNS, KEK, Tsukuba, Ibaraki (Japan); Yamada, Masatoshi [Kanazawa University, Institute for Theoretical Physics, Kanazawa (Japan)

    2017-09-15

    The null result in the LHC may indicate that the standard model is not drastically modified up to very high scales, such as the GUT/string scale. Having this in the mind, we suggest a novel leptogenesis scenario realized in the false vacuum of the Higgs field. If the Higgs field develops a large vacuum expectation value in the early universe, a lepton number violating process is enhanced, which we use for baryogenesis. To demonstrate the scenario, several models are discussed. For example, we show that the observed baryon asymmetry is successfully generated in the standard model with higher-dimensional operators. (orig.)

  3. Vacuum considerations summary

    International Nuclear Information System (INIS)

    1977-01-01

    The vacuum system for Heavy Ion Fusion machines can be divided according to pressure into 4 parts: (a) Ion Sources; (b) Linear Accelerators; (c) Circular Accelerators, Accumulators and Storage Rings; and (d) Reactors. Since ion sources will need rather conventional pumping arrangements and reactors will operate with greater pressures, depending on their mode of operation, only items b and c will be treated in this report. In particular, the vacuum system design will be suggested for the machines proposed by various scenarios arrived at during the workshop. High mass numbers will be assumed

  4. Handbook of vacuum physics

    CERN Document Server

    1964-01-01

    Handbook of Vacuum Physics, Volume 3: Technology is a handbook of vacuum physics, with emphasis on the properties of miscellaneous materials such as mica, oils, greases, waxes, and rubber. Accurate modern tables of physical constants, properties of materials, laboratory techniques, and properties of commercial pumps, gauges, and leak detectors are presented. This volume is comprised of 12 chapters and begins with a discussion on pump oils, divided into rotary pump oils and vapor pump oils. The next chapter deals with the properties and applications of greases, including outgassing and vapor pr

  5. Vacuum phonon tunneling.

    Science.gov (United States)

    Altfeder, Igor; Voevodin, Andrey A; Roy, Ajit K

    2010-10-15

    Field-induced phonon tunneling, a previously unknown mechanism of interfacial thermal transport, has been revealed by ultrahigh vacuum inelastic scanning tunneling microscopy (STM). Using thermally broadened Fermi-Dirac distribution in the STM tip as in situ atomic-scale thermometer we found that thermal vibrations of the last tip atom are effectively transmitted to sample surface despite few angstroms wide vacuum gap. We show that phonon tunneling is driven by interfacial electric field and thermally vibrating image charges, and its rate is enhanced by surface electron-phonon interaction.

  6. Mirror Fusion Test Facility-B (MFTF-B) axicell configuration: NbTi magnet system. Design and analysis summary. Volume 1

    International Nuclear Information System (INIS)

    Heathman, J.H.; Wohlwend, J.W.

    1985-05-01

    This report summarizes the designs and analyses produced by General Dynamics Convair for the four Axicell magnets (A1 and A20, east and west), the four Transition magnets (T1 and T2, east and west), and the twelve Solenoid magnets (S1 through S6, east and west). Over four million drawings and specifications, in addition to detailed stress analysis, thermal analysis, electrical, instrumentation, and verification test reports were produced as part of the MFTF-B design effort. Significant aspects of the designs, as well as key analysis results, are summarized in this report. In addition, drawing trees and lists off detailed analysis and test reports included in this report define the locations of the detailed design and analysis data

  7. Mirror Fusion Test Facility-B (MFTF-B) axicell configuration: NbTi magnet system. Design and analysis summary. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Heathman, J.H.; Wohlwend, J.W.

    1985-05-01

    This report summarizes the designs and analyses produced by General Dynamics Convair for the four Axicell magnets (A1 and A20, east and west), the four Transition magnets (T1 and T2, east and west), and the twelve Solenoid magnets (S1 through S6, east and west). Over four million drawings and specifications, in addition to detailed stress analysis, thermal analysis, electrical, instrumentation, and verification test reports were produced as part of the MFTF-B design effort. Significant aspects of the designs, as well as key analysis results, are summarized in this report. In addition, drawing trees and lists off detailed analysis and test reports included in this report define the locations of the detailed design and analysis data.

  8. Nickel hydrogen multicell common pressure vessel battery development update

    Science.gov (United States)

    Zagrodnik, Jeffrey P.; Jones, Kenneth R.

    1992-01-01

    The technology background and design qualification of the multicell common pressure vessel nickel hydrogen battery are described. The results of full flight qualification, including random vibration at 19.5 g for two minutes in each axis, electrical characterization in a thermal vacuum chamber, and mass spectroscopy vessel leak detection are reviewed and 12.7 cm qualification and 25.4 cm design adaptation are discussed.

  9. LEP vacuum chamber, prototype

    CERN Multimedia

    CERN PhotoLab

    1983-01-01

    Final prototype for the LEP vacuum chamber, see 8305170 for more details. Here we see the strips of the NEG pump, providing "distributed pumping". The strips are made from a Zr-Ti-Fe alloy. By passing an electrical current, they were heated to 700 deg C.

  10. Cryogenic vacuum pump design

    International Nuclear Information System (INIS)

    Bartlett, A.J.; Lessard, P.A.

    1984-01-01

    This paper is a review of the problems and tradeoffs involved in cryogenic vacuum pump analysis, design and manufacture. Particular attention is paid to the several issues unique to cryopumps, e.g., radiation loading, adsorption of noncondensible gases, and regeneration. A general algorithm for cryopump design is also proposed. 12 references

  11. ISR vacuum system

    CERN Multimedia

    CERN PhotoLab

    1971-01-01

    Some of the most important components of the vacuum system are shown. At the left, the rectangular box is a sputter-ion pump inside its bake-out oven. The assembly in the centre includes a sector valve, three roughing valves, a turbomolecular pump, a rotary backing pump and auxiliary equipment. At the right, the small elbow houses a Bayard-

  12. ISR vacuum system

    CERN Multimedia

    CERN PhotoLab

    1970-01-01

    A pressure of 5 x 10-11 Torr has been obtained repreatedly in this pilot section of the ISR vacuum system. The pilot section is 45 m long is pumped by 9 sputter-ion pumps pf 350 l/s pumping speed, and is baked out at 200 degrees C before each pump down.

  13. LEP Vacuum Chamber

    CERN Multimedia

    1983-01-01

    This is a cut-out of a LEP vacuum chamber for dipole magnets showing the beam channel and the pumping channel with the getter (NEG) strip and its insulating supports. A water pipe connected to the cooling channel can also be seen at the back.The lead radiation shield lining is also shown. See also 8305563X.

  14. Scroll vacuum pump

    Energy Technology Data Exchange (ETDEWEB)

    Morishita, Etsuo; Suganami, Takuya; Nishida, Mitsuhiro; Kitora, Yoshihisa; Yamamoto, Sakuei; Fujii, Kosaburo

    1988-02-25

    An effort is made to apply a scroll machine to development of a vacuum pump. In view of mechanical simplification and load patterns, the vacuum pump uses a rotating mechanism to produce paired vortices rotating around each center. Chip seal and atmospheric pressure are utilized for axial gap sealing while a spring and atmospheric pressure for the radial gap sealing. In both gaps, the sealing direction is stationary relative to the environment during rotation, making it much easier to achieve effective sealing as compared to oscillating pumps. Since the compression ratio is high in vacuum pumps, a zero top clearance form is adopted for the central portion of vortices and an gas release valve is installed in the rotating axis. A compact Oldham coupling with a small inertia force is installed behind the vortices to maintain the required phase relations between the vortices. These improvements result in a vacuum of 1 Pa for dry operation and 10/sup -2/ Pa for oil flooded operation of a single-stage scroll machine at 1800 rpm. (5 figs, 1 tab, 4 refs)

  15. On Lovelock vacuum solution

    OpenAIRE

    Dadhich, Naresh

    2010-01-01

    We show that the asymptotic large $r$ limit of all Lovelock vacuum and electrovac solutions with $\\Lambda$ is always the Einstein solution in $d \\geq 2n+1$ dimensions. It is completely free of the order $n$ of the Lovelock polynomial indicating universal asymptotic behaviour.

  16. In-vessel remote maintenance of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref

  17. Commissioning result of the KSTAR in-vessel cryo-pump

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. B.; Lee, H. J.; Park, Y.M. [National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2013-12-15

    KSTAR in-vessel cryo-pump has been installed in the vacuum vessel top and bottom side with up-down symmetry for the better plasma density control in the D-shape H-mode. The cryogenic helium lines of the in-vessel cryo-pump are located at the vertical positions from the vacuum vessel torus center 2,000 mm. The inductive electrical potential has been optimized to reduce risk of electrical breakdown during plasma disruption. In-vessel cryo-pump consists of three parts of coaxial circular shape components; cryo-panel, thermal shield and particle shield. The cryo-panel is cooled down to below 4.5 K. The cryo-panel and thermal shields were made by Inconel 625 tube for higher mechanical strength. The thermal shields and their cooling tubes were annealed in air environment to improve the thermal radiation emissivity on the surface. Surface of cryo-panel was electro-polished to minimize the thermal radiation heat load. The in-vessel cryo-pump was pre-assembled on a test bed in 180 degree segment base. The leak test was carried out after the thermal shock between room temperature to LN2 one before installing them into vacuum vessel. Two segments were welded together in the vacuum vessel and final leak test was performed after the thermal shock. Commissioning of the in-vessel cryo-pump was carried out using a temporary liquid helium supply system.

  18. Multiple shell pressure vessel

    International Nuclear Information System (INIS)

    Wedellsborg, B.W.

    1988-01-01

    A method is described of fabricating a pressure vessel comprising the steps of: attaching a first inner pressure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about the first inner pressure vessel and attaching the second inner pressure vessel to the top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about the second inner pressure vessel and attaching the outer pressure vessel to the top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to the inlet and outlet openings in the first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to the inlet and outlet opening in the second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between the first inner pressure vessel and the second inner pressure vessel with material selected from the group, filling the space between the second inner pressure vessel and the outer pressure vessel with material selected from the group, and pressurizing the material filling the spaces between the pressure vessels to a predetermined pressure, the step comprising: pressurizing the spaces to a pressure whereby the wall of the first inner pressure vessel is maintained in compression during steady state operation of the pressure vessel

  19. High current vacuum closing switch

    International Nuclear Information System (INIS)

    Dolgachev, G.I.; Maslennikov, D.D.; Romanov, A.S.; Ushakov, A.G.

    2005-01-01

    The paper proposes a powerful pulsed closing vacuum switch for high current commutation consisting of series of the vacuum diodes with near 1 mm gaps having closing time determined by the gaps shortening with the near-electrode plasmas [ru

  20. Design of the vacuum control system for DIII-D

    International Nuclear Information System (INIS)

    Campbell, G.L.; Callis, R.W.; Haskovec, J.S.; Heckman, E.J.; Moore, C.D.; Scoville, J.T.

    1986-01-01

    The vacuum control and instrumentation for the DIII-D upgrade was designed using a new large programmable controller with color graphic operator interfaces and intelligent distributed devices. Remote, optically isolated input and output is used as well as optical isolation for the operator and programming consoles. Gate valves between experimental equipment and the vacuum vessel are interlocked for machine safety by an intelligent interface based upon a commercially available microcontroller card. Complete automatic operation with capability for remote operator intervention was one goal of this design effort. The design of the system with emphasis on the graphics, optical isolation and microcontroller implementation will be discussed

  1. Multistage charged particle accelerator, with high-vacuum insulation

    International Nuclear Information System (INIS)

    Holl, P.

    1976-01-01

    A multistage charged-particle accelerator for operating with accelerating voltages higher than 150 kV is described. The device consists essentially of a high-voltage insulator, a source for producing charged particles, a Wehnelt cylinder, an anode, and a post-accelerating tube containing stack-wise positioned post-accelerating electrodes. A high vacuum is used for insulating the parts carrying the high voltages, and at least one cylindrical screen surrounding these parts is interposed between them and the vacuum vessel, which can itself also function as a cylindrical screen

  2. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    Energy Technology Data Exchange (ETDEWEB)

    Urbonavičius, E., E-mail: Egidijus.Urbonavicius@lei.lt; Povilaitis, M., E-mail: Mantas.Povilaitis@lei.lt; Kontautas, A., E-mail: Aurimas.Kontautas@lei.lt

    2015-11-15

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  3. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    International Nuclear Information System (INIS)

    Urbonavičius, E.; Povilaitis, M.; Kontautas, A.

    2015-01-01

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  4. The symmetries of the vacuum

    International Nuclear Information System (INIS)

    Fleming, H.

    1985-01-01

    The vacuum equation of state required by cosmological inflation is taken seriously as a general property of the cosmological vacuum. This correctly restricts the class of theories which admit inflation. A model of such a vacuum is presented that leads naturally to the cosmological principle. (Author) [pt

  5. Elmo Bumpy Torus proof of principle, Phase II: Title 1 report. Volume II. Toroidal vessel

    International Nuclear Information System (INIS)

    1982-01-01

    The Toroidal Vessel provides the vacuum enclosure for containing the high temperature steady state plasma. In addition, the Toroidal Vessel must provide several viewing ports for plasma diagnostics, vacuum pumping ports for both high vacuum and roughing vacuum, feed-through ports for ECRH waveguides, limiter feed throughs for cooling and supporting the limiters, and ports for ion gages. The vessel must operate in an intense environment comprised of x-rays, microwaves, magnetic fields and plasma heat loads as well as the atmosphere pressure and gravity loads and the internal thermal stress loads due to heating and cooling of the torus. A key issue addressed was the choice of vacuum vessel seal and wall materials. In addition, during the course of the study, ORNL requested that horsecollar diagnostic ports be incorporated in the design. A comprehensive trade study was performed considering the vessel material issues in concert with the impact of the horsecollar port design. A change in baseline from an aluminum vessel with elastomer seals and circular diagnostic ports to austenitic stainless steel vessel with metal seals and horsecollar ports was agreed upon by both MDAC and ORNL towards the end of Title I

  6. Nonperturbative QED vacuum birefringence

    Energy Technology Data Exchange (ETDEWEB)

    Denisov, V.I.; Dolgaya, E.E.; Sokolov, V.A. [Physics Department, Moscow State University,Moscow, 119991 (Russian Federation)

    2017-05-19

    In this paper we represent nonperturbative calculation for one-loop Quantum Electrodynamics (QED) vacuum birefringence in presence of strong magnetic field. The dispersion relations for electromagnetic wave propagating in strong magnetic field point to retention of vacuum birefringence even in case when the field strength greatly exceeds Sauter-Schwinger limit. This gives a possibility to extend some predictions of perturbative QED such as electromagnetic waves delay in pulsars neighbourhood or wave polarization state changing (tested in PVLAS) to arbitrary magnetic field values. Such expansion is especially important in astrophysics because magnetic fields of some pulsars and magnetars greatly exceed quantum magnetic field limit, so the estimates of perturbative QED effects in this case require clarification.

  7. Compact vacuum insulation embodiments

    Science.gov (United States)

    Benson, D.K.; Potter, T.F.

    1992-04-28

    An ultra-thin compact vacuum insulation panel is comprised of two hard, but bendable metal wall sheets closely spaced apart from each other and welded around the edges to enclose a vacuum chamber. Glass or ceramic spacers hold the wall sheets apart. The spacers can be discrete spherical beads or monolithic sheets of glass or ceramic webs with nodules protruding therefrom to form essentially point' or line' contacts with the metal wall sheets. In the case of monolithic spacers that form line' contacts, two such spacers with the line contacts running perpendicular to each other form effectively point' contacts at the intersections. Corrugations accommodate bending and expansion, tubular insulated pipes and conduits, and preferred applications are also included. 26 figs.

  8. Compact vacuum insulation

    Science.gov (United States)

    Benson, D.K.; Potter, T.F.

    1993-01-05

    An ultra-thin compact vacuum insulation panel is comprised of two hard, but bendable metal wall sheets closely spaced apart from each other and welded around the edges to enclose a vacuum chamber. Glass or ceramic spacers hold the wall sheets apart. The spacers can be discrete spherical beads or monolithic sheets of glass or ceramic webs with nodules protruding therefrom to form essentially point'' or line'' contacts with the metal wall sheets. In the case of monolithic spacers that form line'' contacts, two such spacers with the line contacts running perpendicular to each other form effectively point'' contacts at the intersections. Corrugations accommodate bending and expansion, tubular insulated pipes and conduits, and preferred applications are also included.

  9. Dry vacuum pumps

    International Nuclear Information System (INIS)

    Sibuet, R

    2008-01-01

    For decades and for ultimate pressure below 1 mbar, oil-sealed Rotary Vane Pumps have been the most popular solution for a wide range of vacuum applications. In the late 80ies, Semiconductor Industry has initiated the development of the first dry roughing pumps. Today SC applications are only using dry pumps and dry pumping packages. Since that time, pumps manufacturers have developed dry vacuum pumps technologies in order to make them attractive for other applications. The trend to replace lubricated pumps by dry pumps is now spreading over many other market segments. For the Semiconductor Industry, it has been quite easy to understand the benefits of dry pumps, in terms of Cost of Ownership, process contamination and up-time. In this paper, Technology of Dry pumps, its application in R and D/industries, merits over conventional pumps and future growth scope will be discussed

  10. Probabilistic retinal vessel segmentation

    Science.gov (United States)

    Wu, Chang-Hua; Agam, Gady

    2007-03-01

    Optic fundus assessment is widely used for diagnosing vascular and non-vascular pathology. Inspection of the retinal vasculature may reveal hypertension, diabetes, arteriosclerosis, cardiovascular disease and stroke. Due to various imaging conditions retinal images may be degraded. Consequently, the enhancement of such images and vessels in them is an important task with direct clinical applications. We propose a novel technique for vessel enhancement in retinal images that is capable of enhancing vessel junctions in addition to linear vessel segments. This is an extension of vessel filters we have previously developed for vessel enhancement in thoracic CT scans. The proposed approach is based on probabilistic models which can discern vessels and junctions. Evaluation shows the proposed filter is better than several known techniques and is comparable to the state of the art when evaluated on a standard dataset. A ridge-based vessel tracking process is applied on the enhanced image to demonstrate the effectiveness of the enhancement filter.

  11. Temperature control in vacuum

    International Nuclear Information System (INIS)

    Dearnaley, G.

    1986-01-01

    The patent concerns a method for controlling the temperature of silicon wafers (or samples), during ion beam treatment of the wafers, in a vacuum. The apparatus and method are described for irradiation and temperature control of the samples. The wafers are mounted on a drum which is rotated through the ion beam, and are additionally heated by infra-red lamps to achieve the desired temperature. (U.K.)

  12. Electroweak vacuum geometry

    International Nuclear Information System (INIS)

    Lepora, N.; Kibble, T.

    1999-01-01

    We analyse symmetry breaking in the Weinberg-Salam model paying particular attention to the underlying geometry of the theory. In this context we find two natural metrics upon the vacuum manifold: an isotropic metric associated with the scalar sector, and a squashed metric associated with the gauge sector. Physically, the interplay between these metrics gives rise to many of the non-perturbative features of Weinberg-Salam theory. (author)

  13. Vacuum inhomogeneous cosmological models

    International Nuclear Information System (INIS)

    Hanquin, J.-L.

    1984-01-01

    The author presents some results concerning the vacuum cosmological models which admit a 2-dimensional Abelian group of isometries: classifications of these space-times based on the topological nature of their space-like hypersurfaces and on their time evolution, analysis of the asymptotical behaviours at spatial infinity for hyperbolical models as well as in the neighbourhood of the singularity for the models possessing a time singularity during their evolution. (Auth.)

  14. EAST-AIA deployment under vacuum: Calibration of laser diagnostic system using computer vision

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yang, E-mail: yangyang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, 350 Shushanhu Rd, Hefei, Anhui (China); Song, Yuntao; Cheng, Yong; Feng, Hansheng; Wu, Zhenwei; Li, Yingying; Sun, Yongjun; Zheng, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, 350 Shushanhu Rd, Hefei, Anhui (China); Bruno, Vincent; Eric, Villedieu [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France)

    2016-11-15

    Highlights: • The first deployment of the EAST articulated inspection arm robot under vacuum is presented. • A computer vision based approach to measure the laser spot displacement is proposed. • An experiment on the real EAST tokamak is performed to validate the proposed measure approach, and the results shows that the measurement accuracy satisfies the requirement. - Abstract: For the operation of EAST tokamak, it is crucial to ensure that all the diagnostic systems are in the good condition in order to reflect the plasma status properly. However, most of the diagnostic systems are mounted inside the tokamak vacuum vessel, which makes them extremely difficult to maintain under high vacuum condition during the tokamak operation. Thanks to a system called EAST articulated inspection arm robot (EAST-AIA), the examination of these in-vessel diagnostic systems can be performed by an embedded camera carried by the robot. In this paper, a computer vision algorithm has been developed to calibrate a laser diagnostic system with the help of a monocular camera at the robot end. In order to estimate the displacement of the laser diagnostic system with respect to the vacuum vessel, several visual markers were attached to the inner wall. This experiment was conducted both on the EAST vacuum vessel mock-up and the real EAST tokamak under vacuum condition. As a result, the accuracy of the displacement measurement was within 3 mm under the current camera resolution, which satisfied the laser diagnostic system calibration.

  15. ELETTRA vacuum system

    International Nuclear Information System (INIS)

    Bernardini, M.; Daclon, F.; Giacuzzo, F.; Miertusova, J.; Pradal, F.; Kersevan, R.

    1993-01-01

    Elettra is a third-generation synchrotron light source which is being built especially for the use of high brilliance radiation from insertion devices and bending magnets. The UHV conditions in a storage ring lead to a longer beam lifetime - one of the most important criterion. The Elettra vacuum system presents some pecularities which cannot be found in any already existing machine. The final version of bending magnet vacuum chamber is presented. After chemical and thermal conditioning the specific outgassing rate of about 1.5e-12 Torr. liters sec -1 cm -2 was obtained. A microprocessor-controlled system has been developed to perform bake-out at the uniform temperature. The etched-foil type heaters are glued to the chamber and Microtherm insulation is used. UHV pumps based on standard triode sputter-ion pumps were modified with ST 707 NEG (Non Evaporable Getter) modules. A special installation enables the resistive activation of getters and significantly increases pumping speed for hydrogen and other residual gases (except methane and argon). All these technological innovations improve vacuum conditions in Elettra storage ring and consequently also the other parameters of the light source

  16. PARAFFIN SEPARATION VACUUM DISTILLATION

    Directory of Open Access Journals (Sweden)

    Zaid A. Abdulrahman

    2013-05-01

    Full Text Available Simulated column performance curves were constructed for existing paraffin separation vacuum distillation column in LAB plant (Arab Detergent Company/Baiji-Iraq. The variables considered in this study are the thermodynamic model option, top vacuum pressure, top and bottom temperatures, feed temperature, feed composition & reflux ratio. Also simulated columns profiles for the temperature, vapor & liquid flow rates composition were constructed. Four different thermodynamic model options (SRK, TSRK, PR, and ESSO were used, affecting the results within 1-25% variation for the most cases.The simulated results show that about 2% to 8 % of paraffin (C10, C11, C12, & C13 present at the bottom stream which may cause a problem in the LAB plant. The major variations were noticed for the top temperature & the  paraffin weight fractions at bottom section with top vacuum pressure. The bottom temperature above 240 oC is not recommended because the total bottom flow rate decreases sharply, where as  the weight fraction of paraffins decrease slightly. The study gives evidence about a successful simulation with CHEMCAD

  17. Vacuum system for LHC

    International Nuclear Information System (INIS)

    Groebner, O.

    1995-01-01

    The Large Hadron Collider (LHC) which is planned at CERN will be housed in the tunnel of the Large Electron Positron collider (LEP) and will store two counter-rotating proton beams with energies of up to 7 TeV in a 27 km accelerator/storage ring with superconducting magnets. The vacuum system for the LHC will be at cryogenic temperatures (between 1.9 and 20 K) and will be exposed to synchrotron radiation emitted by the protons. A stringent limitation on the vacuum is given by the energy deposition in the superconducting coils of the magnets due to nuclear scattering of the protons on residual gas molecules because this may provoke a quench. This effect imposes an upper limit to a local region of increased gas density (e.g. a leak), while considerations of beam lifetime (100 h) will determine more stringent requirements on the average gas density. The proton beam creates ions from the residual gas which may strike the vacuum chamber with sufficient energy to lead to a pressure 'run-away' when the net ion induced desorption yield exceeds a stable limit. These dynamic pressure effects will be limited to an acceptable level by installing a perforated 'beam screen' which shields the cryopumped gas molecules at 1.9 K from synchrotron radiation and which also absorbs the synchrotron radiation power at a higher and, therefore, thermodynamically more efficient temperature. (author)

  18. Anomalous vacuum expectation values

    International Nuclear Information System (INIS)

    Suzuki, H.

    1986-01-01

    The anomalous vacuum expectation value is defined as the expectation value of a quantity that vanishes by means of the field equations. Although this value is expected to vanish in quantum systems, regularization in general produces a finite value of this quantity. Calculation of this anomalous vacuum expectation value can be carried out in the general framework of field theory. The result is derived by subtraction of divergences and by zeta-function regularization. Various anomalies are included in these anomalous vacuum expectation values. This method is useful for deriving not only the conformal, chiral, and gravitational anomalies but also the supercurrent anomaly. The supercurrent anomaly is obtained in the case of N = 1 supersymmetric Yang-Mills theory in four, six, and ten dimensions. The original form of the energy-momentum tensor and the supercurrent have anomalies in their conservation laws. But the modification of these quantities to be equivalent to the original one on-shell causes no anomaly in their conservation laws and gives rise to anomalous traces

  19. Improvement to reactor vessel

    International Nuclear Information System (INIS)

    1974-01-01

    The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

  20. Vacuum pumping concepts for ETF

    International Nuclear Information System (INIS)

    Homeyer, W.G.

    1980-09-01

    The Engineering Test Facility (ETF) poses unique vacuum pumping requirements due to its large size and long burn characteristics. These requirements include torus vacuum pumping initially and between burns and pumping of neutralized gas from divertor collector chambers. It was found that the requirements could be met by compound cryopumps in which molecular sieve 5A is used as the cryosorbent. The pumps, ducts, and vacuum valves required are large but fit with other ETF components and do not require major advances in vacuum pumping technology. Several additional design, analytical, and experimental studies were identified as needed to optimize designs and provide better design definition for the ETF vacuum pumping systems