Automatic and manual operation modes of the TFTR maintenance manipulator

International Nuclear Information System (INIS)

Boehme, G.; Gumb, L.; Lotz, E.; Mueller, G.; Selig, M.

1987-01-01

The remote in-vessel operations scheduled to maintain the TFTR at Princeton, NJ, USA, comprise inspection, calibration, cleaning and protective tile replacement. The environmental conditions inside the torus vessel are ultra high vacuum, moderate γ-radiation and 150 0 C temperature of the vessel structure. The Princeton Plasma Physics Laboratory (PPPL) and KfK are jointly developing a maintenance manipulator (MM) which can perform these tasks. (orig./HP)

TFTR materials issues and problems during design and construction

International Nuclear Information System (INIS)

Sabado, M.; Little, R.

1984-01-01

TFTR as well as its contemporaries, T15, JT60, and JET, have important contributions to make towards our understanding of plasma conditions in the thermonuclear regime. One of the main objectives of TFTR is to produce fusion power densities approaching those in a fusion reactor, approx.= 1 Wcm -3 at Q approx.= 1-2. TFTR will be the first tokamak to routinely use deuterium tritium, and produce approx.= 10 19 fusion neutrons per pulse. With startup of TFTR on December 24, 1982, the demonstration of physics feasibility of 'breakeven' is close at hand. Since TFTR performance will be reactor relevant, the capability of materials/components to withstand the hostile effects of a plasma environment will be presented. It is intended that designers of future fusion devices benefit from the materials technology developments and applications on TFTR. In an attempt to comply with this mandate, this paper will describe TFTR issues on materials, their developments, selections, problems, and solutions. Special emphasis will be given, in particular, to the impurity control devices in TFTR, namely, the limiter and surface pumping system located inside the vacuum vessel. The plasma will interact with these components and they will be subjected to disruptions, a vacuum of 10 -6 to 10 -8 torr and a nominal temperatures of 0 C. 'Painful' materials development problems encountered will be reviewed, as well as important 'lessons learned'. A briefing on the materials of construction will be given, with some comments on the problems that developed and their solutions. (orig.)

Calibration issues of the TFTR multichannel neutron collimator

International Nuclear Information System (INIS)

Goeler, S. von; Johnson, L.C.; Bitter, M.; Efthimion, P.C.; Roquemore, A.L.

1996-01-01

The calibration procedures for the detectors in the Neutron Collimator are reviewed. The absolute calibration was performed for the NE451 detectors, in situ, by moving a DT neutron generator in the TFTR vacuum vessel across each sight line. This calibration was transferred to other detectors in the same channel. Four new sight lines have been installed at a different toroidal location, which view the plasma through the vacuum vessel port cover rather than through thinned windows. The new detectors are cross-calibrated to the NE451 detectors with a jog shot procedure, where the plasma is quickly shifted in major radius over a distance of 30 cm. The jog shot procedure shows that scattered neutrons account approximately for 30% of the signal of the new central channels. The neutron source strength from the collimator agrees within 10% with the source strength from global neutron monitors in the TFTR test cell. Detector non-linearity is discussed. Another special issue is the behavior of the detectors during T-puffs, where the DD/DT neutron ratio changes rapidly

Engineering test station for TFTR blanket module experiments

International Nuclear Information System (INIS)

Jassby, D.L.; Leinoff, S.

1979-12-01

A conceptual design has been carried out for an Engineering Test Station (ETS) which will provide structural support and utilities/instrumentation services for blanket modules positioned adjacent to the vacuum vessel of the TFTR (Tokamak Fusion Test Reactor). The ETS is supported independently from the Test Cell floor. The ETS module support platform is constructed of fiberglass to eliminate electromagnetic interaction with the pulsed tokamak fields. The ETS can hold blanket modules with dimensions up to 78 cm in width, 85 cm in height, and 105 cm in depth, and with a weight up to 4000 kg. Interfaces for all utility and instrumentation requirements are made via a shield plug in the TFTR igloo shielding. The modules are readily installed or removed by means of TFTR remote handling equipment

Vacuum system design and tritium inventory for the TFTR charge exchange diagnostic

International Nuclear Information System (INIS)

Medley, S.S.

1979-05-01

The charge exchange diagnostic for the TFTR is comprised of two analyzer systems which contain a total of twenty independent mass/energy analyzers and one diagnostic neutral beam tentatively rated at 80 keV, 15 A. The associated vacuum systems were analyzed using the Vacuum System Transient Simulator (VSTS) computer program which models the transient transport of multi-gas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced cost, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus

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

1987 calibration of the TFTR neutron spectrometers

International Nuclear Information System (INIS)

Barnes, C.W.; Strachan, J.D.; Princeton Univ., NJ

1989-12-01

The 3 He neutron spectrometer used for measuring ion temperatures and the NE213 proton recoil spectrometer used for triton burnup measurements were absolutely calibrated with DT and DD neutron generators placed inside the TFTR vacuum vessel. The details of the detector response and calibration are presented. Comparisons are made to the neutron source strengths measured from other calibrated systems. 23 refs., 19 figs., 6 tabs

Development of a maintenance manipulator for TFTR

International Nuclear Information System (INIS)

Holloway, C.

1986-01-01

The maintenance manipulator is a device permanently connected to the Tokamak Fusion Test Reactor (TFTR) vacuum vessel and is located in close proximity to the tokamak. It is used for the inspection and maintenance of in-vessel components whilst the machine remains under vacuum. The total system comprises a vacuum vessel ante-chamber that houses the manipulator, an articulated boom and carriage that transports and positions a dexterous end-effector, and end-effector that supports maintenance tooling, and an inspection system. Because of the maintenance manipulator's operating environment, there are many challenging engineering features, i.e., temperatures up to 150 0 C, changing magnetic fields in space and time that act on the manipulator whilst it is at rest, neutron neutron fluxes of up to 10/sup 11/cm/sup -2/s/sup -1/, and, last but not least, UHV conditions. This paper describes the development of the vacuum system, the maintenance manipulator, and inspective devices. It includes the methods employed to overcome the engineering difficulties and the application of information gained from other advanced technology programs, such as space and nuclear fission

In situ calibration of TFTR neutron detectors

International Nuclear Information System (INIS)

Hendel, H.W.; Palladino, R.W.; Barnes, C.W.; Diesso, M.; Felt, J.S.; Jassby, D.L.; Johnson, L.C.; Ku, L.; Liu, Q.P.; Motley, R.W.; Murphy, H.B.; Murphy, J.; Nieschmidt, E.B.; Roberts, J.A.; Saito, T.; Strachan, J.D.; Waszazak, R.J.; Young, K.M.

1990-01-01

We report results of the TFTR fission detector calibration performed in December 1988. A NBS-traceable, remotely controlled 252 Cf neutron source was moved toroidally through the TFTR vacuum vessel. Detection efficiencies for two 235 U detectors were measured for 930 locations of the neutron point source in toroidal scans at 16 different major radii and vertical heights. These scans effectively simulated the volume-distributed plasma neutron source and the volume-integrated detection efficiency was found to be insensitive to plasma position. The Campbell mode is useful due to its large overlap with the count rate mode and large dynamic range. The resulting absolute plasma neutron source calibration has an uncertainty of ±13%

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.)

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

Long Term Tritium Trapping in TFTR and JET

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.A.; Young, K.M.; Coad, J.P.; Hogan, J.T.; Penzhorn, R.-D.; Bekris, N.

2001-01-01

Tritium retention in TFTR [Tokamak Fusion Test Reactor] and JET [Joint European Torus] shows striking similarities and contrasts. In TFTR, 5 g of tritium were injected into circular plasmas over a 3.5 year period, mostly by neutral-beam injection. In JET, 35 g were injected into divertor plasmas over a 6 month campaign, mostly by gas puffing. In TFTR, the bumper limiter provided a large source of eroded carbon and a major part of tritium was co-deposited on the limiter and vessel wall. Only a small area of the co-deposit flaked off. In JET, the wall is a net erosion area, and co-deposition occurs principally in shadowed parts of the inner divertor, with heavy flaking. In both machines, the initial tritium retention, after a change from deuterium [D] to tritium [T] gas puffing, is high and is due to isotope exchange with deuterium on plasma-facing surfaces (dynamic inventory). The contribution of co-deposition is lower but cumulative, and is revealed by including periods of D fueling that reversed the T/D isotope exchange. Ion beam analysis of flakes from TFTR showed an atomic D/C ratio of 0.13 on the plasma facing surface, 0.25 on the back surface and 0.11 in the bulk. Data from a JET divertor tile showed a larger D/C ratio with 46% C, 30% D, 20% H and 4% O. Deuterium, tritium, and beryllium profiles have been measured and show a thin less than 50 micron co-deposited layer. Flakes retrieved from the JET vacuum vessel exhibited a high tritium release rate of 2e10 Bq/month/g. BBQ modeling of the effect of lithium on retention in TFTR showed overlapping lithium and tritium implantation and a 1.3x increase in local T retention

Observations of Flaking of Co-deposited Layers in TFTR

International Nuclear Information System (INIS)

Gentile, C.A.; Skinner, C.H.; Young, K.M.

1999-01-01

Flaking of co-deposited layers in the Tokamak Fusion Test Reactor (TFTR) has been observed after the termination of plasma operations. This unexpected flaking affects approximately 15% of the tiles and appears on isotropic graphite tiles but not on carbon fiber composite tiles. Samples of tiles, flakes and dust were recently collected from the inside of the vacuum vessel and will be analyzed to better characterize the behavior of tritium on plasma facing components in DT fusion devices

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.)

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.)

Remote maintenance of in-vessel components in Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Loesser, G.D.; Heitzenroeder, P.; Kungl, D.; Dylla, H.F.; Cerdan, G.

1990-01-01

The Tokamak Fusion Test Reactor (TFTR) will generate a total of 3 x 10 21 neutrons during its planned D-T operational period. A maintenance manipulator has been designed and tested to minimize personnel radiation during in-vessel maintenance activities. Its functions include visual inspection, first-wall tile replacement, cleaning, diagnostics calibrations and leak detection. To meet these objectives, the TFTR maintenance manipulator is required to be operable in the TFTR high vacuum environment, typically -8 torr, ( -6 Pa). Geometrically, the manipulator must extend 180 0 in either direction around the torus to assure complete coverage of the vessel first-wall. The manipulator consists of a movable carriage, and movable articulated link sections which are driven by electrical actuators. The boom has vertical load capacity of 455 kg and lateral load capacity of 46 kg. The boom can either be fitted with a general inspection arm or dextrous slave arms. The general inspection arm is designed to hold the leak detector and an inspection camera; it is capable of rotation along two axes and has a linkage system which permits motion normal to the vacuum vessel wall. All systems except the dextrous slave arms are operable in a vacuum. (author)

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.)

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.)

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.)

In-situ calibration of TFTR [Tokamak Fusion Test Reactor] neutron detectors

International Nuclear Information System (INIS)

Hendel, H.W.; Palladino, R.W.; Barnes, C.W.; Diesso, M.; Felt, J.S.; Jassby, D.L.; Johnson, L.C.; Ku, L.P.; Liu, Q.P.; Motley, R.W.; Murphy, H.B.; Murphy, J.; Nieschmidt, E.B.; Roberts, J.A.; Saito, T.; Strachan, J.D.; Waszazak, R.J.; Young, K.

1990-03-01

We report results of the TFTR fission detector calibration performed in December 1988. A NBS-traceable, remotely controlled 252 Cf neutron source was moved toroidally through the TFTR vacuum vessel. Detection efficiencies for two 235 U detectors were measured for 930 locations of the neutron point source in toroidal scans at 16 different major radii and vertical heights. These scans effectively simulated the volume-distributed plasma neutron source, and the volume-integrated detection efficiency was found to be insensitive to plasma position. The Campbell mode is useful due to its large overlap with the count rate mode and large dynamic range. The resulting absolute plasma neutron source calibration has an uncertainty of ± 13%. 21 refs., 23 figs., 4 tabs

Assessment of eddy current effects on compression experiments in the TFTR tokamak

International Nuclear Information System (INIS)

Wong, K.L.; Park, W.

1986-05-01

The eddy current induced on the TFTR vacuum vessel during compression experiments is estimated based on a cylindrical model. It produces an error magnetic field that generates magnetic islands at the rational magnetic surfaces. The widths of these islands are calculated and found to have some effect on electron energy confinement. However, resistive MHD simulation results indicate that the island formation process can be slowed down by plasma rotation

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

Operations analysis of the unscheduled summer machine opening of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory

International Nuclear Information System (INIS)

Viola, M.E.; McCann, J.

1985-01-01

During experimental operation, a problem developed with the mechanical integrity of the TFTR surface pumping system neutralizer plates that required a vacuum vessel entry for repairs. This problem, coupled with several less significant machine internal problems that had been developing, forced the decision to make an unscheduled vacuum vessel entry. An extended machine outage at that time would have had a severe impact on the experimental schedule. Therefore, the goal was to make repairs and return the vacuum vessel to a clean condition as quickly as possible. The total time required between the end of regularly scheduled activity and restoration of the machine capability to routinely obtain 1 MA disruption-free plasma was 12 days

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

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.

TFTR bumper limiter and final protective plate engineering, fabrication and assembly

International Nuclear Information System (INIS)

Helmich, R.C.; Snook, P.G.; Loesser, G.D.; Reilly, T.B.; Trachsel, C.A.

1986-01-01

The inner vacuum vessel wall of the Tokamak Fusion Test Reactor (TFTR) is protected from plasma impingement by a bumper limiter and from neutral beam bombardment by protective plates. Engineering problems and solutions relating to Inconel 718, such as welding, machining in the annealed or age-hardened condition, selection of feeds, speeds and the need for rigid tooling are discussed. Vacuum furnace brazing of the 5/16'' Inconel 600 cooling tubing to the backing plates in both horizontal and vertical sections are shown. A detailed description of the plate and tile fabrication and assembly, with manufacturing and management techniques is outlined in this paper

Study of optically thin electron cyclotron emission from TFTR using a Michelson interferometer

International Nuclear Information System (INIS)

Stauffer, F.J.; Boyd, D.A.

1986-01-01

The TFTR Michelson interferometer, which is used as an electron temperature diagnostic, has a spectral range of 75-540 GHz. This range is adequate for measuring at least the first three cyclotron harmonics, and it spans both optically thick and thin portions of the ECE spectrum. During the most recent opening of the TFTR vacuum vessel, a concave, carbon reflector was installed on the back wall of the vessel, opposite the light collecting optic of the Michelson system. The reflector is designed to prevent the observation of optically thin ECE that originates from a location that is outside the field of view of the light collecting optic. If this is achieved, it should be possible to derive the electron density profile from measurements of either the extraordinary mode third harmonic or the ordinary mode second harmonic. An analysis of ECE spectra that have been measured before and after installation of the reflector is presented

Construction of the facility for the testing of the TFTR Neutral Beam Injector

International Nuclear Information System (INIS)

Haughian, J.; Lou, K.; Roth, D.

1979-11-01

The prototype for the TFTR Neutral Beam Injection System has been assembled at the Lawrence Berkeley Laboraory, and is presently under test. Some of the construction features of the shielding enclosure, the cryogenic supply system, control and computer area, and the auxiliary vacuum and utility supply system are described. In addition, the paper describes the target chamber, its beam dump and cryopanels, and the duct that connects the target chamber to the injector vessel

Mechanical engineering aspects of TFTR

International Nuclear Information System (INIS)

Citrolo, J.C.

1983-04-01

This paper briefly presents the principles which characterize a tokamak and discusses the mechanical aspects of TFTR, particularly the toroidal field coils and the vacuum chamber, in the context of being key components common to all tokamaks. The mechanical loads on these items as well as other design requirements are considered and the solutions to these requirements as executed in TFTR are presented. Future technological developments beyond the scope of TFTR, which are necessary to bring the tokamak concept to a full fusion-power system, are also presented. Additional methods of plasma heating, current drive, and first wall designs are examples of items in this category

TFTR grounding scheme and ground-monitor system

International Nuclear Information System (INIS)

Viola, M.

1983-01-01

The Tokamak Fusion Test Reactor (TFTR) grounding system utilizes a single-point ground. It is located directly under the machine, at the basement floor level, and is tied to the building perimeter ground. Wired to this single-point ground, via individual 500 MCM insulated cables, are: the vacuum vessel; four toroidal field coil cases/inner support structure quadrants; umbrella structure halves; the substructure ring girder; radial beams and columns; and the diagnostic systems. Prior to the first machine operation, a ground-loop removal program was initiated. It required insulation of all hangers and supports (within a 35-foot radius of the center of the machine) of the various piping, conduits, cable trays, and ventilation systems. A special ground-monitor system was designed and installed. It actively monitors each of the individual machine grounds to insure that there are no inadvertent ground loops within the machine structure or its ground and that the machine grounds are intact prior to each pulse. The TFTR grounding system has proven to be a very manageable system and one that is easy to maintain

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Design 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.)

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

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

Demonstrating diamond wire cutting of the TFTR

International Nuclear Information System (INIS)

Rule, K.; Perry, E.; Larson, S.; Viola, M.

2000-01-01

The Tokamak Fusion Test Reactor (TFTR) ceased operation in April 1997 and decommissioning commenced in October 1999. The deuterium-tritium fusion experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak (100 cubic meters) present a unique and challenging task for dismantling. Plasma arc cutting is the current baseline technology for the dismantlement of fission reactors. This technology is typically used because of its faster cutting times. Alternatively, an innovative approach for dismantlement of the TFTR is the use of diamond wire cutting technology. Recent improvements in diamond wire technology have allowed the cutting of carbon steel components such as pipe, plate, and tube bundles in heat exchangers. Some expected benefits of this technology include: significantly reduction in airborne contaminates, reduced personnel exposure, a reduced risk of spread of tritium contamination, and reduced overall costs as compared to using plasma arc cutting. This paper will provide detailed results of the diamond wire cutting demonstration that was completed in September of 1999, on a mock-up of this complex reactor. The results will identify cost, safety, industrial and engineering parameters, and the related performance of each situation

Demonstrating diamond wire cutting of the TFTR

Energy Technology Data Exchange (ETDEWEB)

Rule, K.; Perry, E.; Larson, S.; Viola, M. [and others

2000-02-24

The Tokamak Fusion Test Reactor (TFTR) ceased operation in April 1997 and decommissioning commenced in October 1999. The deuterium-tritium fusion experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak (100 cubic meters) present a unique and challenging task for dismantling. Plasma arc cutting is the current baseline technology for the dismantlement of fission reactors. This technology is typically used because of its faster cutting times. Alternatively, an innovative approach for dismantlement of the TFTR is the use of diamond wire cutting technology. Recent improvements in diamond wire technology have allowed the cutting of carbon steel components such as pipe, plate, and tube bundles in heat exchangers. Some expected benefits of this technology include: significantly reduction in airborne contaminates, reduced personnel exposure, a reduced risk of spread of tritium contamination, and reduced overall costs as compared to using plasma arc cutting. This paper will provide detailed results of the diamond wire cutting demonstration that was completed in September of 1999, on a mock-up of this complex reactor. The results will identify cost, safety, industrial and engineering parameters, and the related performance of each situation.

Tritium pellet injector for TFTR

International Nuclear Information System (INIS)

Gouge, M.J.; Baylor, L.R.; Cole, M.J.; Combs, S.K.; Dyer, G.R.; Fehling, D.T.; Fisher, P.W.; Foust, C.R.; Langley, R.A.; Milora, S.L.; Qualls, A.L.; Wilgen, J.B.; Schmidt, G.L.; Barnes, G.W.; Persing, R.G.

1992-01-01

The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) phase. The existing TFTR deuterium pellet injector (DPI) has been modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller. The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed, and the TPI was tested at ORNL with deuterium pellet. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

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.

Plasma-wall interaction: Recent TFTR results and implications on design and construction of limiters

International Nuclear Information System (INIS)

Owens, D.K.; Ulrickson, M.A.

1987-01-01

The first wall of the Tokamak Fusion Test Reactor (TFTR) consists of a water cooled toroidal belt limiter, a cooled moveable limiter, and cooled protective plates to shield the vacuum vessel from neutral beam shinethrough. Each of these systems consists of Inconel support plates covered with graphite tiles. In addition, there are Inconel and stainless steel bellows cover plates to protect the bellows and the surface pumping system which provides enhanced pumping in the torus and also serves to protect the bellows. These systems are described and the design requirements, simulations and actual thermal and mechanical loads reviewed. The normal and off-normal operating conditions which were considered in the design of the TFTR components include thermal loading during normal and disruptive plasma operation, eddy-current induced mechanical forces and arcing. The failures which have occurred are generally associated with thermal stress rather than mechanical failure due to disruption induced eddy currents. The models which were developed to design the TFTR hardware appear to have worked well as the performance of these systems has generally been satisfactory at loads approaching design limits. The implications of the TFTR experience for reactor design are discussed

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

Studies of tritiated co-deposited layers in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.A.; Ascione, G.; Causey, R.A.; Hayaski, T.; Hogan, J.; Nishi, M.; Shu, W.M.; Wampler, William R.; Young, K.M.

2000-01-01

Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.5 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition

Studies of tritiated co-deposited Layers in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.A.; Ascione, G.; Carpe, A.; Causey, R.A.; Hayashi, T.; Hogan, J.; Langish, S.W.; Nishi, M.F.; Shu, W.M.; Wampler, W.R.; Young, K.M.

2000-01-01

Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons, a stainless steel shutter and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.56 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling

Studies of tritiated co-deposited layers in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.A.; Ascione, G.; Carpe, A.; Causey, R.A.; Hayashi, T.; Hogan, J.; Langish, S.; Nishi, M.; Shu, W.M.; Wampler, W.R.; Young, K.M.

2000-01-01

Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons, a stainless steel shutter and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.56 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling

TFTR movable limiter instrumentation and controls

International Nuclear Information System (INIS)

Frankenberg, J.; Collins, D.; Kaufmann, D.; Mamoun, A.

1983-01-01

The TFTR movable limiter is a single poloidal limiter located within one 18 /SUP o/ segment of the vacuum vessel. It consists of three (3) interconnected inconel backing plates covered with titanium carbide coated graphite tiles. The backing plates are positioned by three independent screw drive actuators. Cooling water is fed through the horizontal port cover to tubes brazed onto the backs of the backing plates. Thermocouples monitor the limiter temperature. (1) and more fully described in refs. (1) and (2). The positioning actuators are driven by independently controlled DC servo motors, controlled either locally or from CICADA. Drive motor shaft position is monitored by chain driven encoders and potentiometers. Limiter blade position can be varied to suit any plasma within the operating range. CICADA is programmed to keep the limiter stroke within safe operating limits. A microprocessor duplicates the CICADA protective function allowing limiter operation without CICADA. The potentiometer signal is sent to an analog computer, which safeguards the limiter against failure of the encoders or the micro-processor. Cooling water flows through the limiter in 3 separate paths, one for each blade. The flow rate and temperature rise through each loop are measured accurately to allow CICADA to calculate the heat into each blade. The water system is also interlocked and alarmed to prevent dumping of water into the vacuum vessel

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.).

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.)

Conceptual thermal-mechanical design of the TFTR first wall armor against neutral beam impingement

International Nuclear Information System (INIS)

Chi, J.W.H.; Flaherty, R.

1976-01-01

The Tokamak Fusion Test Reactor (TFTR) is designed to operate in a pulsed mode with relatively low duty cycles. Each pulse consists of a short plasma heat-up period, a reaction period, followed by a relatively long cooldown period. Plasma heating is accomplished by ohmic heating by a current induced change in the magnetically linked ohmic heating coils, followed by neutral beam injection for further preheat and the initiation of fusion reactions. During normal operation, the bulk of the neutral beam energy will be absorbed by the plasma, while the remainder will impinge on the vacuum vessel wall. The amount of thermal energy deposited on an unprotected wall is expected to be excessive, limiting the frequency of pulses and requiring frequent wall replacement. A faulted condition would cause penetration of an unprotected wall. As a consequence, a wall armoring (or liner) concept was developed to protect the vacuum vessel wall and to permit ease of liner replacement

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

ITER cryostat main chamber and vacuum vessel pressure suppression system design

International Nuclear Information System (INIS)

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

1999-03-01

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

ITER cryostat main chamber and vacuum vessel pressure suppression system design

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

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.

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.)

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.

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

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

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.)

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.

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...

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.

Possible neutral beam requirements for TFTR upgrades

International Nuclear Information System (INIS)

Prichard, B.A. Jr.; Little, R.; Post, D.E.; Schmidt, J.A.

1977-01-01

A discussion is provided of possible neutral beam requirements and constraints for a TFTR upgrade. The time scale is the early 80s and beams of 250 keV D 0 , probably using 65 ampere negative ion sources, existing power supplies and vacuum enclosures would be required

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)

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.)

Discharge cleaning on TFTR after boronization

International Nuclear Information System (INIS)

Mueller, D.; Dylla, H.F.; LaMarche, P.H.; Bell, M.G.; Blanchard, W.; Bush, C.E.; Gentile, C.; Hawryluk, R.J.; HIll, K.W.; Janos, A.C.; Jobes, F.C; Owens, D.K.; Pearson, G.; Schivell, J.; Ulrickson, M.A.; Vannoy, C.; Wong, K.L.

1991-05-01

At the beginning of the 1990 TFTR experimental run, after replacement of POCO-AXF-5Q graphite tiles on the midplane of the bumper limiter by carbon fiber composite (CFC) tiles and prior to any Pulse Discharge Cleaning (PDC), boronization was performed. Boronization is the deposition of a layer of boron and carbon on the vacuum vessel inner surface by a glow discharge in a diborane, methane and helium mixture. The amount of discharge cleaning required after boronization was substantially reduced compared to that which was needed after previous openings when boronization was not done. Previously, after a major shutdown, about 10 5 low current (∼20 kA) Taylor Discharge Cleaning (TDC) pulses were required before high current (∼400 kA) aggressive Pulse Discharge Cleaning (PDC) pulses could be performed successfully. Aggressive PDC is used to heat the limiters from the vessel bakeout temperature of 150 degrees C to 250 degrees C for a period of several hours. Heating the limiters is important to increase the rate at which water is removed from the carbon limiter tiles. After boronization, the number of required TDC pulses was reduced to <5000. The number of aggressive PDC pulses required was approximately unchanged. 14 refs., 1 tab

Periscope-camera system for visible and infrared imaging diagnostics on TFTR

International Nuclear Information System (INIS)

Medley, S.S.; Dimock, D.L.; Hayes, S.; Long, D.; Lowrence, J.L.; Mastrocola, V.; Renda, G.; Ulrickson, M.; Young, K.M.

1985-05-01

An optical diagnostic consisting of a periscope which relays images of the torus interior to an array of cameras is used on the Tokamak Fusion Test Reactor (TFTR) to view plasma discharge phenomena and inspect vacuum vessel internal structures in both visible and near-infrared wavelength regions. Three periscopes view through 20-cm-diameter fused-silica windows which are spaced around the torus midplane to provide a viewing coverage of approximately 75% of the vacuum vessel internal surface area. The periscopes have f/8 optics and motor-driven controls for focusing, magnification selection (5 0 , 20 0 , and 60 0 field of view), elevation and azimuth setting, mast rotation, filter selection, iris aperture, and viewing port selection. The four viewing ports on each periscope are equipped with multiple imaging devices which include: (1) an inspection eyepiece, (2) standard (RCA TC2900) and fast (RETICON) framing rate television cameras, (3) a PtSi CCD infrared imaging camera, (4) a 35 mm Nikon F3 still camera, or (5) a 16 mm Locam II movie camera with variable framing up to 500 fps. Operation of the periscope-camera system is controlled either locally or remotely through a computer-CAMAC interface. A description of the equipment and examples of its application are presented

Periscope-camera system for visible and infrared imaging diagnostics on TFTR

International Nuclear Information System (INIS)

Medley, S.S.; Dimock, D.L.; Hayes, S.; Long, D.; Lowrance, J.L.; Mastrocola, V.; Renda, G.; Ulrickson, M.; Young, K.M.

1985-01-01

An optical diagnostic consisting of a periscope which relays images of the torus interior to an array of cameras is used on the Tokamak Fusion Test Reactor (TFTR) to view plasma discharge phenomena and inspect the vacuum vessel internal structures in both the visible and near-infrared wavelength regions. Three periscopes view through 20-cm-diam fused-silica windows which are spaced around the torus midplane to provide a viewing coverage of approximately 75% of the vacuum vessel internal surface area. The periscopes have f/8 optics and motor-driven controls for focusing, magnification selection (5 0 , 20 0 , and 60 0 field of view), elevation and azimuth setting, mast rotation, filter selection, iris aperture, and viewing port selection. The four viewing ports on each periscope are equipped with multiple imaging devices which include: (1) an inspection eyepiece, (2) standard (RCA TC2900) and fast (RETICON) framing rate television cameras, (3) a PtSi CCD infrared imaging camera, (4) a 35-mm Nikon F3 still camera, or (5) a 16-mm Locam II movie camera with variable framing rate up to 500 fps. Operation of the periscope-camera system is controlled either locally or remotely through a computer-CAMAC interface. A description of the equipment and examples of its application are presented

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.)

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

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.

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

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.

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

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.

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.

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...

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.

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

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)

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)

Plasma-material interactions in TFTR

International Nuclear Information System (INIS)

Dylla, H.F.; Bell, M.G.; Blanchard, W.R.; Boody, F.P.; Bretz, N.; Budny, R.; Bush, C.E.; Cecchi, J.L.; Cohen, S.A.; Combs, S.K.; Davis, S.L.; Doyle, B.L.; Efthimion, P.C.; England, A.C.; Eubank, H.P.; Fonck, R.; Fredrickson, E.; Grisham, L.R.; Goldston, R.J.; Grek, B.; Groebner, R.; Hawryluk, R.J.; Heifetz, D.; Hendel, H.; Hill, K.W.; Hiroe, S.; Hulse, R.; Johnson, D.; Johnson, L.C.; Kilpatrick, S.; Lamarche, P.H.; Little, R.; Manos, D.M.; Mansfield, D.; Meade, D.M.; Medley, S.S.; Milora, S.L.; Mikkelsen, D.R.; Mueller, D.; Murakami, M.; Nieschmidt, E.; Owens, D.K.; Park, H.; Pontau, A.; Prichard, B.; Ramsey, A.T.; Redi, M.H.; Schivell, J.; Schmidt, G.L.; Scott, S.D.; Sesnic, S.; Shimada, M.; Simpkins, J.E.; Sinnis, J.; Stauffer, F.; Stratton, B.; Tait, G.D.; Taylor, G.; Ulrickson, M.; Von Goeler, S.; Wampler, W.R.; Wilson, K.; Williams, M.; Wong, K.L.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.

1987-01-01

This paper presents a summary of plasma-material interactions which influence the operation of TFTR with high current (≤ 2.2 MA) ohmically heated, and high-power (≅ 10 MW) neutral-beam heated plasmas. The conditioning procedures which are applied routinely to the first-wall hardware are reviewed. Fueling characteristics during gas, pellet, and neutral-beam fueling are described. Recycling coefficients near unity are observed for most gas fueled discharges. Gas fueled discharges after helium discharge conditioning of the toroidal bumper limiter, and discharges fueled by neutral beams and pellets, show R e = 5-6x10 19 m -3 ) values of Z eff are ≤ 1.5. Increases in Z eff of ≤ 1 have been observed with neutral beam heating of 10 MW. The primary low Z impurity is carbon with concentrations decreasing from ≅ 10% to e . Oxygen densities tend to increase with n e , and at the ohmic plasma density limit oxygen and carbon concentrations are comparable. Chromium getter experiments and He 2+ /D + plasma comparisons indicate that the limiter is the primary source of carbon and that the vessel wall is a significant source of the oxygen impurity. Metallic impurities, consisting of the vacuum vessel metals (Ni, Fe, Cr) have significant (≅ 10 -4 n e ) concentrations only at low plasma densities (n e 19 m -3 ). The primary source of metallic impurities is most likely ion sputtering from metals deposited on the carbon limiter surface. (orig.)

Alignment of the TFTR bumper limiter

International Nuclear Information System (INIS)

Barnes, G.W.; Owens, D.K.; Loesser, G.D.; Ulrickson, M.

1989-01-01

The TFTR Bumper Limiter (BL) is an axisymmetric toroidal limiter mounted on the inner wall of the vacuum vessel. It subtends 120 degree poloidally and has a surface area of 22 m 2 . The plasma facing surface consists of 1,000 kg of graphite tiles mounted on watercooled Inconel backing plates. During the initial installation in the Spring of 1985, the limiter surface was aligned to the toroidal magnetic field by mechanical and magnetic measurements to an estimated accuracy of ±2 mm. During subsequent operation, especially in the 1988 run period in which 30 MW of Neutral Beam Injection routinely occurred, several tiles at points on the limiter which protruded slightly into the plasma were severely damaged. The damage, cracked and spalled tiles, is believed to be initiated by high energy disruptions and aggravated by normal high power operation. The damage pattern and temperature rise during normal operation are consistent with this interpretation. A vacuum vessel opening to replace the damaged tiles and realign the limiter was required. The bumper limiter was reshaped to be circular to ±0.5 mm at the midplane by means of mechanical measurements in order to better distribute the heat loads and eliminate hot spots. The ±0.5 mm accuracy is determined by the variation in individual tile thickness which is ±0.5 mm. This paper describes the methods used to mechanically align the limiter and presents evidence based on machine operation with plasma that the limiter is reasonably well aligned with the toroidal field. Future work dealing with the alignment of the total limiter to the toroidal field using mechanical and magnetic measurements and the replacement of a subset of the carbon tiles with carbon-carbon composite material is also discussed. 7 refs., 4 figs

Progress of ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Progress of ITER vacuum vessel

International Nuclear Information System (INIS)

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

2013-01-01

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

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.)

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

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.)

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.)

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

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

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.

In-situ Tritium Measurements of the Tokamak Fusion Test Reactor Bumper Limiter Tiles Post D-T Operations

International Nuclear Information System (INIS)

C.A. Gentile; C.H. Skinner; K.M. Young; M. Nishi; S. Langish; et al

1999-01-01

The Princeton Plasma Physics Laboratory (PPPL) Engineering and Research Staff in collaboration with members of the Japan Atomic Energy Research Institute (JAERI), Tritium Engineering Laboratory have commenced in-situ tritium measurements of the TFTR bumper limiter. The Tokamak Fusion Test Reactor (TFTR) operated with tritium from 1993 to 1997. During this time ∼ 53,000 Ci of tritium was injected into the TFTR vacuum vessel. After the cessation of TFTR plasma operations in April 1997 an aggressive tritium cleanup campaign lasting ∼ 3 months was initiated. The TFTR vacuum vessel was subjected to a regimen of glow discharge cleaning (GDC) and dry nitrogen and ''moist air'' purges. Currently ∼ 7,500 Ci of tritium remains in the vacuum vessel largely contained in the limiter tiles. The TFTR limiter is composed of 1,920 carbon tiles with an average weight of ∼ 600 grams each. The location and distribution of tritium on the TFTR carbon tiles are of considerable interest. Future magnetically confined fusion devices employing carbon as a limiter material may be considerably constrained due to potentially large tritium inventories being tenaciously held on the surface of the tiles. In-situ tritium measurements were conducted in TFTR bay L during August and November 1998. During the bay L measurement campaign open wall ion chambers and ultra thin thermoluminscent dosimeters (TLD) affixed to a boom and end effector were deployed into the vacuum vessel. The detectors were designed to make contact with the surface of the bumper limiter tile and to provide either real time (ion chamber) or passive (TLD) indication of the surface tritium concentration. The open wall ion chambers were positioned onto the surface of the tile in a manner which employed the surface of the tile as one of the walls of the chamber. The ion chambers, which are (electrically) gamma insensitive, were landed at four positions per tile. The geometry for landing the TLD's provided measurement at 24

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.)

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

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.

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.

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

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)

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

Hardware design of a microcomputer controlled diagnostic vacuum controller

International Nuclear Information System (INIS)

Marsala, R.J.

1983-01-01

The TFTR diagnostic vacuum controller (DVC) has been designed and built to control and monitor the pumps, valves and gauges which comprise a diagnostic vacuum system. The DVC is a microcomputer based self-contained controller with battery backup which may be controlled manually from front panel controls or remotely via CICADA. The DVC implements all pump and valve sequencing and provides protection against incorrect operation. There are presently two versions of the DVC operating on TFTR and a third version being used on the S-1 machine

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)

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.

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.

Escaping 1 MeV tritons in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Strachan, J.D.; Boivin, R.; Cavallo, A.; Fredrickson, E.D.; McGuire, K.; Mynick, H.E.; White, R.B.

1989-01-01

1 MeV tritons created by D-D reactions can simulate the 'single-particle' behavior expected with 3.5 MeV D-T alphas, since the gyroradii and slowing-down of these two particles are similar. This paper describes measurements of the flux of escaping 1 MeV tritons from the TFTR plasma during high power D 0 →D neutral beam injection, and shows that in most cases the observed triton loss is consistent with the classical (single-particle) first-orbit loss model. In this model tritons are lost if their first orbit intersects the wall due to their large banana width, while almost all tritons confined on their first orbit should stay confined until thermalized. The triton detectors are ZnS(Ag) scintillator screens housed in light-tight boxes located just outside the plasma boundary at the bottom of the TFTR vessel. They are particle 'pinhole' cameras which can resolve the triton flux vs. pitch angle (to ±5 o ), energy (to ±50 %), and time (to <20 μsec). The 2-D images of triton flux onto these scintillators are optically coupled to either an intensified TV camera or to photomultiplyer tubes for fast time resolution. The soft x-ray background in an earlier prototype has been eliminated. Although there are presently 8 such detectors in TFTR, this paper discusses results from only the detector located just below the vessel center (R=259 cm, r=102 cm). Note that the '1 MeV triton' signal discussed below also has about a 30 % contribution from 3 MeV protons; however, since these two particles have identical gyroradii they should behave alike. 5 refs., 5 figs

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

Operation of the repeating pneumatic injector on TFTR and design of an 8-shot deuterium pellet injector

International Nuclear Information System (INIS)

Combs, S.K.; Milora, S.L.; Foust, C.R.

1985-01-01

The repeating pneumatic hydrogen pellet injector, which was developed at the Oak Ridge National Laboratory (ORNL), has been installed and operated on the Tokamak Fusion Test Reactor (TFTR). The injector combines high-speed extruder and pneumatic acceleration technologies to propel frozen hydrogen isotope pellets repetitively at high speeds. The pellets are transported to the plasma in an injection line that also serves to minimize the gas loading on the torus; the injection line incorporates a fast shutter valve and two stages of guide tubes with intermediate vacuum pumping stations. A remote, stand-alone control and data acquisition system is used for injector and vacuum system operation. In early pellet fueling experiments on TFTR, the injector has been used to deliver deuterium pellets at speeds ranging from 1.0 to 1.5 km/s into plasma discharges. First, single large (nominal 4-mm-dia) pellets provided high densities in TFTR (1.8 x 10 14 cm -3 on axis); after conversion to smaller (nominal 2.7-mm-dia) pellets, up to five pellets were injected at 0.25-s intervals into a plasma discharge, giving a line-averaged density of 1 x 10 14 cm -3 . Operating characteristics and performance of the injector in initial tests on TFTR are presented

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

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.

Tritium processing and management during D-T experiments on TFTR

International Nuclear Information System (INIS)

La Marche, P.H.; Anderson, J.L.; Gentile, C.A.; Hawryluk, R.J.; Hosea, J.; Kalish, M.; Kozub, T.; Murray, H.; Nagy, A.; Raftopoulos, S.

1994-11-01

TFTR performance has surpassed many of the previous tokamak records. This has been made possible by the use of tritium as fuel for DT plasma discharges. Stable operations of tritium systems provide for safe, routine DT operation of TFTR. In the preparation for DT operation, in the commissioning of the tritium systems and in the operation of the Nuclear Facility several key lessons have been learned. They include: the facility must take the lead in interpreting the applicable regulations and orders and then seek regulator approval; the use of ultra high vacuum technology in tritium system design and construction simplifies and enhances operations and maintenance; and central facility control under a single supervisory position is crucial to safely orchestrate operational and maintenance activities

Remote leak detection for the TFTR

International Nuclear Information System (INIS)

Walthers, C.R.

1977-01-01

The planned design for the TFTR (TOKAMAK Fusion Test Reactor) remote leak detection system consists of a central console which controls the application of tracer gas to possible leak areas. Seals are tested by admitting tracer gas to machined cavities on the atmospheric side of the seal. The tracer gas is brought to the seal cavity by 1 / 8 -inch diameter tubes which connect to local tracer gas/vacuum manifolds located outside the protective radiation shielding. Vacuum shell walls and welds are checked by flowing tracer gas through annular heating/cooling passages. The detector will be either an MSLD (mass spectrometer leak detector) or an RGA (residual gas analyzer), the location of which is not finalized. Feasibility tests performed and planned include response and sensitivity measurements of possible tubing/detector configurations with several tracer gases

Fokker-Planck Modelling of Delayed Loss of Charged Fusion Products in TFTR

International Nuclear Information System (INIS)

Edenstrasser, J.W.; Goloborod'ko, V.Ya.; Reznik, S.N.; Yavorskij, V.A.; Zweben, S.

1998-01-01

The results of a Fokker-Planck simulation of the ripple-induced loss of charged fusion products in the Tokamak Fusion Test Reactor (TFTR) are presented. It is shown that the main features of the measured ''delayed loss'' of partially thermalized fusion products, such as the differences between deuterium-deuterium and deuterium-tritium discharges, the plasma current and major radius dependencies, etc., are in satisfactory agreement with the classical collisional ripple transport mechanism. The inclusion of the inward shift of the vacuum flux surfaces turns out to be necessary for an adequate and consistent explanation of the origin of the partially thermalized fusion product loss to the bottom of TFTR

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

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

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.

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.

Tokamak Fusion Test Reactor neutral beam injection system vacuum chamber

International Nuclear Information System (INIS)

Pedrotti, L.R.

1977-01-01

Most of the components of the Neutral Beam Lines of the Tokamak Fusion Test Reactor (TFTR) will be enclosed in a 50 cubic meter box-shaped vacuum chamber. The chamber will have a number of unorthodox features to accomodate both neutral beam and TFTR requirements. The design constraints, and the resulting chamber design, are presented

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.)

Tritium Decontamination of TFTR D-T Graphite Tiles Employing Ultra Violet Light and a Nd:YAG Laser

International Nuclear Information System (INIS)

Gentile, C.A.; Skinner, C.H.; Young, K.M.; Ciebiera, L.

1999-01-01

The use of an ultra violet (UV) light source (wavelength = 172 nm) and a Nd:YAG Laser for the decontamination of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles will be investigated at the Princeton Plasma Physics Laboratory (PPPL). The development of this form of tritium decontamination may be useful for future D-T burning fusion devices which employ carbon plasma-facing components on the first wall. Carbon tiles retain hydrogen isotopes, and the in-situ tritium decontamination of carbon can be extremely important in maintaining resident in-vessel tritium inventory to a minimum. A test chamber has been designed and fabricated at PPPL. The chamber has the ability to be maintained under vacuum, be baked to 200 *C, and provides sample ports for gas analyses. Tiles from TFTR that have been exposed to D-T plasmas will be placed within the chamber and exposed to either an UV light source or the ND:YAG Laser. The experiment will determine the effectiveness of these two techniques for the removal of tritium. In addition, exposure rates and scan times for the UV light source and/or Nd:YAG Laser will be determined for tritium removal optimization from D-T tiles

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

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)

ITER vacuum vessel structural analysis completion during manufacturing phase

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

ITER vacuum vessel structural analysis completion during manufacturing phase

International Nuclear Information System (INIS)

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

2016-01-01

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

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)

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.

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.)

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.

Demonstration tests for manufacturing the ITER vacuum vessel

International Nuclear Information System (INIS)

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

2007-01-01

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

Demonstration tests for manufacturing the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

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

2007-10-15

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

A tritium vessel cleanup experiment in TFTR

International Nuclear Information System (INIS)

Caorlin, M.; Kamperschroer, J.; Owens, D.K.; Voorhees, D.; Mueller, D.; Ramsey, A.T.; La Marche, P.H.; Loughlin, M.J.

1995-03-01

A simple tritium cleanup experiment was carried out in TFTR following the initial high power deuterium-tritium discharges in December 1993. A series of 34 ohmic and deuterium neutral beam fueled shots was used to study the removal of tritium implanted into the wall and limiters. A very large plasma was created in each discharge to ''scrub'' an area as large as possible. Beam-fueled shots at 2.5 to 7.5 MW of injected power were used to monitor tritium concentration levels in the plasma by detection of DT-neutrons. The neutron signal decreased by a factor of 4 during the experiment, remaining well above the expected T-burnup level. The amount of tritium recovered at the end of the cleanup was about 8% of the amount previously injected with high power DT discharges. The experience gained suggests that measurements of tritium inventory in the torus are very difficult to execute and require dedicated systems with overall accuracy of 1%

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

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.

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.

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)

Status report on TFTR

International Nuclear Information System (INIS)

Reardon, P.J.

1978-01-01

The primary objectives of the TFTR are the generation and confinement of 5 to 10 keV (50 to 100 million degrees) reactor-grade plasmas in a tokamad magnetic-field configuration, and the production of fusion energy on a pulsed basis, from the reaction of deuterum and tritium. The TFTR will be used to study the physics of burning plasmas and the engineering aspects of a D-T burning tokamak operating with reactor-level plasma conditions. The overall TFTR program is intended to produce scientific and technical information, component hardware, and the design, construction, and operating experience necessary as input for the future design, construction, and operation of ignition and experimental fusion power reactors. In a very real way the TFTR is prototypical of an Experimenta Power Reactor

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

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)

Development of the TFTR neutral beam injection system

International Nuclear Information System (INIS)

Prichard, B.A. Jr.

1977-01-01

The TFTR Neutral Beam Lines are designed to inject 20 MW of 120 keV neutral deuterium atoms into the plasma. This is accomplished using 12 sources, 65 amperes each, mounted in 4 beam lines. The 120 kV sources and a prototype beam line are being developed. The implementation of these beam lines has required the development of several associated pieces of hardware. 200 kV switch tubes for the power supplies are being developed for modulation and regulation of the accelerating supplies. A 90 cm metallic seal gate valve capable of sealing against atmosphere in either direction is being developed for separating the torus and beam line vacuum systems. A 70 x 80 cm fast shutter valve is also being developed to limit tritium migration from the torus into the beam line. Internal to the beam line a calorimeter, ion dump and deflection magnet have been designed to handle three beams, and optical diagnostics utilizing the doppler broadening and doppler shift of light emitted from the accelerated beam are being developed. The control and monitoring of the 12 sources will be done via the TFTR computer control system (CICADA) as will other parts of the machine, and software is being developed to condition and operate the sources automatically. The prototype beam line is scheduled to begin operation in the fall of 1978 and all four production beam lines on TFTR in 1982

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

Vacuum vessel port structures for ITER-FEAT

International Nuclear Information System (INIS)

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

2001-01-01

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

Vacuum vessel port structures for ITER-FEAT

Energy Technology Data Exchange (ETDEWEB)

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

2001-11-01

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

Charged fusion product and fast ion loss in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Darrow, D.S.; Fredrickson, E.D.; Mynick, H.E.; White, R.B.; Biglari, H.; Bretz, N.; Budny, R.; Bush, C.E.; Chang, C.S.; Chen, L.; Cheng, C.Z.; Fu, G.Y.; Hammett, G.W.; Hawryluk, R.J.; Hosea, J.; Johnson, L.; Mansfield, D.; McGuire, K.; Medley, S.S.; Nazikian, R.; Owens, D.K.; Park, H.; Park, J.; Phillips, C.K.; Schivell, J.; Stratton, B.C.; Ulrickson, M.; Wilson, R.; Young, K.M.; Fisher, R.; McChesney, J.; Fonck, R.; McKee, G.; Tuszewski, M.

1993-03-01

Several different fusion product and fast ion loss processes have been observed in TFTR using an array of pitch angle, energy and time resolved scintillator detectors located near the vessel wall. For D-D fusion products (3 MeV protons and 1 MeV tritons) the observed loss is generally consistent with expected first-orbit loss for Ip I MA. However, at higher currents, Ip = 1.4--2.5 MA, an NM induced D-D fusion product loss can be up to 3-4 times larger than the first-orbit loss, particularly at high beam powers, P ≥ 25 MW. The MHD induced loss of 100 KeV neutron beam ions and ∼0.5 MeV ICRF minority tail tons has also been measured ≤ 459 below the outer midplane. be potential implications of these results for D-T alpha particle experiments in TFTR and ITER are described

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...

Control of TFTR during DT operations

International Nuclear Information System (INIS)

Pearson, G.G.; Alling, P.D.; Blanchard, W.; Camp, R.A.; Hawryluk, R.J.; Hosea, J.C.; Nagy, A.

1995-01-01

Since beginning routine D-T operations in December, 1993, there have been more than 500 DT plasmas and approximately 600,000 Ci of tritium processed through TFTR culminating in greater than 10 MW of fusion power produced in a single discharge in November, 1994. These performance levels were achieved while maintaining the highest levels of personnel and equipment safety. Prior to D-T operations, a Chain of Command structure and a TFTR Shift Supervisor (TFTRSS) position were developed for centralized control of the facility with all subsystems reporting to this position. A comprehensive surveillance system was incorporated such that the TFTR SS could easily review the operational readiness of subsystems for D-T operations. A TFTR SS Station was constructed to facilitate monitoring and control of TFTR. This station includes a camera system, FAX, a networked personal computer, a computerized tritium monitor and control system and a hardware interlock system. In the transition from D-D to D-T operations, TFTR's procedures were reviewed/revised and a number of additional procedures developed for control of activities at the facility. This paper details the equipment, administrative and organizational configurations used for controlling TFTR during D-T operations

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.

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.)

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.

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.

Mechanical design of the folded waveguide for PBX-M and TFTR

International Nuclear Information System (INIS)

Fogelman, C.H.; Bigelow, T.S.; Yugo, J.J.

1995-01-01

The folded waveguide (FWG) antenna is an advanced Cyclotron Range of Frequencies launcher being designed at Oak Ridge National Laboratory in collaboration with Princeton Plasma Physics Laboratory. The FWG offers a drastic increase in radio frequency (RF) power density over typical loop antennas. It also results in internal electric fields of much lower magnitude near the plasma. It is scheduled for installation on either the Tokamak Fusion Test Reactor (TFTR) or the Princeton Beta Experiment-Modified (PBX-M) tokamak in January 1996. The design objective is to provide an FWG that can withstand the thermal loads and disruption scenarios and meet the space constants of both machines. The design is also intended to be prototypical for the International Thermonuclear Experimental Reactor (ITER). The FWG is fully retractable, and maintenance operations can be performed while the vessel remains under vacuum. The FWG can operate in fast-wave mode, or it can be retracted, rotated 90 degrees, and reengaged for the ion-Bernstein wave launch. The polarizing plate completely covers the front of the antenna, except for slots cut at every other gap between with plates of other configurations such as a 0-π dipole plate

Phenomenology of high density disruptions in the TFTR tokamak

International Nuclear Information System (INIS)

Fredrickson, E.D.; McGuire, K.M.; Bell, M.G.

1993-01-01

Studies of high density disruptions on TFTR, including a comparison of minor and major disruptions at high density, provide important new information regarding the nature of the disruption mechanism. Further, for the first time, an (m,n)=(1,1) 'cold bubble' precursor to high density disruptions has been experimentally observed in the electron temperature profile. The precursor to major disruptions resembles the 'vacuum bubble' model of disruptions first proposed by B.B. Kadomtsev and O.P. Pogutse (Sov. Phys. - JETP 38 (1974) 283). (author). Letter-to-the-editor. 25 refs, 3 figs

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)

Coil protection calculator for TFTR

International Nuclear Information System (INIS)

Marsala, R.J.; Woolley, R.D.

1987-01-01

A new coil protection calculator (CPC) is presented in this paper. It is now being developed for TFTR's magnetic field coils will replace the existing coil fault detector. The existing fault detector sacrifices TFTR operating capability for simplicity. The new CPC will permit operation up to the actual coil limits by accurately and continuously computing coil parameters in real-time. The improvement will allow TFTR to operate with higher plasma currents and will permit the optimization of pulse repetition rates

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.

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.

Diamond Wire Cutting of the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Keith Rule; Erik Perry; Robert Parsells

2003-01-01

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of DandD (Decontamination and Decommissioning) activity

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

High beta and confinement studies on TFTR

International Nuclear Information System (INIS)

1990-01-01

In the first 8 months of this project we have made substantial progress toward the goals set out in our original proposal. Our plan to access new regimes of operation at high values of var-epsilon β p using low current discharges in TFTR has worked extremely well and a new regime of operation has indeed been found in the course of our execution of TFTR Experimental Proposal 146 which involved our operation of TFTR on 9 November 1989, 19--20 January 1990 and 1--2 February 1990. The status of our high var-epsilon β p work on TFTR is given and is extracted from our paper submitted for presentation to the 1990 EPS meeting in Amsterdam. We have also performed an analysis of the energetic particle stabilization requirements for TFTR Supershots, and developed methods for analysis and a theory of perturbative transport measurements in TFTR

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

Alpha particle loss in the TFTR DT experiments

International Nuclear Information System (INIS)

Zweben, S.J.; Darrow, D.S.; Herrmann, H.W.

1995-01-01

Alpha particle loss was measured during the TFTR DT experiments using a scintillator detector located at the vessel bottom in the ion grad-B drift direction. The DT alpha particle loss to this detector was consistent with the calculated first-orbit loss over the whole range of plasma current I=0.6-2.7 MA. In particular, the alpha particle loss rate per DT neutron did not increase significantly with fusion power up to 10.7 MW, indicating the absence of any new ''collective'' alpha particle loss processes in these experiments

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

Synchronization of timing systems on TFTR

International Nuclear Information System (INIS)

Montague, J.; Sichta, P.

1992-01-01

This paper reports on the TOKAMAK Fusion Test Reactor (TFTR) facility clock system which has four related timing subsystems: the TFTR shot clock, the Neutral Beams clocks, the Ion Cyclotron Range of Frequencies (ICRF) system clock, and the Disruption Trigger System. These systems have been integrated to support increasingly fast sampling rates in data acquisition and greater accuracy in the firing of the Neutral Beams and ICRF systems during TFTR shots

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.

TFTR DT preparation project status

Energy Technology Data Exchange (ETDEWEB)

Perry, E.D.; Dudek, L.E.

1993-11-01

The Tokamak Fusion Test Reactor (TFTR) research program is preparing to commence the first high power Deuterium-Tritium (DT) experiments of the US Fusion Program. Hardware upgrades to TFTR required for DT operations have been completed. This paper discusses these hardware preparations.

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.

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)

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)

TFTR D and D Project: Final Examination and Testing of the TFTR TF-Coils

International Nuclear Information System (INIS)

Zatz, Irving J.

2003-01-01

In operation for nearly 15 years, TFTR (Tokamak Fusion Test Reactor) was not only a fusion science milestone, but a milestone of achievement in engineering as well. The TFTR DandD (Decommissioning and Decontamination) program provided a rare opportunity to examine machine components that had been exposed to a unique performance environment of greater than 100,000 mechanical and thermal load cycles. In particular, the possible examination of the TFTR toroidal-field (TF) coils, which met, then exceeded, the 5.2 Tesla magnetic field machine specification, could supply the answers to many questions that have been asked and debated since the coils were originally designed and built. A test program conducted in parallel with the DandD effort was the chance to look inside and examine, in detail, the TFTR TF coils for the first time since they were delivered encased to PPPL (Princeton Plasma Physics Laboratory). The results from such a program would provide data and insight that would not only be nefit PPPL and the fusion community, but the broader scientific community as well

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)

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

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

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

Handling of tritium at TFTR

International Nuclear Information System (INIS)

Pierce, C.W.; Howe, H.J.; Yemin, L.; Lind, K.

1977-01-01

Some of the engineering approaches taken at TFTR for the tritium control systems are discussed as the requirements being placed on the tritium systems by the operating scenarios of the Tokamak. The tritium control systems presently being designed for TFTR will limit the annual release to the environment to less than 100 curies

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

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)

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

The tritium operations experience on TFTR

International Nuclear Information System (INIS)

Halle, A. von; Anderson, J.L.; Gentile, C.; Grisham, L.; Hosea, J.; Kamperschroer, J.; LaMarche, P.; Oldaker, M.; Nagy, A.; Raftopoulos, S.; Stevenson, T.

1995-01-01

The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grams of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the U.S. Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described. (orig.)

The tritium operations experience on TFTR

International Nuclear Information System (INIS)

von Halle, A.; Gentile, C.

1994-01-01

The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grains of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes- the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the US Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described

The ICRF antennas for TFTR

International Nuclear Information System (INIS)

Hoffman, D.J.; Colestock, P.L.; Gardner, W.L.; Hosea, J.C.; Nagy, A.; Stevens, J.; Swain, D.W.; Wilson, J.R.

1988-01-01

Two compact loop antennas have been designed to provide ion cyclotron resonant frequency (ICRF) heating for TFTR. The antennas can convey a total of 10 MW to accomplish core heating in either high-density or high-temperature plasmas. The near-term goal of heating TFTR plasmas and the longer-term goals of ease in handling (for remote maintenance) and high reliability (in an inaccessible tritium tokamak environment) were major considerations in the antenna designs. The compact loop configuration facilitates handling because the antennas fit completely through their ports. Conservative design and extensive testing were used to attain the reliability required for TFTR. This paper summarizes how these antennas will accomplish these goals. 5 figs, 1 tab

Chromium getter studies in TFTR

International Nuclear Information System (INIS)

Dylla, H.F.; LaMarche, P.H.; Blanchard, W.R.

1986-02-01

We have studied the effects of the deposition of thin films (approx.0.1 μm) of chromium onto approx.70% of the torus area of the Tokamak Fusion Test Reactor (TFTR). The purpose of these experiments was to test the difference between high surface coverage and high pumping speed gettering schemes with respect to minimizing oxygen impurity generation in high power tokamak discharges. The initial Cr deposition had significant effects on vessel outgassing and subsequent plasma performance: the outgassing of H 2 O, CO, and CO 2 decreased by a factor of ten, oxygen impurity radiation decreased by a factor of two, the plasma Z/sub eff/ decreased from 1.3 to 1.1, and the plasma density limit increased by 20%. This improvement correlates with a significant reduction of the edge radiation as the density limit is approached. The effects of the initial and subsequent Cr depositions were relatively long lasting, exhibiting time constants of the order of weeks. We attribute the observed impurity reduction to a modification of the oxide surface on the vessel wall, which is apparently a significant impurity source for oxygen. 17 refs., 6 figs

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)

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

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.

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.

Tritium decontamination of TFTR carbon tiles employing ultra violet light

International Nuclear Information System (INIS)

Shu, W.M.; Ohira, S.; Gentile, C.A.; Oya, Y.; Nakamura, H.; Hayashi, T.; Iwai, Y.; Kawamura, Y.; Konishi, S.; Nishi, M.F.; Young, K.M.

2001-01-01

Tritium decontamination on the surface of Tokamak Fusion Test Reactor (TFTR) bumper limiter tiles used during the Deuterium-Deuterium (D-D) phase of TFTR operations was investigated employing an ultra violet light source with a mean wavelength of 172 nm and a maximum radiant intensity of 50 mW/cm 2 . The partial pressures of H 2 , HD, C and CO 2 during the UV exposure were enhanced more than twice, compared to the partial pressures before UV exposure. In comparison, the amount of O 2 decreased during the UV exposure and the production of a small amount of O 3 was observed when the UV light was turned on. Unlike the decontamination method of baking in air or oxygen, the UV exposure removed hydrogen isotopes from the tile to vacuum predominantly in forms of gases of hydrogen isotopes. The tritium surface contamination on the tile in the area exposed to the UV light was reduced after the UV exposure. The results show that the UV light with a wavelength of 172 nm can remove hydrogen isotopes from carbon-based tiles at the very surface

Disruptions in the TFTR tokamak

International Nuclear Information System (INIS)

Janos, A.; Fredrickson, E.D.; McGuire, K.; Batha, S.H.; Bell, M.G.; Bitter, M.; Budny, R.; Bush, C.E.; Efthimion, P.C.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Jobes, F.C.; Johnson, D.W.; Levinton, F.; Mansfield, D.; Meade, D.; Medley, S.S.; Monticello, D.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Park, H.; Park, W.; Post, D.E.; Schivell, J.; Strachan, J.D.; Taylor, G.; Ulrickson, M.; Goeler, S. von; Wilfrid, E.; Wong, K.L.; Yamada, M.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.; Drake, J.F.; Kleva, R.G.; Fleischmann, H.H.

1993-03-01

For a successful reactor, it will be useful to predict the occurrence of disruptions and to understand disruption effects including how a plasma disrupts onto the wall and how reproducibly it does so. Studies of disruptions on TFTR at both high-β pol and high-density have shown that, in both types, a fast growing m/n=1/1 mode plays an important role. In highdensity disruptions, a newly observed fast m/n = 1/1 mode occurs early in the thermal decay phase. For the first time in TFTR q-profile measurements just prior to disruptions have been made. Experimental studies of heat deposition patterns on the first wall of TFTR due to disruptions have provided information on MHD phenomena prior to or during the disruption, how the energy is released to the wall, and the reproducibility of the heat loads from disruptions. This information is important in the design of future devices such as ITER. Several new processes of runaway electron generation are theoretically suggested and their application to TFTR and ITER is considered, together with a preliminary assessment of x-ray data from runaways generated during disruptions

TFTR tritium operations lessons learned

International Nuclear Information System (INIS)

Gentile, C.A.; Raftopoulos, S.; LaMarche, P.

1996-01-01

The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980's The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America's first operational D-T fusion reactor. This paper will discuss negative pressure employing 'elephant trunks' in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and Δ pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs

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

PBX/TFTR pellet program PPPL

International Nuclear Information System (INIS)

Schmidt, G.

1986-01-01

Goals, current results and plans for pellet injection work for the PBX and TFTR programs are outlined. The present PBX injector is a prototype for ORNL 4 pellet condensing injectors. It has demonstrated that pellet injection on PBX can be used to increase overall density and alter the density profile. Future PBX operation requires reliable operation in deuterium and tritium, multiple pellet capability and ability to vary the size of pellets. These goals will require the construction of a new injector similar to the TFTR DPI system. It has also been demonstrated that pellets can efficiently fuel TFTR, producing a clean, high density plasma. Issues which are still outstanding include isotope exchange effects, use of different pellet sizes, optimization of pellet density perturbations and pellet penetration at high beam power

Material selection for TFTR limiters

International Nuclear Information System (INIS)

Ulrickson, M.

1980-10-01

The requirements for the material to be used as the first surface of limiters in TFTR are that it: (1) withstand a heat flux of 1 kw/cm 2 for a pulse length of 1.5s and a duty cycle of 1/200 for 10 5 cycles, (2) withstand the thermal and electro-magnetic loads from 10 4 plasma current disruptions lasting about 200 μs, (3) generate impurities at a rate low enough to meet impurity control requirements (which depend on the atomic number of the material) for TFTR, and (4) have tritium retention characteristics consistent with tritium inventory requirements for TFTR. An extensive set of material tests using electron beams, neutral beams, and plasma bombardment have been carried out to identify materials which can meet the thermal requirements of the above

Safety review and approval process for the TFTR

International Nuclear Information System (INIS)

Levine, J.D.; Howe, H.J.; Howe, K.E.

1983-01-01

The design, construction, and operation of the Tokamak Fusion Test Reactor (TFTR) has undergone an extensive safety and enviromental analysis involving Princeton Plasma Physics Laboratory (PPPL), the U.S. Department of Energy (DOE), the Ebasco/Grumman Industrial Subcontractor Team, and other organizations. This analysis, which is continuing during the TFTR operational phase, has been facilitated by the preparation, review and approval of several documents, including an Environmental Statement (Draft and Final), a Preliminary Safety Analysis Report (PSAR), a Final Safety Analysis Report (FSAR), Operations Safety Requirements (OSRs) and Safety Requirements (SRs), and various Operating and Maintenance Manuals. Through TFTR Safety Group participation in formal system design evaluations, change control boards, and reviews of project procurement and installation documentation, the TFTR Management Configuration Control System assures that all aspects of the project, including proposed design, installation and operational changes, receive prompt and thorough safety analyses. These efforts will continue as the TFTR Program moves into the neutral beam and D-T operational phases. The safety review and approval experience that has been acquired on the TFTR Project should serve as a foundation for similar efforts on future fusion devices

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

Measurements of charged fusion product diffusion in TFTR

Energy Technology Data Exchange (ETDEWEB)

Boivin, Rejean Louis [Princeton Univ., NJ (United States)

1991-12-01

The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator [ZnS(Ag)] and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current (≳ 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model.

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.)

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)

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.

Operations and maintenance plans for the TFTR

International Nuclear Information System (INIS)

Allen, H.L.; Fedor, B.J.

1978-01-01

Princeton University Plasma Physics Laboratory (PPPL) is constructing a Tokamak Fusion Test Reactor (TFTR) scheduled to begin operation for fusion research experiments in late 1981, first with hydrogen and deuterium plasmas and later, in the second phase, using tritium for high power fusion studies. This latter mode will introduce considerable complexity to operation and maintenance of the TFTR in terms of meeting requirements for tritium handling, adequate radiation shielding, and corrective and preventive maintenance procedures. In this paper we discuss plans for the installation and preoperational testing of the major subsystems of TFTR, proposed start-up and operating scenarios for the device and the system of operational control. In addition, the TFTR Maintenance Plan and related procedures for specific major maintenance tasks are described, including the use of remote handling equipment and remote manipulators. Each of these topics is addressed in terms of the current status of planning and development

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

Layout of the manipulator-arm (boom) for the TFTR fusion reactor (Princeton, USA) under UHV-conditions

International Nuclear Information System (INIS)

Klaubert, J.

1987-01-01

This presentation shows the main criteria for the layout of the manipulator - arm and the antechamber - vessel of the TFTR - FUSION - REACTOR at Princeton University, PLASMA PHYSICS LABORATORY (USA). The main problem during layout of a manipulator system like the TFTR - Boom has been the limitation of the vertical deflections due to deadweight of the construction. The design problem is rather a deformation problem and a problem of stability than a stress problem. The way of optimizing the ratio between stiffness and deadweight is the most important part during the complete design - process. Additional earthquake requirements need further investigations for a satisfying layout (horizontal forces, weak-axis of moment of inertia). The details of the construction (welding, connections etc.) have to be designed in respect to UHV - requirements --> no holes and no fillet welds (outgasing - rate.) are allowed. All weldings have to be designed as bevel-welds. This manipulator system is designed for working in a plane system (two degrees of freedom). A manipulator system with the same operating capabilities in a three degree of freedom system needs larger cross sections for the different beam-elements than those of the discussed TFTR - BOOM

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.

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.

Experimental results from detached plasmas in TFTR

International Nuclear Information System (INIS)

Strachan, J.D.; Boody, F.P.; Bush, C.E.

1986-10-01

Detached plasmas are formed in TFTR which have the principal property of the boundary to the high temperature plasma core being defined by a radiating layer. This paper documents the properties of TFTR ohmic-detached plasmas with a range of plasma densities at two different plasma currents

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

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.

High beta and confinement studies on TFTR

International Nuclear Information System (INIS)

Navratil, G.A.; Bhattacharjee, A.; Iacono, R.; Mauel, M.E.; Sabbagh, S.A.; Kesner, J.

1992-01-01

A new regime of high poloidal beta operation in TFTR was developed in the course of the first two years of this project (9/25/89 to 9/24/91). Our proposal to continue this successful collaboration between Columbia University and the Massachusetts Institute of Technology with the Princeton Plasma Physics Laboratory for a three year period (9/25/91 to 9/24/94) to continue to investigate improved confinement and tokamak performance in high poloidal beta plasmas in TFTR through the DT phase of operation was approved by the DOE and this is a report of our progress during the first 9 month budget period of the three year grant (9/25/91 to 6/24/92). During the approved three year project period we plan to (1) extend and apply the low current, high QDD discharges to the operation of TFTR using Deuterium and Tritium plasma; (2) continue the analysis and plan experiments on high poloidal beta phenomena in TFTR including: stability properties, enhanced global confinement, local transport, bootstrap current, and divertor formation; (3) plan and carry out experiments on TFTR which attempt to elevate the central q to values > 2 where entry to the second stability regime is predicted to occur; and (4) collaborate on high beta experiments using bean-shaped plasmas with a stabilizing conducting shell in PBX-M. In the seven month period covered by this report we have made progress in each of these four areas through the submission of 4 TFTR Experimental Proposals and the partial execution of 3 of these using a total of 4.5 run days during the August 1991 to February 1992 run

Fast current ramp experiments on TFTR

International Nuclear Information System (INIS)

Fredrickson, E.D.; McGuire, K.; Goldston, R.J.

1987-05-01

Electron heat transport on TFTR and other tokamaks is several orders of magnitude larger than neoclassical calculations would predict. Despite considerable effort, there is still no clear theoretical understanding of this anomalous transport. The electron temperature profile T/sub e/(r), shape has shown a marked consistency on many machines, including TFTR, for a wide range of plasma parameters and heating profiles. This could be an important clue as to the process responsible for this enhanced thermal transport. In this paper 'profile consistency' in TFTR is described and an experiment which uses a fast current ramp to transiently decouple the current density profile J(r), and the T/sub e/(r) profiles is discussed. From this experiment the influence of J(r) on electron temperature profile consistency can be determined

The high density and high βpol disruption mechanism on TFTR

International Nuclear Information System (INIS)

Fredrickson, E.D.; Manickam, J.; McGuire, K.M.; Monticello, D.; Nagayama, Y.; Park, W.; Taylor, G.

1992-01-01

Studies of disruptions on TFTR have been extended to include high density disruptions as well as the high β pol disruptions. The data strongly suggests that the (m,n)=(1,1) mode plays an important role in both types of disruptions. Further, for the first time, it is unambiguously shown, using a fast electron cyclotron emission (ECE) instrument for the electron temperature profile measurements, that the (m,n)=(1,1) precursor to the high density disruptions has a 'cold bubble' structure. The precursor to the major disruption at high density resembles the 'vacuum bubble' model of disruptions first proposed by Kadomtsev and Pogutse. (author) 2 refs., 2 figs

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

Engineering design of TFTR and it's impact on future tokamaks

International Nuclear Information System (INIS)

Sabado, M.M.

1981-01-01

TFTR is a second generation tokamak whose key objective is scientific break-even. TFTR is expected to be the first machine to demonstrate proper combination of plasma confinement time, density, and temperature to obtain this objective. A summary of major TFTR design parameters, including TFM, is presented, and their potential impact on future tokamaks discussed. Details of the updated engineering design and analysis of components are described. Status of major hardware fabrication, assembly installation and test are reviewed. TFTR features, technology, predicted performance and their potential implication for future tokamaks are summarized

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.

Residual effects of chromium gettering on the outgassing behavior of a stainless steel vacuum vessel

International Nuclear Information System (INIS)

Simpkins, J.E.; Blanchard, W.R.; Dylla, H.F.; LaMarche, P.H.

1986-05-01

Laboratory experiments that compared chromium and titanium gettering showed that with chromium, unlike titanium, there is no appreciable diffusion of hydrogen isotopes into the film. It was concluded from these experiments that chromium gettering on tokamaks is more desirable than titanium gettering, since chromium should provide higher hydrogen recycling, minimize tritium inventories, and avoid hydrogen embrittlement. Large-scale sublimation sources, consisting of hollow elongated chromium spheres with internal resistance heaters, were developed for use on tokamaks. These sources have been used to getter both the Impurity Study Experiment (ISX) and the Tokamak Fusion Test Reactor (TFTR). In both bases, significant effects on plasma performance were observed, including lower Z/sub eff/ and radiated power losses and an increase in the density limit. In TFTR these effects were observed for a period of weeks after a single chromium deposition. This paper reports the results of laboratory experiments made to examine the gettering characteristics of chromium films under conditions simulating those in TFTR

Experimental results from the TFTR tokamak

International Nuclear Information System (INIS)

Hawryluk, R.J.; Arunasalam, V.; Bell, J.D.

1986-10-01

Recent experiments on TFTR have extended the operating regime of TFTR in both ohmic- and neutral-beam-heated discharges. The TFTR tokamak has reached its original machine design specifications (I/sub p/ = 2.5 MA and B/sub T/ = 5.2 T). Initial neutral-beam-heating experiments used up to 6.3 MW of deuterium beams. With the recent installation of two additional beamlines, the power has been increased up to 11 MW. A deuterium pellet injector was used to increase the central density to 2.5 x 10 20 m -3 in high current discharges. At the opposite extreme, by operating at low plasma current (I/sub p/ ∼ 0.8 MA) and low density (anti n/sub e/ ∼ 1 x 10 19 m -3 ), high ion temperatures (9 +- 2 keV) and rotation speeds (7 x 10 5 m/s) have been achieved during injection. In addition, plasma compression experiments have demonstrated acceleration of beam ions from 82 keV to 150 keV, in accord with expectations. The wide operating range of TFTR, together with an extensive set of diagnostics and a flexible control system, has facilitated transport and scaling studies of both ohmic- and neutral-beam-heated discharges. The results of these confinement studies are presented

Enhanced carbon influx into TFTR supershots

International Nuclear Information System (INIS)

Ramsey, A.T.; Bush, C.E.; Dylla, H.F.; Owens, D.K.; Pitcher, C.S.; Ulrickson, M.

1990-12-01

Under some conditions, a very large influx of carbon into TFTR occurs during beam injection into low recycling plasmas (the Supershot regime). These carbon ''blooms'' result in serious degradation of plasma parameters. The sources of this carbon have been identified as hot spots on the TFTR bumper limiter at or near the last closed flux surface. Two separate temperature thresholds have been identified. One, at about 1650 degree C, is consistent with radiation enhanced sublimation. The other, at about 2300 degree C, appears to be thermal sublimation of carbon from the limiter. To account for the increased density caused by the blooms, near unity recycling of the carbon at the limiter by physical sputtering is required; this effect is expected from laboratory measurements, and we believe we are seeing it on TFTR. The sources of the carbon blooms are sites which have either loosely attached fragments of limiter material (caused by damage) or surfaces nearly perpendicular to the magnetic field lines. Such surfaces may have local power depositions two orders of magnitude higher than usual. The TFTR team modified the limiter during the opening of Winter 1989--90. The modifications greatly reduced the number and magnitude of the blooms, so that they are no longer a problem

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)

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.)

Measurements of charged fusion product diffusion in TFTR

Energy Technology Data Exchange (ETDEWEB)

Boivin, R.L.

1991-12-01

The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator (ZnS(Ag)) and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current ({approx gt} 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model.

Measurements of charged fusion product diffusion in TFTR

International Nuclear Information System (INIS)

Boivin, R.L.

1991-12-01

The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator [ZnS(Ag)] and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current (approx-gt 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model

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

Measurements of TFTR D-T radiation shielding efficiency

International Nuclear Information System (INIS)

Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Ku, L.P.; Levine, J.; Rule, K.; Azziz, N.; Goldhagen, P.; Hajnal, F.

1994-11-01

Measurements of neutron and gamma dose-equivalents were performed in the Test Cell, at the outer Test Cell wall, in nearby work areas, and out to the nearest property lines at a distance of 180 m. Argon ionization chambers, moderated 3 He proportional counters, and fission chamber detectors were used to obtain measurements of neutron and gamma dose-equivalents per D-T neutron during individual TFTR discharges. These measured neutron and gamma D-T dose-equivalents per TFTR neutron characterize the effects of local variations in material density resulting from the complex asymmetric site geometry. The measured dose-equivalents per TFTR D-T neutron and the cumulative neutron production were used to determine that the planned annual TFTR neutron production of 1 x 10 21 D-T neutrons is consistent with the design objective of limiting the total dose-equivalent at the property line, from all radiation sources and pathways, to less than 10 mrem per year

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)

Neutron emission from TFTR supershots

International Nuclear Information System (INIS)

Strachan, J.D.; Bell, M.G.; Bitter, M.; Budny, R.; Hawryluk, R.; Hill, K.W.; Hsuan, H.; Jassby, D.L.; Johnson, L.C.; LeBlanc, B.; Mansfield, D.; Meade, D.; Mikkelsen, D.R.; Mueller, D.; Park, H.; Ramsey, A.; Scott, S.; Synakowski, E.; Taylor, G.; Marmer, E.; Snipes, J.; Terry, J.

1992-10-01

Empirical scaling relations are deduced describing the neutron emission from TFTR supershots using a data base that includes all of the supershot plasmas (525) from the 1990 campaign. A physics-based scaling for the neutron emission is derived from the dependence of the central plasma parameters on machine settings and the energy confinement time. This scaling has been used to project the fusion rate for equivalent DT plasmas in TFTR, and to explore machine operation space which optimizes the fusion rate. Increases in neutron emission are possible by either increasing the toroidal magnetic field or further improving the limiter conditioning

TFTR Motor Generator

International Nuclear Information System (INIS)

Murray, J.G.; Bronner, G.; Horton, M.

1977-01-01

A general description is given of 475 MVA pulsed motor generators for TFTR at Princeton Plasma Physics Laboratory. Two identical generators operating in parallel are capable of supplying 950 MVA for an equivalent square pulse of 6.77 seconds and 4,500 MJ at 0.7 power factor to provide the energy for the pulsed electrical coils and heating system for TFTR. The description includes the operational features of the 15,000 HP wound rotor motors driving each generator with its starting equipment and cycloconverter for controlling speed, power factor, and regulating line voltage during load pulsing where the generator speed changes from 87.5 to 60 Hz frequency variation to provide the 4,500 MJ or energy. The special design characteristics such as fatigue stress calculations for 10 6 cycles of operation, forcing factor on exciter to provide regulation, and low generator impedance are reviewed

Enhanced carbon influx into TFTR supershots

International Nuclear Information System (INIS)

Ramsey, A.T.; Bush, C.E.; Dylla, H.F.; Owens, D.K.; Pitcher, C.S.; Ulrickson, M.A.

1991-01-01

Under some conditions, a very large influx of carbon into TFTR occurs during neutral beam injection into low recycling plasmas (the supershot regime). These carbon ''blooms'' result in serious degradation of plasma parameters. The sources of this carbon have been identified as hot spots on the TFTR bumper limiter at or near the last closed flux surface. Two separate temperature thresholds have been identified. One threshold, at about 1650 deg. C, is consistent with radiation enhanced sublimation (RES). The other, at about 2300 deg. C, appears to be thermal sublimation of carbon from the limiter. The carbon influx can be quantitatively accounted for by taking laboratory values for RES rates, making reasonable assumptions about the extent of the blooming area and assuming unity carbon recycling at the limiter. Such high carbon recycling is expected, and it is shown that, in target plasmas at least, it is observed on TFTR. The sources of the carbon blooms are sites which have either loosely attached fragments of limiter material (caused by damage) or surfaces that are nearly perpendicular to the magnetic field lines. Such surfaces may have local power depositions two orders of magnitude higher than usual. The TFTR team modified the limiter during the opening of winter 1989-1990. The modifications greatly reduced the number and magnitude of the blooms, so that they are no longer a problem. (author). 27 refs, 9 figs

FW/Blanket and vacuum vessel for RTO/RC ITER

International Nuclear Information System (INIS)

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

2000-01-01

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

FW/Blanket and vacuum vessel for RTO/RC ITER

Energy Technology Data Exchange (ETDEWEB)

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

2000-11-01

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

Recent D-T results on TFTR

International Nuclear Information System (INIS)

Johnson, D.W.; Arunasalam, V.

1995-10-01

Routine tritium operation in TFTR has permitted investigations of alpha particle physics in parameter ranges resembling those of a reactor core. ICRF wave physics in a DT plasma and the influence of isotopic mass on supershot confinement have also been studied. Continued progress has been made in optimizing fusion power production in TFTR, using extended machine capability and Li wall conditioning. Performance is currently limited by MHD stability. A new reversed magnetic shear regime is being investigated with reduced core transport and a higher predicted stability limit

TFTR initial operations

International Nuclear Information System (INIS)

Young, K.M.; Bell, M.; Blanchard, W.R.

1984-01-01

TFTR (Tokamak Fusion Test Reactor) has operated since December 1982 with ohmically heated plasmas. Routine operation with feedback control of plasma current, position and density has been obtained for plasmas with Isub(p) approx.= 800 kA, a = 68 cm, R = 250 cm, and Bsub(t) = 27 kG. A maximum plasma current of 1 MA was achieved with q approx.= 2.5. Energy confinement times of approx. 150 msec were measured for hydrogen and deuterium plasmas with n-barsub(e) approx.= 2 x 10 13 cm -3 , Tsub(e)(O) approx.= 1.5 keV, Tsub(i)(O) approx.= 1.5 keV and Zsub(eff) approx.= 3. The preliminary results suggest a size-cubed scaling from PLT, and are consistent with Alcator C scaling where tau approx. nR 2 a. Initial measurements of plasma disruption characteristics indicate current decay rates of approx. 800 kA in 8 ms which is within the TFTR design requirement of 3 MA in 3 ms. (author)

TFTR neutral-beam power system

International Nuclear Information System (INIS)

Winje, R.A.

1982-10-01

The TFTR Neutral Beam Power System (NBPS) consists of the accelerator grid power supply and the auxiliary power supplies required to operate the TFTR 120-keV ion sources. The current configuration of the NBPS including the 11-MVA accelerator grid power supply and the Arc and Filament power supplies isolated for operation at accelerator grid voltages up to 120 kV, is described. The prototype NBPS has been assembled at the Princeton Plasma Physics Laboratory and has been operated. The results of the initial operation and the description and resolution of some of the technical problems encountered during the commissioning tests are presented

TFTR neutral beam power system

International Nuclear Information System (INIS)

Deitz, A.; Murray, H.; Winje, R.

1977-01-01

The TFTR NB System will be composed of four beam lines, each containing three ion sources presently being developed for TFTR by the Lawrence Berkeley Laboratories (LBL). The Neutral Beam Power System (NBPS) will provide the necessary power required to operate these Ion Sources in both an experimental or operational mode as well as test mode. This paper describes the technical as well as the administrative/management aspects involved in the development and building of this system. The NBPS will combine the aspects of HV pulse (120 kV) and long pulse width (0.5 sec) together to produce a high power system that is unique in the Electrical Engineering field

Overview of the first workshop on alpha particle physics in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Biglari, H.

1991-07-01

The ''First Workshop on Alpha Physics in TFTR'' was held at the Princeton Plasma Physics Lab March 28--29, 1991. The motivation for this meeting was to clarify and strengthen the TFTR alpha physics program, and to increase the involvement of the fusion community outside PPPL in the TFTR D-T experiments. Therefore the meeting was sharply focused on alpha physics relevant to the upcoming TFTR D-T simulation, and was asked to devote half of his talk to specific TFTR issues. The Workshop consisted of 27 talks on: (1) experimental possibilities; (2) theoretical possibilities; (3) diagnostic possibilities; (4) relevance for future machines; and (5) discussion/summary session. This summary contains a brief sampling of the new results and ideas brought out by these talks, followed by two more general overviews of the status of experiment and theory

Radiation shielding for TFTR DT diagnostics

International Nuclear Information System (INIS)

Ku, L.P.; Johnson, D.W.; Liew, S.L.

1994-01-01

The authors illustrate the designs of radiation shielding for the TFTR DT diagnostics using the ACX and TVTS systems as specific examples. The main emphasis here is on the radiation transport analyses carried out in support of the designs. Initial results from the DT operation indicate that the diagnostics have been functioning as anticipated and the shielding designs are satisfactory. The experience accumulated in the shielding design for the TFTR DT diagnostics should be useful and applicable to future devices, such as TPX and ITER, where many similar diagnostic systems are expected to be used

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.)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-11-01

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

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

Impurity pellet injection experiments at TFTR

International Nuclear Information System (INIS)

Marmar, E.S.

1991-01-01

Impurity (Li and C) pellet experiments, which began at TFTR in 1989, and are expected to continue at least through 1991, have continued to produce new and significant results. The most significant of these are: (1) improvements in TFTR supershots after wall-conditioning by Li pellet injection; (2) accurate measurements of the pitch angle profiles of the internal magnetic field using the polarization angles of line emission from Li + in the pellet ablation cloud; and (3) initial measurements of pitch angle profiles using the tilt of the LI + emission region of the ablation cloud which is stretched out along the field lines

TFTR initial operations

International Nuclear Information System (INIS)

Young, K.M.; Bell, M.; Blanchard, W.R.

1983-11-01

The Tokamak Fusion Test Reactor (TFTR) has operated since December 1982 with ohmically heated plasmas. Routine operation with feedback control of plasma current, position, and density has been obtained for plasmas with I/sub p/ approx. = 800 kA, a = 68 cm, R = 250 cm, and B/sub t/ = 27 kG. A maximum plasma current of 1 MA was achieved with q approx. = 2.5. Energy confinement times of approx. 150 msec were measured for hydrogen and deuterium plasmas with anti n/sub e/ approx. = 2 x 10 13 cm -3 , T/sub e/ (0) approx. = 1.5 keV, T/sub i/ (0) approx. = 1.5 keV, and Z/sub eff/ approx. = 3. The preliminary results suggest a size-cubed scaling from PLT and are consistent with Alcator C scaling where tau approx. nR 2 a. Initial measurements of plasma disruption characteristics indicate current decay rates of approx. 800 kA in 8 ms which is within the TFTR design requirement of 3 MA in 3 ms

Transport analysis of TFTR experiments

International Nuclear Information System (INIS)

Goldston, R.; McCune, D.; Zarnstorff, M.; Hammett, G.; Scott, S.

1991-01-01

The purpose of this investigation was to complete the analysis of TFTR data which was under progress. The main emphasis was to study the effects of heating profile and resulting density and temperature profiles on transport through the comparison between beam heated plasmas with hollow and centrally peaked heating profiles (edge vs. center heating). The analysis has been completed and a manuscript has been prepared for publication in Nuclear Fusion. A proposal to perform a similar experiment using ICRF heating to decouple heating profile effects from density profile effects was submitted and was approved by the TFTR. ICRF heating enables the heating profile and the power partition between ions and electrons to be controlled. The experiment was scheduled twice, but it had to be postponed both times

Impurity control in TFTR

International Nuclear Information System (INIS)

Cecchi, J.L.

1980-06-01

The control of impurities in TFTR will be a particularly difficult problem due to the large energy and particle fluxes expected in the device. As part of the TFTR Flexibility Modification (TEM) project, a program has been implemented to address this problem. Transport code simulations are used to infer an impurity limit criterion as a function of the impurity atomic number. The configurational designs of the limiters and associated protective plates are discussed along with the consideration of thermal and mechanical loads due to normal plasma operation, neutral beams, and plasma disruptions. A summary is given of the materials-related research, which has been a collaborative effort involving groups at Argonne National Laboratory, Sandia Laboratories, and Princeton Plasma Physics Laboratory. Conceptual designs are shown for getterng systems capable of regenerating absorbed tritium. Research on this topic by groups at the previously mentioned laboratories and SAES Research Laboratory is reviewed

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.

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.

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)

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.

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

Coil protection calculator for TFTR

International Nuclear Information System (INIS)

Marsala, R.J.; Lawson, J.E.; Persing, R.G.; Senko, T.R.; Woolley, R.D.

1989-01-01

A new coil protection system (CPS) is being developed to replace the existing TFTR magnetic coil fault detector. The existing fault detector sacrifices TFTR operating capability for simplicity. The new CPS, when installed in October of 1988, will permit operation up to the actual coil stress limits parameters in real-time. The computation will be done in a microprocessor based Coil Protection Calculator (CPC) currently under construction at PPL. THe new CPC will allow TFTR to operate with higher plasma currents and will permit the optimization of pulse repetition rates. The CPC will provide real-time estimates of critical coil and bus temperatures and stresses based on real-time redundant measurements of coil currents, coil cooling water inlet temperature, and plasma current. The critical parameter calculations are compared to prespecified limits. If these limits are reached or exceeded, protection action will be initiated to a hard wired control system (HCS), which will shut down the power supplies. The CPC consists of a redundant VME based microprocessor system which will sample all input data and compute all stress quantities every ten milliseconds. Thermal calculations will be approximated every 10ms with an exact solution occurring every second. The CPC features continuous cross-checking of redundant input signal, automatic detection of internal failure modes, monitoring and recording of calculated results, and a quick, functional verification of performance via an internal test system. (author)

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

International Nuclear Information System (INIS)

Pessi, P.

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Pessi, P.

2009-07-01

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

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 �

Stack and area tritium monitoring systems for the tokamak fusion test reactor (TFTR)

International Nuclear Information System (INIS)

Pearson, G.G.; Meixler, L.D.; Sirsingh, R.A.P.

1992-01-01

This paper reports on the TFTR Tritium Stack and Area Monitoring Systems which have been developed to provide the required level of reliability in a cost effective manner consistent with the mission of the Tritium Handling System on TFTR. Personnel protection, environmental responsibility, and tritium containing system integrity have been the considerations in system design. During the Deuterium-Tritium (D-T) experiments on TFTR, tritium will be used for the first time as one of the fuels. Area monitors provide surveillance of the air in various rooms at TFTR. Stack monitors monitor the air at the TFTR test site that is exhausted through the HVAC systems, from the room exhaust stacks and the tritium systems process vents. The philosophies for the implementation of the Stack and Area Tritium Monitoring Systems at TFTR are to use hardwired controls wherever personnel protection is involved, and to take advantage of modern intelligent controllers to provide a distributed system to support the functions of tracking, displaying, and archiving concentration levels of tritium for all of the monitored areas and stacks

Recent results from TFTR

International Nuclear Information System (INIS)

Hawryluk, R.J.; Bell, M.G.; Bitter, M.

1984-05-01

During the past year, the research activities on TFTR have encompassed three broad areas. The first was to extend the operating range of TFTR. Plasma currents up to 1.5 MA were achieved in discharges with a = 0.83 m, R = 2.55 m at a toroidal field of 2.7 T. In these large plasmas, the maximum line average density was 3.35 x 10 19 m -3 . The second activity was a study of the scaling of the energy confinement time, tau/sub E/, in ohmically heated discharges as a function of plasma current, density, and plasma size. These experiments indicate a favorable scaling of tau/sub E/ with size and density. Energy confinement times in excess of 0.25 s were obtained in deuterium discharges. The third activity was a study of adiabatic compression. During compression, the plasma current and ion temperature scaled approximately as predicted; however, the electron temperature and density scaled less strongly than predicted for ideal compression

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)

Upgrade to the Tritium Remote Control and Monitoring System for TFTR D and D

International Nuclear Information System (INIS)

Sichta, P.; Oliaro, G.; Sengupta, S.

2002-01-01

Since 1988, the Tritium Remote Control and Monitoring System (TRECAMS) has performed crucial functions in support of D-T [deuterium-tritium] operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). Although plasma operations on TFTR were completed in 1997, the need for TRECAMS continued. During this period TRECAMS supported the TFTR tritium systems, the TFTR's Shutdown and Safing phase, and the TFTR Decontamination and Decommissioning (D and D) project. The most critical function of the TRECAMS in the post-TFTR era has been to provide a real-time indication of the airborne tritium levels in the tritium areas and the (HVAC) stacks. TRECAMS is a critical tool in conducting safe TFTR D and D tritium-line breaks and other tritium-related work activities. Beginning in 1998, the failure rate of the system's hardware sharply increased. Furthermore, the specialized knowledge required to maintain the original software and hardware was diminishing. It soon became apparent that a failure of the TRECAMS could significantly impact the TFTR D and D project's cost and schedule. To preclude this, the TRECAMS hardware and software was upgraded in the year 2000 to use modern components. This paper will describe that successful upgrade, including a review of the engineering processes and our operating experiences with the upgraded system

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

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

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.)

Plasma wall interaction and tritium retention in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Amarescu, E.; Ascione, G.

1996-01-01

The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced an impressive number of record breaking results including core fusion power, ∼ 2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently, the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. The authors present recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues

Plasma wall interaction and tritium retention in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Amarescu, E.; Ascione, G.

1997-01-01

The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced a number of record breaking results including core fusion power, ∝2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. Recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues are presented. (orig.)

Development of the TFTR neutral beam injection system

International Nuclear Information System (INIS)

Prichard, B.A. Jr.

1978-01-01

The TFTR Neutral Beam Lines are designed to inject 20 MW of 120 keV neutral deuterium atoms into the plasma. This is accomplished using 12 sources, 65 amperes each, mounted in 4 beam lines. The 120 kV sources are being developed by LBL and a prototype beam line which will be tested at Berkeley is being developed as a cooperative effort by LLL and LBL. The implementation of these beam lines has required the development of several associated pieces of hardware. The control and monitoring of the 12 sources will be done via the TFTR computer control system (CICADA) as will other parts of the machine, and software is being developed to condition and operate the sources automatically. The prototype beam line is scheduled to begin operation in the fall of 1978 and all four production beam lines on TFTR in 1982

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

Helium transport in TFTR

International Nuclear Information System (INIS)

Strachan, J.D.; Chan, A.

1986-09-01

Initial measurements of the 15 MeV protons produced in TFTR by the d( 3 He, p)α fusion reaction have been used to determine the time evolution of the central 3 He density. The signals following short 3 He gas puffs indicate inward transport times of about 100 msec

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

TFTR L mode energy confinement related to deuterium influx

International Nuclear Information System (INIS)

Strachan, J.D.

1999-01-01

Tokamak energy confinement scaling in TFTR L mode and supershot regimes is discussed. The main result is that TFTR L mode plasmas fit the supershot scaling law for energy confinement. In both regimes, plasma transport coefficients increased with increased edge deuterium influx. The common L mode confinement scaling law on TFTR is also inversely proportional to the volume of wall material that is heated to a high temperature, possibly the temperature at which the deuterium sorbed in the material becomes detrapped and highly mobile. The deuterium influx is increased by: (a) increased beam power due to a deeper heated depth in the edge components and (b) decreased plasma current due to an increased wetted area as governed by the empirically observed dependence of the SOL width upon plasma current. (author). Letter-to-the-editor

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

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.

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.

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)

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

Tritium pellet injection sequences for TFTR

International Nuclear Information System (INIS)

Houlberg, W.A.; Milora, S.L.; Attenberger, S.E.; Singer, C.E.; Schmidt, G.L.

1983-01-01

Tritium pellet injection into neutral deuterium, beam heated deuterium plasmas in the Tokamak Fusion Test Reactor (TFTR) is shown to be an attractive means of (1) minimizing tritium use per tritium discharge and over a sequence of tritium discharges; (2) greatly reducing the tritium load in the walls, limiters, getters, and cryopanels; (3) maintaining or improving instantaneous neutron production (Q); (4) reducing or eliminating deuterium-tritium (D-T) neutron production in non-optimized discharges; and (5) generally adding flexibility to the experimental sequences leading to optimal Q operation. Transport analyses of both compression and full-bore TFTR plasmas are used to support the above observations and to provide the basis for a proposed eight-pellet gas gun injector for the 1986 tritium experiments

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.)

TFTR neutral-beam test facility

International Nuclear Information System (INIS)

Turitzin, N.M.; Newman, R.A.

1981-11-01

TFTR Neutral Beam System will have thirteen discharge ion sources, each with its own power supply. Twelve of these will be utilized for supplemental heating of the TFTR tokamak plasma, while the thirteenth will be dedicated to an off-machine test chamber for source development and/or conditioning. A test installation for one source was set up using prototype equipment to discover and correct possible deficiencies, and to properly coordinate the equipment. This test facility represents the first opportunity for assembling an integrated system of hardware supplied by diverse vendors, each of whom designed and built his equipment to performance specifications. For the installation and coordination of the different portions of the total system, particular attention was given to personnel safety and safe equipment operation. This paper discusses various system components, their characteristics, interconnection and control. Results of the recently initiated test phase will be reported at a later date

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

Neutron diagnostics on TFTR utilizing the Campbelling technique

International Nuclear Information System (INIS)

England, A.C.; Hendel, H.W.; Neischmidt, E.B.

1986-01-01

The authors report modified commercial neutron counting equipment installed on a tokamak fusion test reactor (TFTR) which utilizes the Campbelling theorem to monitor the neutron source strength at very high neutron count rates. Campbelling utilizes the large amplitude fluctuation from neutron events in the detectors to discriminate against small amplitude noise events. Source strengths yielding equivalent count rates a factor of five higher than possible in the conventional count rate mode have been obtained to date. The concept of Campbelling is discussed and the particular application to TFTR is illustrated

Transport simulations TFTR: Theoretically-based transport models and current scaling

International Nuclear Information System (INIS)

Redi, M.H.; Cummings, J.C.; Bush, C.E.; Fredrickson, E.; Grek, B.; Hahm, T.S.; Hill, K.W.; Johnson, D.W.; Mansfield, D.K.; Park, H.; Scott, S.D.; Stratton, B.C.; Synakowski, E.J.; Tang, W.M.; Taylor, G.

1991-12-01

In order to study the microscopic physics underlying observed L-mode current scaling, 1-1/2-d BALDUR has been used to simulate density and temperature profiles for high and low current, neutral beam heated discharges on TFTR with several semi-empirical, theoretically-based models previously compared for TFTR, including several versions of trapped electron drift wave driven transport. Experiments at TFTR, JET and D3-D show that I p scaling of τ E does not arise from edge modes as previously thought, and is most likely to arise from nonlocal processes or from the I p -dependence of local plasma core transport. Consistent with this, it is found that strong current scaling does not arise from any of several edge models of resistive ballooning. Simulations with the profile consistent drift wave model and with a new model for toroidal collisionless trapped electron mode core transport in a multimode formalism, lead to strong current scaling of τ E for the L-mode cases on TFTR. None of the theoretically-based models succeeded in simulating the measured temperature and density profiles for both high and low current experiments

Anomalous delayed loss of trapped D-D fusion products in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Darrow, D.S.; Fredrickson, E.D.; Mynick, H.E.

1993-02-01

A new anomalous delayed loss of D-D fusion products has been measured at the bottom of the TFRR vessel. This loss is delayed by ∼ 0.2 sec with respect to the usual prompt first-orbit loss, and has a correspondingly lower energy, i.e. about half the fusion product birth energy. This loss process dominates the total fusion product loss measured 90 degrees below the midplane for plasma currents. I≥ 1.8 MA and major radii near R=2.45 m, e.g. for recent TFTR supershots. This delayed feature can occur without large coherent MED activity, although it can be strongly modulated by such activity. Several possible causes for this phenomenon are discussed, but no clear explanation for this delayed loss has yet been found

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-11-01

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

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.)

Two frequency ICRF operation on TFTR

International Nuclear Information System (INIS)

Rogers, J.H.; Majeski, R.; Wilson, J.R.; Hosea, J.C.; Schilling, G.; Stevens, J.; Phillips, C.K.

1993-01-01

Modifications have been made recently to allow two of the ICRF antennas (bays L and M) on TFTR to operate at either of two frequencies, 43 MHz or 64 MHz. This was accomplished by lengthening the resonant loops (2Λ at 43 MHz, 3Λ at 64 MHz) and replacing the conventional quarter wave impedance transformers with a tapered impedance design. The other two antennas (bays K and N) will operate at a fixed frequency, 43 MHz. The two frequency operation will allow a combination of 3 He-minority and H-minority heating at near full field on TFTR. The higher frequency, 64 MHz, may also be useful in direct electron heating and current drive experiments at lower toroidal fields. Models of the antenna, resonant loops and impedance matching system are presented

Diagnostic interface problems on TFTR

International Nuclear Information System (INIS)

Goldfarb, S.

1977-01-01

Diagnostic equipment on TFTR has functional interfaces with many machine systems. Salient requirements include plasma access, environmental resistance to thermal, magnetic and radiation effects, automated data acquisition and controls, remote handling and personnel safety. Problems imposed by these requirements and the solutions being considered are described

Design and R and D for the ITER vacuum vessel

International Nuclear Information System (INIS)

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

1998-01-01

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

Design and R and D for the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

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

1998-07-01

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

Design of a tritium pellet injector for TFTR

International Nuclear Information System (INIS)

Milora, S.L.; Gouge, M.J.; Fisher, P.W.; Combs, S.K.; Cole, M.J.; Wysor, R.B.; Fehling, D.T.; Foust, C.R.; Baylor, L.R.; Schmidt, G.L.; Barnes, G.W.; Persing, R.G.

1991-01-01

The TFTR tritium pellet injector (TPI) is designed to provide a tritium pellet fueling capability with pellet speeds in the 1- to 3 km/s-range for the TFTR D-T phase. The existing TFTR deuterium pellet injector is being modified at Oak Ridge National Laboratory to provide a fourshot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns a two -stage light gas gun driver. The pipe gun concept has been qualified for tritium operation by the tritium proof-of-principle injector experiments conducted on the Tritium Systems Test Assembly at Los Alamos National Laboratory. In these experiments, tritium and D-T pellets were accelerated to speeds near 1.5 km/s. The TPI is being designed for pellet sizes in the range from 3.43 to 4.0 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation will be controlled by a programmable logic controller. 7 refs., 4 figs

TFTR toroidal field coil design

International Nuclear Information System (INIS)

Smith, G.E.; Punchard, W.F.B.

1977-01-01

The design of the Tokamak Fusion Test Reactor (TFTR) Toroidal Field (TF) magnetic coils is described. The TF coil is a 44-turn, spiral-wound, two-pancake, water-cooled configuration which, at a coil current of 73.3 kiloamperes, produces a 5.2-Tesla field at a major radius of 2.48 meters. The magnetic coils are installed in titanium cases, which transmit the loads generated in the coils to the adjacent supporting structure. The TFTR utilizes 20 of these coils, positioned radially at 18 0 intervals, to provide the required toroidal field. Because it is very highly loaded and subject to tight volume constraints within the machine, the coil presents unique design problems. The TF coil requirements are summarized, the coil configuration is described, and the problems highlighted which have been encountered thus far in the coil design effort, together with the development tests which have been undertaken to verify the design

Measurement of Tritium Surface Distribution on TFTR Bumper Limiter Tiles

International Nuclear Information System (INIS)

Sugiyama, K.; Tanabe, T.; Skinner, C.H.; Gentile, C.A.

2004-01-01

The tritium surface distribution on graphite tiles used in the Tokamak Fusion Test Reactor (TFTR) bumper limiter and exposed to TFTR deuterium-tritium (D-T) discharges from 1993 to 1997 was measured by the Tritium Imaging Plate Technique (TIPT). The TFTR bumper limiter shows both re-/co-deposition and erosion. The tritium images for all tiles measured are strongly correlated with erosion and deposition patterns, and long-term tritium retention was found in the re-/co-depositions and flakes. The CFC tiles located at erosion dominated areas clearly showed their woven structure in their tritium images owing to different erosion yields between fibers and matrix. Significantly high tritium retention was observed on all sides of the erosion tiles, indicating carbon transport via repetition of local erosion/deposition cycles

Decontamination and decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Walton, G.R.; Perry, E.D.; Commander, J.C.; Spampinato, P.T.

1994-01-01

The Tokamak Fusion Test Reactor (TFTR) is scheduled to complete its end-of-life deuterium-tritium (D-T) experiments in September 1994. The D-T operation will result in the TFTR machine structure becoming activated, and plasma facing and vacuum components will be contaminated with tritium. The resulting machine activation levels after a two year cooldown period will allow hands on dismantling for external structures, but require remote dismantling for the vacuum vessel. The primary objective of the Decontamination and Decommissioning (D ampersand D) Project is to provide a facility for construction of a new Department of Energy (DOE) experimental fusion reactor by March 1998. The project schedule calls for a two year shutdown period when tritium decontamination of the vacuum vessel, neutral beam injectors and other components will occur. Shutdown will be followed by an 18 month period of D ampersand D operations. The technical objectives of the project are to: safely dismantle and remove components from the test cell complex; package disassembled components in accordance with applicable regulations; ship packages to a DOE approved disposal or material recycling site; and develop expertise using remote disassembly techniques on a large scale fusion facility. This paper discusses the D ampersand D objectives, the facility to be decommissioned, and the technical plan that will be implemented

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

High performance deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

Sabbagh, S.A.; Bell, M.G.

1995-03-01

Plasmas composed of nominally equal concentrations of deuterium and tritium (DT) have been created in TFTR with the goals of producing significant levels of fusion power and of examining the effects of DT fusion alpha particles. Conditioning of the limiter by the injection of lithium pellets has led to an approximate doubling of the energy confinement time, τ E , in supershot plasmas at high plasma current (I p ≤ 2.5 MA) and high heating power (P b ≤ 33 MW). Operation with DT typically results in an additional 20% increase in τ E . In the high poloidal beta, advanced tokamak regime in TFTR, confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.5 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasmas, exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Confinement of alpha particles appears to be classical and losses due to collective effects have not been observed. While small fluctuations in fusion product loss were observed during ELMs, no large loss was detected in DT plasmas

TFTR vertically viewing electron cyclotron emission diagnostic

International Nuclear Information System (INIS)

Taylor, G.

1990-01-01

The Tokamak Fusion Test Reactor (TFTR) Michelson interferometer has a spectral coverage of 75--540 GHz, allowing measurement of the first four electron cyclotron harmonics. Until recently the instrument has been configured to view the TFTR plasma on the horizontal midplane, primarily in order to measure the electron temperature profile. Electron cyclotron emission (ECE) extraordinary mode spectra from TFTR Supershot plasmas exhibit a pronounced, spectrally narrow feature below the second harmonic. A similar feature is seen with the ECE radiometer diagnostic below the electron cyclotron fundamental frequency in the ordinary mode. Analysis of the ECE spectra indicates the possibility of a non-Maxwellian 40--80 keV tail on the electron distribution in or near the core. During 1990 three vertical views with silicon carbide viewing targets will be installed to provide a direct measurement of the electron energy distribution at major radii of 2.54, 2.78, and 3.09 m with an energy resolution of approximately 20% at 100 keV. To provide the maximum flexibility, the optical components for the vertical views will be remotely controlled to allow the Michelson interferometer to be reconfigured to either the midplane horizontal view or one of the three vertical views between plasma shots

Extension of TFTR operations to higher toroidal field levels

International Nuclear Information System (INIS)

Woolley, R.D.

1995-01-01

For the past year, TFTR has sometimes operated at extended toroidal field (TF) levels. The extension to 5.6 Tesla (79 kA) was crucial for TFTR's November 1994 10.7 MW DT fusion power record. The extension to 6.0 Tesla (85 kA) was commissioned on 9 September 1995. There are several reasons that one could expect the TF coils to survive the higher stresses that develop at higher fields. They were designed to operate at 5.2 Tesla with a vertical field of 0.5 Tesla, whereas the actual vertical field needed for the plasma does not exceed 0.35 Tesla. Their design specification explicitly required they survive some pulses at 6.0 Tesla. TF coil mechanical analysis computer models available during coil design were crude, leading to conservative design. And design analyses also had to consider worst-case misoperations that TFTR's real time Coil Protection Calculators (CPCs) now positively prevent from occurring

Decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Spampinato, P.T.; Walton, G.R.

1993-01-01

The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) will complete its experimental lifetime with a series of deuterium-tritium pulses in 1994. As a result, the machine structures will become radioactive, and vacuum components will also be contaminated with tritium. Dose rate levels will range from less than 1 mr/h for external structures to hundreds of mr/h for the vacuum vessel. Hence, decommissioning operations will range from hands on activities to the use of remotely operated equipment. After 21 months of cool down, decontamination and decommissioning (D and D) operations will commence and continue for approximately 15 months. The primary objective is to render the test cell complex re-usable for the next machine, the Tokamak Physics Experiment (TPX). This paper presents an overview of decommissioning TFTR and discusses the D and D objectives

TFTR plasma feedback systems

International Nuclear Information System (INIS)

Efthimion, P.; Hawryluk, R.J.; Hojsak, W.; Marsala, R.J.; Mueller, D.; Rauch, W.; Tait, G.D.; Taylor, G.; Thompson, M.

1985-01-01

The Tokamak Fusion Test Reactor employs feedback control systems for four plasma parameters, i.e. for plasma current, for plasma major radius, for plasma vertical position, and for plasma density. The plasma current is controlled by adjusting the rate of change of current in the Ohmic Heating (OH) coil system. Plasma current is continuously sensed by a Rogowski coil and its associated electronics; the error between it and a preprogrammed reference plasma current history is operated upon by a ''proportional-plusintegral-plus-derivative'' (PID) control algorithm and combined with various feedforward terms, to generate compensating commands to the phase-controlled thyristor rectifiers which drive current through the OH coils. The plasma position is controlled by adjusting the currents in Equilibrium Field and Horizontal Field coil systems, which respectively determine the vertical and radial external magnetic fields producing J X B forces on the plasma current. The plasma major radius position and vertical position, sensed by ''B /sub theta/ '' and ''B /sub rho/ '' magnetic flux pickup coils with their associated electronics, are controlled toward preprogrammed reference histories by allowing PID and feedforward control algorithms to generate commands to the EF and HF coil power supplies. Plasma density is controlled by adjusting the amount of gas injected into the vacuum vessel. Time-varying gains are used to combine lineaveraged plasma density measurements from a microwave interferometer plasma diagnostic system with vacuum vessel pressure measurements from ion gauges, with various other measurements, and with preprogrammed reference histories, to determine commands to piezoelectric gas injection valves

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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.

TFTR D-T results

International Nuclear Information System (INIS)

Meade, D.M.

1994-01-01

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of ∼ 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of ∼1 and yielded a maximum fusion power of ∼ 9.2 MW. The fusion power density in the core of the plasma was ∼ 1.8 MW m -3 approximating that expected in a D-T fusion reactor. A TFTR plasma with T/D density ratio of ∼ 1 was found to have ∼ 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of τ E ∼ A 0.6 . The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. The ∼ 5% loss of alpha particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined energetic alphas and the resultant alpha ash density. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfven Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed

Assessment of the dynamic behaviours of the ITER Vacuum Vessel

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Design 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

Thermal consequences of plasma disruptions in TFTR and ETF

International Nuclear Information System (INIS)

Budny, R.; Ludescher, C.

1981-01-01

We studied thermal responses of first walls for TFTR and ETF during plasma disruptions. To model the flux, we assumed the entire kinetic energy is deposited by axisymmetric horizontal displacement of the plasma. The deposition time is a free parameter. In TFTR, the minimum deposition time which does not cause the toroidal limiter to melt is 7 or 14 ms depending on whether or not the limiter is actively cooled. In ETF, the minimum time which does not cause surface melting of the cooling tubes is 80 ms. (author)

Engineering analysis of TFTR disruption

Energy Technology Data Exchange (ETDEWEB)

Murray, J.G.; Rothe, K.E.; Bronner, G.

1984-09-01

This report covers an engineering approach quantifying the currents, forces, and times, as well as plasma position, for the worst-case disruption based on engineerign circuit assumptions for the plasma. As the plasma moves toward the wall during the current-decay phase of disruption, the wall currents affect the rate of movement and, hence, the decay time. The calculated structure-induced currents differ considerably from those calculated using a presently available criterion, which specifies that the plasma remains stationary in the center of the torus while decaying in 10 ms. This report outlines the method and basis for the engineering calculation used to determine the current and forces as a function of the circuit characteristics. It provides specific calculations for the Tokamak Fusion Test Reactor (TFTR) with variations in parameters such as the thermal decay time, the torus resistance, and plasma temperature during the current decay. The study reviews possible ways to reduce the disruption damage of TFTR by reducing the magnitude of the plasma external field energy that is absorbed by the plasma during the current decay.

Engineering analysis of TFTR disruption

International Nuclear Information System (INIS)

Murray, J.G.; Rothe, K.E.; Bronner, G.

1984-09-01

This report covers an engineering approach quantifying the currents, forces, and times, as well as plasma position, for the worst-case disruption based on engineerign circuit assumptions for the plasma. As the plasma moves toward the wall during the current-decay phase of disruption, the wall currents affect the rate of movement and, hence, the decay time. The calculated structure-induced currents differ considerably from those calculated using a presently available criterion, which specifies that the plasma remains stationary in the center of the torus while decaying in 10 ms. This report outlines the method and basis for the engineering calculation used to determine the current and forces as a function of the circuit characteristics. It provides specific calculations for the Tokamak Fusion Test Reactor (TFTR) with variations in parameters such as the thermal decay time, the torus resistance, and plasma temperature during the current decay. The study reviews possible ways to reduce the disruption damage of TFTR by reducing the magnitude of the plasma external field energy that is absorbed by the plasma during the current decay

120-keV beam direct conversion system for TFTR injectors

International Nuclear Information System (INIS)

Hamilton, G.W.

1976-01-01

Several practical motivations exist for the development of beam direct conversion systems that are compatible with the injection systems of large experiments such as the Tokamak Fusion Test Reactor (TFTR). We present a preliminary design in which we analyze the most acute problems involved in scaling up existing designs and apparatus to fulfill TFTR requirements. Some of the questions addressed are the requirements for electron suppression, gas pumping, compactness, and power densities. A new idea is presented that allows for the handling of higher beam power. The gross savings in the capital cost of injector power supplies for the TFTR will be about $7.2 million, but the net savings will be somewhat less than this. This preliminary design has not yet revealed fundamental limitations with respect to the development of beam energy-recovery systems operating at high levels of current, voltage, and power densities

A Neutral Beam Injector Upgrade for NSTX

International Nuclear Information System (INIS)

Stevenson, T.; McCormack, B.; Loesser, G.D.; Kalish, M.; Ramakrishnan, S.; Grisham, L.; Edwards, J.; Cropper, M.; Rossi, G.; Halle, A. von; Williams, M.

2002-01-01

The National Spherical Torus Experiment (NSTX) capability with a Neutral Beam Injector (NBI) capable of 80 kiloelectronvolt (keV), 5 Megawatt (MW), 5 second operation. This 5.95 million dollar upgrade reused a previous generation injector and equipment for technical, cost, and schedule reasons to obtain these specifications while retaining a legacy capability of 120 keV neutral particle beam delivery for shorter pulse lengths for possible future NSTX experiments. Concerns with NBI injection included power deposition in the plasma, aiming angles from the fixed NBI fan array, density profiles and beam shine through, orbit losses of beam particles, and protection of the vacuum vessel wall against beam impingement. The upgrade made use of the beamline and cryo panels from the Neutral Beam Test Stand facility, existing power supplies and controls, beamline components and equipment not contaminated by tritium during DT [deuterium-tritium] experiments, and a liquid Helium refrigerator plant to power and cryogenically pump a beamline and three ion sources. All of the Tokamak Fusion Test Reactor (TFTR) ion sources had been contaminated with tritium, so a refurbishment effort was undertaken on selected TFTR sources to rid the three sources destined for the NSTX NBI of as much tritium as possible. An interconnecting duct was fabricated using some spare and some new components to attach the beamline to the NSTX vacuum vessel. Internal vacuum vessel armor using carbon tiles was added to protect the stainless steel vacuum vessel from beam impingement in the absence of plasma and interlock failure. To date, the NBI has operated to 80 keV and 5 MW and has injected requested power levels into NSTX plasmas with good initial results, including high beta and strong heating characteristics at full rated plasma current

Automatic generation of computer programs servicing TFTR console displays

International Nuclear Information System (INIS)

Eisenberg, H.

1983-01-01

A number of alternatives were considered in providing programs to support the several hundred displays required for control and monitoring of TFTR equipment. Since similar functions were performed, an automated method of creating programs was suggested. The complexity of a single program servicing as many as thirty consoles mitigated against that approach. Similarly, creation of a syntactic language while elegant, was deemed to be too time consuming, and had the disadvantage of requiring a working knowledge of the language on a programming level. It was elected to pursue a method of generating an individual program to service a particular display. A feasibility study was conducted and the Control and Monitor Display Generator system (CMDG) was developed. A Control and Monitor Display Service Program (CMDS) provides a means of performing monitor and control functions for devices associated with TFTR subsystems, as well as other user functions, via TFTR Control Consoles. This paper discusses the specific capabilities provided by CMDS in a usage context, as well as the mechanics of implementation

Compilation of TFTR materials data

International Nuclear Information System (INIS)

Havener, W.J.

1975-12-01

In order to document the key thermophysical property data used in the conceptual design of Tokamak Fusion Test Reactor (TFTR) systems and components, a series of data packages has been prepared. It is expected that data for additional materials will be added and the information already provided will be updated to provide a project-wide data base

Parametric variations of ion transport in TFTR

International Nuclear Information System (INIS)

Scott, S.D.; Ernst, D.

1993-01-01

This paper is divided into three roughly independent sections. The first is a historical review of the twenty year history of experimental ion heat transport measurements from many tokamaks. The second is a study of ion heat transport in Ohmic TFTR plasmas which shows that χi ∼ χe ∼ 15χi neo . Thus, ion heat transport is demonstrated to be strongly anomalous even the absence of auxiliary heating. The third section describes the variation of χi with local ion temperature in TFTR during auxiliary heating, with emphasis on characterizing the differecens between transport in the L-mode and supershot regimes. The results are consistent with the conjecture that improved ion energy confinement in supershot plasmas is caused by a high ratio of T 1 /T e

β limit disruptions in the TFTR tokamak

International Nuclear Information System (INIS)

Fredrickson, E.D.; McGuire, K.; Janos, A.; Bell, M.; Budny, R.V.; Bush, C.E.; Manickam, J.; Mynick, H.; Nazikian, R.; Taylor, G.

1994-11-01

A disruptive β limit (β = plasma pressure/magnetic pressure) is observed in high performance plasmas in TFTR. The MHD character of these disruptions differs substantially from the disruptions in high density plasmas (density limit disruptions) on TFTR. The high β disruptions can occur with less than a milliseconds warning in the form of a fast growing precursor. The precursor appears to be an external kink or internal (m,n)=(1,1) kink strongly coupled through finite β effects and toroidal terms to higher m components. It does not have the open-quote cold bubble close-quote structure found in density limit disruptions. There is also no evidence for a change in the internal inductance, i.e., a major reconnection of the flux, at the time of the thermal quench

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Finite element modeling of TFTR poloidal field coils

International Nuclear Information System (INIS)

Baumgartner, J.A.; O'Toole, J.A.

1986-01-01

The Tokamak Fusion Test Reactor (TFTR) Poloidal Field (PF) coils were originally analyzed to TFTR design conditions. The coils have been reanalyzed by PPPL and Grumman to determine operating limits under as-built conditions. Critical stress levels, based upon data obtained from the reanalysis of each PF coil, are needed for input to the TFTR simulation code algorithms. The primary objective regarding structural integrity has been to ascertain the magnitude and location of critical internal stresses in each PF coil due to various combinations of electromagnetic and thermally induced loads. For each PF coil, a global finite element model (FEM) of a coil sector is being analyzed to obtain the basic coil internal loads and displacements. Subsequent fine mesh local models of the coil lead stem and lead spur regions produce the magnitudes and locations of peak stresses. Each copper turn and its surrounding insulation are modeled using solid finite elements. The corresponding electromagnetic and thermal analyses are similarly modeled. A series of test beams were developed to determine the best combination of MSC/NASTRAN-type finite elements for use in PF coil analysis. The results of this analysis compare favorably with those obtained by the earlier analysis which was limited in scope

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)

Alpha diagnostics using pellet charge exchange: Results on TFTR and prospects for ITER

International Nuclear Information System (INIS)

Fisher, R.K.; Duong, H.H.; McChesney, J.M.

1996-05-01

Confinement of alpha particles is essential for fusion ignition and alpha physics studies are a major goal of the TFTR, JET, and ITER DT experiments, but alpha measurements remain one of the most challenging plasma diagnostic tasks. The Pellet Charge Exchange (PCX) diagnostic has successfully measured the radial density profile and energy distribution of fast (0.5 to 3.5 MeV) confined alpha particles in TFTR. This paper describes the diagnostic capabilities of PCX demonstrated on TFTR and discusses the prospects for applying this technique to ITER. Major issues on ITER include the pellet's perturbation to the plasma and obtaining satisfactory pellet penetration into the plasma

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Meteorological data summaries for the TFTR from March 1984 to February 1985

International Nuclear Information System (INIS)

Kolibal, J.; Ku, L.P.; Liew, S.L.; Pierce, C.

1985-06-01

This report reviews the first year of meteorological data gathered for the Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) from March 1, 1984 to February 28, 1985. The meteorological station at TFTR is located at D-Site, to the east of the motor generator building as shown in Fig. 1. The station consists of a 60 m tower which is instrumented at 10, 30, and 60 m along with the associated equipment for data acquisition and logging. Instrumentation for the tower consists of measuring the temperature, wind speed, wind direction, dew point, and the standard deviation of the horizontal wind direction. The purpose of the station is to gather site specific meteorological data to assess atmospheric transport and dispersion for TFTR

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.

TFTR CAMAC power supplies reliability

International Nuclear Information System (INIS)

Camp, R.A.; Bergin, W.

1989-01-01

Since the expected life of the Tokamak Fusion Test Reactor (TFTR) has been extended into the early 1990's, the issues of equipment wear-out, when to refurbish/replace, and the costs associated with these decisions, must be faced. The management of the maintenance of the TFTR Central Instrumentation, Control and Data Acquisition System (CICADA) power supplies within the CAMAC network is a case study of a set of systems to monitor repairable systems reliability, costs, and results of action. The CAMAC network is composed of approximately 500 racks, each with its own power supply. By using a simple reliability estimator on a coarse time interval, in conjunction with determining the root cause of individual failures, a cost effective repair and maintenance program has been realized. This paper describes the estimator, some of the specific causes for recurring failures and their correction, and the subsequent effects on the reliability estimator. By extension of this program the authors can assess the continued viability of CAMAC power supplies into the future, predicting wear-out and developing cost effective refurbishment/replacement policies. 4 refs., 3 figs., 1 tab

Ideal MHD stability of high poloidal beta equilibria in TFTR

International Nuclear Information System (INIS)

Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.; Bell, M.G.; Budny, R.V.; Chance, M.S.; Fredrickson, E.D.; Jardin, S.C.; Manickam, J.; McCune, D.C.; McGuire, K.M.; Wieland, R.M.; Zarnstorff, M.C.; Phillips, M.W.; Hughes, M.H.; Kesner, J.

1991-01-01

Recent experiments in TFTR have expanded the operating space of the device to include plasmas with values of var-epsilon β p dia ≡ 2μ 0 var-epsilon perpendicular >/ p >> 2 as large as 1.6, and Troyon normalized diamagnetic beta β N dia ≡ β t perpendicular aB t /10 -8 I p as large as 4.7. At values of var-epsilon β p dia ≥ 1.3, a separatrix was observed to enter the vacuum vessel, producing a naturally diverted discharge. Plasmas with large values of var-epsilon β p dia were created with both the plasma current, I p , held constant and with I p decreased, or ramped down, before the start of neutral beam injection. A convenient characterization of the change in I p using experimental parameters can be defined by the ratio of I p before the ramp down, to I p during the neutral beam heating phase, F I p . The ideal MHD stability of these equilibria is investigated to determine their location in stability space, and to study the role of plasma current and pressure profile modification in the creation of these high var-epsilon β p and β N plasmas. The evolution of these plasmas is modelled from experimental data using the TRANSP code. Two-dimensional equilibria are computed from the TRANSP results and used as input to both high and low-n stability codes including PEST. The high var-epsilon β p equilibria, which generally have an oblate cross-sectional shape, are in the first stability region to high-n ballooning modes. At constant I p , these equilibria generally have maximum pressure gradients near the magnetic axis and are stable to n=1 modes without a stabilizing conducting wall. The effect of the current profile shape on the stability of low-n kink/ballooning modes and the requirements for these plasmas to access the second stability region are examined. 6 refs

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)

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

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.)

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)

X-ray diagnostics for TFTR

International Nuclear Information System (INIS)

von Goeler, S.; Hill, K.W.; Bitter, M.

1982-12-01

A short description of the x-ray diagnostic preparation for the TFTR tokamak is given. The x-ray equipment consists of the limiter x-ray monitoring system, the soft x-ray pulse-height-analysis-system, the soft x-ray imaging system and the x-ray crystal spectrometer. Particular attention is given to the radiation protection of the x-ray systems from the neutron environment

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

TFTR 60 GHz alpha particle collective Thomson Scattering diagnostic

International Nuclear Information System (INIS)

Machuzak, J.S.; Woskov, P.P.; Gilmore, J.; Bretz, N.L.; Park, H.K.; Bindslev, H.

1995-03-01

A 60 GHz gyrotron collective Thomson Scattering alpha particle diagnostic has been implemented for the D-T period on TFM. Gyrotron power of 0.1-1 kW in pulses of up to 1 second can be launched in X-mode. Efficient corrugated waveguides are used with antennaes and vacuum windows of the TFTR Microwave Scattering system. A multichannel synchronous detector receiver system and spectrum analyzer acquire the scattered signals. A 200 Megasample/sec digitizer is used to resolve fine structure in the frequency spectrum. By scattering nearly perpendicular to the magnetic field, this experiment will take advantage of an enhancement of the scattered signal which results from the interaction of the alpha particles with plasma resonances in the lower hybrid frequency range. Significant enhancements are expected, which will make these measurements possible with gyrotron power less than 1 kW, while maintaining an acceptable signal to noise ratio. We hope to extract alpha particle density and velocity distribution functions from the data. The D and T fuel densities and temperatures may also be obtainable by measurement of the respective ion cyclotron harmonic frequencies

TFTR generator load assessment

International Nuclear Information System (INIS)

Heck, F.M.

1975-10-01

Typical experimental load demands on the TFTR generators are illustrated based on the electrical characteristics of the field coils, the coil leads, the main bus work, the various auxiliary bus work, the rectifiers, and transformers. The generator MW capacities are shown to be adequate for the proposed experimental operations with allowances made for variations in the final designs. The generator MVA capacities are shown to be adequate provided portions of the TF and EF rectifiers are freewheeled at selected times

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

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.

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.

Four-channel ZnS scintillator measurements of escaping tritons in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.

1988-10-01

A four-channel scintillation detector capable of measuring tritons, protons, and alphas escaping from a tokamak plasma was operated during the 1986 run period of the Tokamak Fusion Test Reactor (TFTR). Signals consistent with the expected 1 MeV triton behavior have been observed during deuterium operation. Backgrounds associated with neutrons, gammas, and soft x-rays have been evaluated in situ. Such a detector should be capable of measuring escaping alphas during the D/T phase of TFTR. 16 refs., 10 figs

Four-channel ZnS scintillator measurements of escaping tritons in TFTR

Energy Technology Data Exchange (ETDEWEB)

Zweben, S.J.

1988-10-01

A four-channel scintillation detector capable of measuring tritons, protons, and alphas escaping from a tokamak plasma was operated during the 1986 run period of the Tokamak Fusion Test Reactor (TFTR). Signals consistent with the expected 1 MeV triton behavior have been observed during deuterium operation. Backgrounds associated with neutrons, gammas, and soft x-rays have been evaluated in situ. Such a detector should be capable of measuring escaping alphas during the D/T phase of TFTR. 16 refs., 10 figs.

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].

Coherent and turbulent fluctuations in TFTR

International Nuclear Information System (INIS)

McGuire, K.; Arunasalam, V.; Bell, M.G.

1987-04-01

Classification of the sawteeth observed in the TFTR tokamak has been carried out to highlight the differences between the many types observed. Three types of sawteeth are discussed: ''simple,'' ''small,'' and ''compound.'' During the enhanced confinement discharges on TFTR, sawteeth related to q = 1 are usually not present, but a sawtooth-like event is sometimes observed. β approaches the Troyon limit only at low q/sub cyl/ with a clear reduction of achievable β/sub n/ at high q/sub cyl/. This suggests that a β/sub p/ limit, rather than the Troyon-Gruber limit, applies at high q/sub cyl/ in the enhanced confinement discharges. These discharges also reach the stability boundary for n → ∞ ideal MHD ballooning modes. Turbulence measurements in the scrape-off region with Langmuir and magnetic probes show strong edge density turbulence n/n = 0.3 - 0.5, with weak magnetic turbulence B/sub θ/B/sub θ/ > 5 x 10 -6 measured at the wall, but these measurements are very sensitive to local edge conditions

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-12-01

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

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

Pitch angle resolved measurements of escaping charged fusion products in TFTR

Energy Technology Data Exchange (ETDEWEB)

Zweben, S.J.

1989-01-01

Measurements of the flux of charged fusion products escaping from the TFTR plasma have been made with a new type of detector which can resolve the particle flux vs. pitch angle, energy, and time. The design of this detector is described, and results from the 1987 TFTR run are presented. These results are roughly consistent with predictions from a simple first-orbit particle loss model with respect to the pitch angle, energy, time, and plasma current dependence of the signals. 11 refs., 9 figs.

Pitch angle resolved measurements of escaping charged fusion products in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.

1989-01-01

Measurements of the flux of charged fusion products escaping from the TFTR plasma have been made with a new type of detector which can resolve the particle flux vs. pitch angle, energy, and time. The design of this detector is described, and results from the 1987 TFTR run are presented. These results are roughly consistent with predictions from a simple first-orbit particle loss model with respect to the pitch angle, energy, time, and plasma current dependence of the signals. 11 refs., 9 figs

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)

EMI free fiber optic strain sensor system for TFTR

International Nuclear Information System (INIS)

Szuchy, N.C.; Caserta, A.L.; Ferrara, A.A.; Squires, R.W.; Sredniawski, J.J.

1983-01-01

In certain applications, structural components are subjected to loadings in high electromagnetic interference (EMI) environments. The mechanical responses of these components must be monitored under rapidly varying electromagnetic fields. A Fiber Optic Strain Sensor System (FOSSS) is an acceptable solution since it is immune to EMI. Grumman Aerospace Corporation initiated the development of a FOSSS that can be used in high EMI situations where resistive/electronic-based strain measurement systems would not be effective, such as on the Tokamak Fusion Test Reactor (TFTR) during plasma disruption. Tests have indicated that because of their increased sensitivity due to the size of the fiber optic (FO) transducer (1-in. 2 ) and responsiveness due to the areal changes of the FO sensor, the strain tracking capability of FO sensors are excellent. For the TFTR application a jacketed 400-micron fiber capable of operating in a 250 0 C temperature environment was used. Continuous 30 foot lengths of high-temperature FO cables were affixed to 304 LN SS tabs, forming an integrated strain sensor and pigtail unit. By fusion splicing 400-micron room temperature fibers to the pigtails, the required runs (approximately 200 feet) to the TFTR data acquisition room were made with minimum coupling attenuation. Development methodology is discussed and test data presented

The TFTR 40 MW neutral beam injection system and DT operations

International Nuclear Information System (INIS)

Stevenson, T.; O'Connor, T.; Garzotto, V.

1995-01-01

Since December 1993, TFTR has performed DT experiments using tritium fuel provided mainly by neutral beam injection. Significant alpha particle populations and reactor-like conditions have been achieved at the plasma core, and fusion output power has risen to a record 10.7 MW using a record 40 MW NB heating. Tritium neutral beams have injected into over 480 DT plasmas and greater than 500 kCi have been processed through the neutral beam gas, cryo, and vacuum systems. Beam tritium injections, as well as tritium feedstock delivery and disposal, have now become part of routine operations. Shot reliability with tritium is about 90% and is comparable to deuterium shot reliability. This paper describes the neutral beam DT experience including the preparations, modifications, and operating techniques that led to this high level of success, as well as the critical differences in beam operations encountered during DT operations. Also, the neutral beam maintenance and repair history during DT operations, the corrective actions taken, and procedures developed for handling tritium contaminated components are discussed in the context of supporting a continuous DT program

TFTR magnetic field design analyses

International Nuclear Information System (INIS)

Davies, K.; Iwinski, E.; McWhirter, J.M.

1975-11-01

The three main magnetic field windings for the TFTR are the toroidal field (TF) windings, the ohmic heating (OH) winding, and the equilibrium field (EF) winding. The following information is provided for these windings: (1) descriptions, (2) functions, (3) magnetic designs, e.g., number and location of turns, (4) design methods, and (5) descriptions of resulting magnetic fields. This report does not deal with the thermal, mechanical support, or construction details of the windings

Central ignition scenarios for TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Redi, M.H.; Bateman, G.

1986-03-01

The possibility of obtaining ignition in TFTR by means of very centrally peaked density profiles is examined. It is shown that local central alpha heating can be made to exceed local central energy losses (''central ignition'') under global conditions for which Q greater than or equal to 1. Time dependent 1-D transport simulations show that the normal global ignition requirements are substantially relaxed for plasmas with peaked density profiles. 18 refs., 18 figs

TFTR Experimental Data Analysis Collaboration

International Nuclear Information System (INIS)

Callen, J.D.

1993-01-01

The research performed under the second year of this three-year grant has concentrated on a few key TFTR experimental data analysis issues: MHD mode identification and effects on supershots; identification of new MHD modes; MHD mode theory-experiment comparisons; local electron heat transport inferred from impurity-induced cool pulses; and some other topics. Progress in these areas and activities undertaken in conjunction with this grant are summarized briefly in this report

TFTR radiation contour and shielding efficiency measurements during D-D operations

International Nuclear Information System (INIS)

Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Hwang, D.; Lewis, M.; Levine, J.; Ku, L.P.; Rule, K.; Hajnal, F.

1994-11-01

Extensive neutron and gamma radiation contour, shielding efficiency, and spectral measurements were performed during high power TFTR D-D operations at the tokamak Test Cell inner walls, ceiling, roof, and outer walls, in nearby control rooms, work areas, and personnel pathways, outdoors along the site fence at 125 m, and out to the nearest property lines at 180 m. The results confirmed that the efficiency of the basic radiation shielding was sufficient to allow the TFTR D-T experimental plan, and provide empirical guidance for simulating the radiation fields of future fusion reactors

Proposed TFTR electrical system

International Nuclear Information System (INIS)

Bronner, G.; Murray, J.

1975-01-01

The development of controlled thermonuclear fusion has progressed to the stage where the present facilities and energy available for future devices are not sufficient and must be increased by about a factor of ten. This report describes the proposed TFTR ac utility power distribution system, an energy storage motor generator flywheel facility, and the rectifier conversion equipment for the Toroidal Field Confining System (TF), Ohmic Heating System (OH), Equilibrium Field System (EF) and the Neutral Beam Heating System (NB). The general requirements are described and the special design considerations identified

Evolution of the electron temperature profile of ohmically heated plasmas in TFTR

International Nuclear Information System (INIS)

Taylor, G.; Efthimion, P.C.; Arunasalam, V.

1985-08-01

Blackbody electron cyclotron emission was used to ascertain and study the evolution and behavior of the electron temperature profile in ohmically heated plasmas in the Tokamak Fusion Test Reactor (TFTR). The emission was measured with absolutely calibrated millimeter wavelength radiometers. The temperature profile normalized to the central temperature and minor radius is observed to broaden substantially with decreasing limiter safety factor q/sub a/, and is insensitive to the plasma minor radius. Sawtooth activity was seen in the core of most TFTR discharges and appeared to be associated with a flattening of the electron temperature profile within the plasma core where q less than or equal to 1. Two types of sawtooth behavior were identified in large TFTR plasmas (minor radius, a less than or equal to 0.8 m) : a typically 35 to 40 msec period ''normal'' sawtooth, and a ''compound'' sawtooth with 70 to 80 msec period

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

ORNL compact loop antenna design for TFTR and Tore Supra

International Nuclear Information System (INIS)

Taylor, D.J.; Baity, F.W.; Bryan, W.E.; Hoffman, D.J.; McIlwain, R.L.; Ray, J.M.

1987-01-01

The goal supplemental ion cyclotron resonance heating (ICRH) of fusion plasma is to deliver power at high efficiencies deep within the plasma. The technology for fast-wave ICRH has reached the point of requiring ''proof-of-performance'' demonstration of specific antenna configurations of specific antenna configurations and their mechanical adequacy for operating in a fusion environment. Oak Ridge National Laboratory (ORNL) has developed the compact loop antenna concept based on a resonant double loop (RDL) configuration for use in both Tokamak Fusion Test Reactor (TFTR) and the Tore Supra ICRH programs. A description and a comparison of the technologies developed in the two designs are presented. The electrical circuit and the mechanical philosophy employed are the same for both antennas, but different operating environments result in substantial differences in the design of specific components. The ORNL TFTR antenna is designed to deliver 4 MW over a 2-s pulse, and the ORNL Tore Supra antenna is designed for 4 MW and essentially steady-state conditions. The TFTR design embodies the first operations compact RDL antenna, and the Tore Supra antenna extends the technology to an operational duty cycle consistent with reactor-relevant applications. 7 refs., 5 figs

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

Neutron dosimetry for the TFTR Lithium-Blanket-Module program

International Nuclear Information System (INIS)

Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

1981-01-01

The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory

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.

DT results of TFTR's alpha collector

International Nuclear Information System (INIS)

Herrmann, H.W.; Zweben, S.J.; Darrow, D.S.; Timberlake, J.R.; Macaulay-Newcombe, R.G.

1996-01-01

An escaping alpha collector probe has been developed for TFTR's DT phase to complement the results of the lost alpha scintillator detectors which have been operating on TFTR since 1988. Measurements of the energy distribution of escaping alphas have been made by measuring the range of alphas implanted into nickel foils located within the alpha collector. Exposed samples have been analyzed for 4 DT plasma discharges at plasma currents of 1.0 and 1.8 MA. The results at 1.0 MA are in good agreement with predictions for first orbit alpha loss at 3.5 MeV. The 1.8 MA results, however, indicate a large anomalous loss of partially thermalized alphas at an energy ∼30% below the birth energy and at a total fluence nearly an order of magnitude above expected first orbit loss. This anomalous loss is not observed with the lost alpha scintillator detectors in DT plasmas but does resemble the anomalous delayed loss seen in DD plasmas. Several potential explanations for this loss process are examined. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations

The TFTR lithium blanket module program

International Nuclear Information System (INIS)

Jassby, D.L.; Bertone, P.C.; Creedon, R.L.; File, J.; Graumann, D.W.

1985-01-01

The Lithium Blanket Module (LBM) is an approximately 80X80X80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li 2 O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li 2 O pellets with satisfactory reproducibility were developed using purified Li 2 O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1-1nCi/g)

Impurity pellet injection experiments at TFTR

International Nuclear Information System (INIS)

Marmar, E.S.

1992-01-01

Impurity (Li and C) pellet injection experiments on TFTR have produced a number of new and significant results. (1) We observe reproducible improvements of TFTR supershots after wall-conditioning by Li pellet injection ('lithiumization'). (2) We have made accurate measurements of the pitch angle profiles of the internal magnetic field using two novel techniques. The first measures the internal field pitch from the polarization angles of Li + line emission from the pellet ablation cloud, while the second measures the pitch angle profiles by observing the tilt of the cigar-shaped Li + emission region of the ablation cloud. (3) Extensive measurements of impurity pellet penetration into plasmas with central temperatures ranging from ∼0.3 to ∼7 keV have been made and compared with available theoretical models. Other aspects of pellet cloud physics have been investigated. (4) Using pellets as a well defined perturbation has allowed study of transport phenomena. In the case of small pellet perturbations, the characteristics of the background plasmas are probed, while with large pellets, pellet induced effects are clearly observed. These main results are discussed in more detail in this paper

Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Perry, E.; Chrzanowski, J.; Rule, K.; Viola, M.; Williams, M.; Strykowsky, R.

1999-01-01

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling

A labview approach to instrumentation for the TFTR bumper limiter alignment project

International Nuclear Information System (INIS)

Skelly, G.N.; Owens, D.K.

1992-01-01

This paper reports on a project recently undertaken to measure the alignment of the TFTR bumper limiter in relation to the toroidal magnetic field axis. The process involved the measurement of the toroidal magnetic field, and the positions of the tiles that make up the bumper limiter. The basis for the instrument control and data acquisition system was National Instrument's LabVIEW 2. LabVIEW is a graphical programming system for developing scientific and engineering applications on a Macintosh. For this project, a Macintosh IIci controlled the IEEE-488 GPIB programmable instruments via an interface box connected to the SCSI port of the computer. With LabVIEW, users create graphical software modules called virtual instruments instead of writing conventional text-based code. To measure the magnetic field, the control system acquired data from two nuclear magnetic resonance magnetometers while the torroidal field coils were pulsed. To measure the position of the tiles on the limiter, an instrumented mechanical arm was used inside the vessel

ICRF antenna modifications and additions for TFTR: Relevance to BPX/ITER projections

International Nuclear Information System (INIS)

Hosea, J.; Phillips, C.K.; Raftopoulos, S.; Stevens, J.; Wilson, J.R.

1991-01-01

The TFTR Bay L and M antennas have been modified to improve their power handling capability. In particular, the Bay L antenna, which exhibited a lower than expected loading resistance, now has a configuration similar to that of Bay M -- slotted walls and septum -- and together with Bay M is expected to support 7 MW operations. The in situ loading enhancement achieved for the Modified Bay L design will serve to quantify models for the coupling effects of slots. Also, comparisons with Bay M loading performance will elucidate wave spectrum and antenna location (relative to in-vessel structures) effects. Two new antennas, with single/double row shields slanted at 6 degree (along B) are to be added in the near future to augment the power capability to ∼12.5 MW. The relevance of the four antenna array features to quantifying BPX/ITER antenna characteristic projections for heating and current drive is discussed. 8 refs., 5 figs

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-10-15

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

Probes for edge plasma studies of TFTR (invited)

International Nuclear Information System (INIS)

Manos, D.M.; Budny, R.V.; Kilpatrick, S.; Stangeby, P.; Zweben, S.

1986-01-01

Tokamak fusion test reactor (TFTR) probes are designed to study the interaction of the plasma with material surfaces such as the wall and limiters, and to study the transport of particles and energy between the core and edge. Present probe heads have evolved from prototypes in Princeton large torus (PLT), poloidal divertor experiment (PDX) [Princeton BETA experiment (PBX)], and the initial phase of TFTR operation. The newest heads are capable of making several simultaneous measurements and include Langmuir probes, heat flux probes, magnetic coils, rotating calorimeter fast ion probes, and sample exposure specimens. This paper describes these probe heads and presents some of the data they and their prototypes have acquired. The paper emphasizes measurement of transient plasma effects such as fast ion loss during auxiliary heating, the evolution of the edge plasma during heating, compression, and free expansion, and fluctuations in the edge plasma

ICRF sawtooth stabilization: Application on TFTR and CIT

International Nuclear Information System (INIS)

Hosea, J.C.; Phillips, C.K.; Stevens, J.E.; Wilson, J.R.; Bell, M.; Boivin, R.; Cavallo, A.; Colestock, P.; Fredrickson, E.; Hammett, G.; Hsuan, H.; Janos, A.; Jassby, D.; Jobes, F.; McGuire, K.; Mueller, D.; Nagayama, Y.; Owens, K.; Park, H.; Schmidt, G.; Stratton, B.; Taylor, G.; Wong, K.L.; Zweben, S.

1991-03-01

The use of ICRF heating to stabilize the core plasma sawtooth relaxations has been extended to TFTR where such stabilization has been produced at relatively low power in the L Mode regime at moderate density (P RF = 4 MW, 2.6 MW in helium and deuterium discharges, respectively, for the minority hydrogen ICRF heating regime with bar n e ∼2.5 x 10 13 cm -3 ). These results, as in the case of those obtained on JET, are qualitatively consistent with energetic ion stabilization of the m = 1, n = 1 ideal/resistive kink mode. The relevance of sawtooth stabilization to the primary regimes of interest on TFTR -- the high-Q supershot regime and the high density pellet injection regimes -- and on CIT -- the high density ICRF heated regime -- is considered in the context of the present theory and the projected ICRF power deposition characteristics. 35 refs., 11 figs

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)

Large eddy simulation of Loss of Vacuum Accident in STARDUST facility

International Nuclear Information System (INIS)

Benedetti, Miriam; Gaudio, Pasquale; Lupelli, Ivan; Malizia, Andrea; Porfiri, Maria Teresa; Richetta, Maria

2013-01-01

Highlights: ► Fusion safety, plasma material interaction. ► Numerical and experimental data comparison to analyze the consequences of Loss of Vacuum Accident that can provoke dust mobilization inside the Vacuum Vessel of the Nuclear Fusion Reactor ITER-like. -- Abstract: The development of computational fluid dynamic (CFD) models of air ingress into the vacuum vessel (VV) represents an important issue concerning the safety analysis of nuclear fusion devices, in particular in the field of dust mobilization. The present work deals with the large eddy simulations (LES) of fluid dynamic fields during a vessel filling at near vacuum conditions to support the safety study of Loss of Vacuum Accidents (LOVA) events triggered by air income. The model's results are compared to the experimental data provided by STARDUST facility at different pressurization rates (100 Pa/s, 300 Pa/s and 500 Pa/s). Simulation's results compare favorably with experimental data, demonstrating the possibility of implementing LES in large vacuum systems as tokamaks

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-11-01

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

Mechanical design of epithermal neutron diagnostic for TFTR

International Nuclear Information System (INIS)

Groo, R.C.

1981-01-01

The mechanical design of the Epithermal Neutron Diagnostic for TFTR is described. This fission detector system measures the time resolution of the neutron flux for folding into the Neutron Activation system and also provides continuous, wide range coverage of all expected fusion reaction rates

Three novel tokamak plasma regimes in TFTR

International Nuclear Information System (INIS)

Furth, H.P.

1985-10-01

Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

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

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

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)

Characteristics of radiated power for various TFTR [Tokamak Fusion Test Reactor] regimes

International Nuclear Information System (INIS)

Bush, C.E.; Schivell, J.; McNeill, D.H.

1988-04-01

Power loss studies were carried out to determine the impurity radiation and energy transport characteristics of various TFTR operation and confinement regimes including L-Mode, detached plasma, co-only neutral beam injection (energetic ion regime), and the enhanced confinement (''supershot'') regime. Combined bolometric, spectroscopic, and infrared photometry measurements provide a picture of impurity behavior and power accounting in TFTR. The purpose of this paper is to make a survey of the various regimes with the aim of determining the radiated power signatures of each. 10 refs., 6 figs., 1 tab

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

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

Expansion of the TFTR neutral beam computer system

International Nuclear Information System (INIS)

McEnerney, J.; Chu, J.; Davis, S.; Fitzwater, J.; Fleming, G.; Funk, P.; Hirsch, J.; Lagin, L.; Locasak, V.; Randerson, L.; Schechtman, N.; Silber, K.; Skelly, G.; Stark, W.

1992-01-01

Previous TFTR Neutral Beam computing support was based primarily on an Encore Concept 32/8750 computer within the TFTR Central Instrumentation, Control and Data Acquisition System (CICADA). The resources of this machine were 90% utilized during a 2.5 minute duty cycle. Both interactive and automatic processes were supported, with interactive response suffering at lower priority. Further, there were additional computing requirements and no cost effective path for expansion within the Encore framework. Two elements provided a solution to these problems: improved price performance for computing and a high speed bus link to the SELBUS. The purchase of a Sun SPARCstation and a VME/SELBUS bus link, allowed offloading the automatic processing to the workstation. This paper describes the details of the system including the performance of the bus link and Sun SPARCstation, raw data acquisition and data server functions, application software conversion issues, and experiences with the UNIX operating system in the mixed platform environment

Perspectives gained from ICRF physics studies on TFTR

International Nuclear Information System (INIS)

Phillips, C.K.; Bell, M.; Batha, S.

1998-01-01

The physics of ICRF heating and current drive has been studied on TFTR for over a decade. Following the early low power coupling studies, high power experiments resulted in sawtooth stabilization, the first observation of RF-driven excitation of toroidal Alfven eigenmodes, and the discovery of a mode conversion scenario for localized off-axis electron heating. The program culminated with the first studies of high power ICRF heating and profile control in tritium-rich high performance plasmas. A significant part of the concluding experiments centered on the potential of ICRF to drive sheared flows in order to suppress turbulence in the plasma core. Initial measurements taken with a novel poloidal velocity diagnostic suggest that localized sheared poloidal flows can be driven with ion Bernstein waves excited directly or else via mode conversion from a propagating fast magnetosonic wave. In this paper, recent results from TFTR on wave-based profile control techniques will be summarized along with suggestions for future studies elsewhere

Neutron diagnostics on TFTR utilizing the Campbelling technique

International Nuclear Information System (INIS)

England, A.C.; Hendel, H.W.; Nieschmidt, E.B.

1986-01-01

Modified commercial equipment installed on the tokamak fusion test reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) utilizes Campbell's mean square voltage theorem to monitor the neutron source strength at neutron count rates orders of magnitude above the capability of the count rate mode. Campbelling uses the large amplitude fluctuations from neutron fission events in the detectors to discriminate against small amplitude γ ray and other noise events. Source strengths yielding equivalent count rates a factor of 5 greater than possible in the conventional count rate mode have been obtained to date. The concept of Campbelling is discussed and the particular application to TFTR is illustrated. Fundamental advantages are the extended useful range of the detectors by a factor of --10 4 and gamma rejection by a factor of --10 3 . Some results are shown and the neutron source strengths obtained are compared to those from conventional counting circuits and from other detectors whose outputs have not yet suffered counting losses

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

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

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

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.

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.

Estimated neutron-activation data for TFTR. Part II. Biological dose rate from sample-materials activation

International Nuclear Information System (INIS)

Ku, L.; Kolibal, J.G.

1982-06-01

The neutron induced material activation dose rate data are summarized for the TFTR operation. This report marks the completion of the second phase of the systematic study of the activation problem on the TFTR. The estimations of the neutron induced activation dose rates were made for spherical and slab objects, based on a point kernel method, for a wide range of materials. The dose rates as a function of cooling time for standard samples are presented for a number of typical neutron spectrum expected during TFTR DD and DT operations. The factors which account for the variations of the pulsing history, the characteristic size of the object and the distance of observation relative to the standard samples are also presented

Deuterium-tritium TFTR plasmas in the high poloidal beta regime

International Nuclear Information System (INIS)

Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.

1995-03-01

Deuterium-tritium plasmas with enhanced energy confinement and stability have been produced in the high poloidal beta, advanced tokamak regime in TFTR. Confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.46 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasma,s exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Fusion power exceeding 6.7 MW with a fusion power gain Q DT = 0.22 has been produced with reduced alpha particle first orbit loss provided by the increased l i

Power and particle balance during neutral beam injection in TFTR

International Nuclear Information System (INIS)

Pitcher, C.S.; Budny, R.V.; Hill, K.W.; Kilpatrick, S.J.; Manos, D.M.; Medley, S.S.; Ramsey, A.T.

1991-05-01

Detailed boundary plasma measurements on TFTR have been made during a NBI power scan in the range P tot = 1MW--20MW in the L-mode regime. The behavior of the plasma density left-angle n e right-angle, radiated power P rad , carbon and deuterium fluxes Γ C , Γ D , and Ζ eff can be summarized as, left-angle n e right-angle ∝ P tot 1/2 , P rad , Γ C , Γ D ∝ P tot , and Ζ eff ∼ constant. It is shown that central fuelling by the neutral beams plays a minor role in the particle balance of the discharge. More important is the NBI role in the power balance. The TFTR data during NBI originate primarily at the graphite limiter

TFTR control and monitoring system (CICADA)

International Nuclear Information System (INIS)

Daniels, R.E.

1981-01-01

The TFTR Central Instrumentation, Control and Data Acquisition System (CICADA) is described. This is a computer based system, supporting three types of user interfaces and supporting real time, terminal, and batch operations. Over one hundred graphic display generators will be supported by the system, four array processors will greatly increase the analysis capabilities, and closed circuit television will distribute performance data throughout the facility. Approximately twenty thousand points wll be interfaced to the system

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.

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)

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

Manufacture of EAST VS In-Vessel Coil

International Nuclear Information System (INIS)

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

2013-01-01

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

Recent TFTR results

International Nuclear Information System (INIS)

Meade, D.M.; Arunasalam, V.; Bell, M.G.; Bell, R.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.; Cavallo, A.; Cheng, C.Z.; Chu, T.K.; Cohen, S.A.; Cowley, S.; Davis, S.L.; Dimock, D.L.; Efthimion, P.C.; Ehrhardt, A.B.; Fredrickson, E.; Furth, H.P.; Goldston, R.J.; Greene, G.; Grek, B.; Grisham, L.R.; Hammett, G.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.C.; Hulse, R.A.; Hsuan, H.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kaita, R.; Kaye, S.; Kieras-Phillips, C.; Kilpatrick, S.J.; Kugel, H.; La Marche, P.H.; LeBlanc, B.; Manos, D.M.; Mansfield, D.K.; Mazzucato, E.; McCarthy, M.P.; McCune, D.C.; McGuire, K.M.; Medley, S.S.; Mikkelsen, D.R.; Monticello, D.; Motley, R.; Mueller, D.; Murphy, J.; Nazikian, R.; Owens, D.K.; Park, H.; Park, W.; Paul, S.; Perkins, R.; Ramsey, A.T.; Redi, M.H.; Rewoldt, G.; Roquemore, A.L.; Rutherford, P.H.; Schilling, G.; Schivell, J.; Schmidt, G.L.; Scott, S.D.; Stevens, J.; Stodiek, W.; Stratton, B.C.; Synakowski, E.; Tang, W.A.; Taylor, G.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.; Williams, M.D.; Wilson, J.R.; Wong, K.L.; Yamada, M.; Yoshikawa, S.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.; Bush, C.E.; Dooling, J.; Dylla, H.F.; Fonck, R.J.; Roberts, D.; Howell, R.B.; Kesner, J.; Marmar, E.S.; Snipes, J.; Terry, J.L.; Nagayama, Y.; Pitcher, S.

1991-07-01

TFTR experiments have emphasized the optimization of high performance plasmas as well as studies of transport in high temperature plasmas. The recent installation of carbon composite tiles on the main bumper limiter has allowed operation with up to 32 MW of neutral beam injection without degradation of plasma performance by large bursts of carbon impurities (''carbon blooms''). Plasma parameters have been extended to T i (0) ∼ 35 keV, T e (0) ∼ 12 keV, n e (0) ∼1.2 x 10 20 m -3 producing D-D reaction rates of 8.8 x 10 16 reactions per second. The fusion parameter n e (0)τ E T i (0) in supershot plasmas is an increasing function of heating power up to an MHD stability limit, reaching values of ∼4.4 x 10 20 m -3 sec keV. Peaked-density-profile hot-ion plasmas with the edge characteristics of the H-mode have been produced in a circular cross-section limiter configuration with n e (0)τ E T i (0) values characteristic of supershots, namely up to four times those projected for standard H-modes with broad density profiles. Reduced transport is also observed in the core of high-density ICRF-heated plasmas when the density profile is peaked. At the highest performance, the central plasma pressure in TFTR reaches reactor level values of 6.5 atmospheres. In these regimes, MHD instabilities with m/n = 1/1, 2/1, 3/2 and 4/3 are often observed concurrent with a degradation in performance. High β p plasmas with var-epsilon β p ∼ 1.6 and β/(I/aB) ∼ 4.7 (%mT/MA) have demonstrated confinement enhancement over the low-mode confinement time with τ E /τ L ∼ 3.5 and a bootstrap current of about 65% of the total plasma current

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

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)

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)

A review of carbon blooms on JET and TFTR

International Nuclear Information System (INIS)

Ulrickson, M.

1990-01-01

Operation of JET and TFTR at high auxiliary heating power has resulted in the occurrence of phenomena called carbon blooms. The carbon bloom is characterized by a rapid increases in the emission of carbon spectral lines, the Z eff , the radiated power, and the plasma density. There is also a concurrent decrease in the neutron emission rate, stored energy, and plasma pressure. On both machines the source of the carbon is observed to be at localized (both toroidally and polidally) hot spots on either the divertor plates or limiters. The localized hot spots are due to one or more of the following: disruption damage spots, misalignment of tiles, and/or exposed edges of tiles. The occurrence of carbon blooms limits the performance of the highest input power plasmas on both machines. This paper reviews the carbon bloom phenomenon as it occurs on both JET and TFTR. (orig.)

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

Neutron spectroscopy on TFTR [Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Nishitani, T.; Strachan, J.D.

1988-05-01

This paper describes the use of an 3 He ionization chamber for neutron spectroscopy on TFTR during 1987. The ion temperature was measured using neutron spectroscopy for one set of ohmically heated plasmas. The deduced ion temperatures agreed to within 20% with those measured by other diagnostics. 11 refs., 11 figs., 1 tab

Theory-based transport simulations of TFTR L-mode temperature profiles

International Nuclear Information System (INIS)

Bateman, G.

1991-01-01

The temperature profiles from a selection of TFTR L-mode discharges are simulated with the 1-1/2-D BALDUR transport code using a combination of theoretically derived transport models, called the Multi-Mode Model. The present version of the Multi-Mode Model consists of effective thermal diffusivities resulting from trapped electron modes and ion temperature gradient (η i ) modes, which dominate in the core of the plasma, together with resistive ballooning modes, which dominate in the periphery. Within the context of this transport model and the TFTR simulations reported here, the scaling of confinement with heating power comes from the temperature dependence of the η i and trapped electron modes, while the scaling with current comes mostly from resistive ballooning modes. 24 refs., 16 figs., 3 tabs

DT simulation of ICRF heated supershots in TFTR using TRANSP

International Nuclear Information System (INIS)

Goldfinger, R.C.; Batchelor, D.B.; Phillips, C.K.; Budny, R.; Hammett, G.W.; Hosea, J.C.; McCune, D.M.; Stevens, J.E.; Wilson, J.R.

1993-01-01

The principal goal of ion cyclotron range of frequency (ICRF) heating on the Tokamak Fusion Test Reactor (TFTR) is to enhance plasma performance during the deuterium-tritium (DT) physics phase of operations. Strongly centralized ICRF heating may play a critical role in obtaining high Q DT and high β α operation in TFTR, as well as in future fusion reactors. ICRF heating of a dilute minority species leads to the formation of an energetic ion population that, in turn, provides strong central electron heating. The corresponding rise in the central electron temperature translates into an increase in the slowing-down time of either neutral beam or alpha particles in the discharge. Preliminary DT simulations of the experimental results in deuterium-deuterium (DD) plasmas performed with the TRANSP code are presented in this paper

Simulation of α-particle redistribution due to sawteeth on TFTR

International Nuclear Information System (INIS)

Yi Zhao; White, R.B.

1996-01-01

In recent Deuterium-Tritium experiments on the Tokamak Fusion Test Reactor (TFTR), both the Pellet Charge Exchange (PCX) and the alpha Charge Exchange Recombination Spectroscopy (α-CHERS) diagnostics indicate that sawtooth oscillations can cause significant broadening of the fusion alpha radial density profile. The authors investigate this sawtooth mixing phenomenon by applying a Hamiltonian guiding center approach. A model of time evolution of the Kadomtsev-type sawtooth is constructed. The presence of more than one mode in the nonlinear stage of the sawtooth crash is necessary to cause significant broadening of the alpha density profile. Use of numerical equilibria allows us to perform detailed comparisons with TFTR experimental data. The results are in reasonable agreement with α-CHERS and show a broadening of alpha particles similar to that seen in PCX measurements

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

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

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

Measurement of TFTR D-T radiation shielding efficiency

International Nuclear Information System (INIS)

Kugel, H.W.; Ascione G.; Elwood, S.

1994-01-01

High power D-T fusion reactor designs presently exhibit complex geometric and material density configurations. Simulations of the radiation shielding required for safe operation and full compliance with all regulatory requirements must include sufficient margin to accommodate uncertainties in material properties and distributions, uncertainties in the final configurations, and uncertainties in approximations employing the homogenization of complex geometries. Measurements of radiation shielding efficiency performed in a realistic D-T tokamak environment can provide empirical guidance for simulating safe, efficient, and cost effective shielding systems for future high power fusion reactors. In this work, the authors present the results of initial measurements of the TFTR radiation shielding efficiency during high power D-T operations with record neutron yields. The TFTR design objective is to limit the total dose-equivalent at the nearest PPPL property lines from all radiation pathways to 10 mrem per calendar year. Compliance with this design objective over a calendar year requires measurements in the presence of typical site backgrounds of about 80 mrem per year

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)

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

TFTR data management system

International Nuclear Information System (INIS)

Randerson, L.; Chu, J.; Ludescher, C.; Malsbury, J.; Stark, W.

1986-01-01

Developments in the tokamak fusion test reactor (TFTR) data-management system supporting data acquisition and off-line physics data reduction are described. Data from monitor points, timing channels, transient recorder channels, and other devices are acquired and stored for use by on-line tasks. Files are transferred off line automatically. A configuration utility determines data acquired and files transferred. An event system driven by file arrival activates off-line reduction processes. A post-run process transfers files not shipped during runs. Files are archived to tape and are retrievable by digraph and shot number. Automatic skimming based on most recent access, file type, shot numbers, and user-set protections maintains the files required for post-run data reduction

Mechanical engineering problems in the TFTR neutral beam system

International Nuclear Information System (INIS)

Cannon, D.D.; Bryant, E.H.; Johnson, R.L.; Kim, J.; Queen, C.C.; Schilling, G.

1975-01-01

A conceptual design of a prototype beam line for the TFTR Neutral Beam System has been developed. The basic components have been defined, cost estimates prepared, and the necessary development programs identified. Four major mechanical engineering problems, potential solutions and the required development programs are discussed

Measurements of tritium recycling and isotope exchange in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Kamperschroer, J.; Mueller, D.; Nagy, A.; Stotler, D.P.

1996-05-01

Tritium Balmer-alpha (T α ) emission, along with H α and D α is observed in the current D-T experimental campaign in TFTR. The data are a measure of the fueling of the plasma by tritium accumulated in the TFTR limiter and the spectral profile maps neutral hydrogenic velocities. T α is relatively slow to appear in tritium neutral beam heated discharges, (T α /(H α + D α + T α ) = 11% after 8 tritium-only neutral beam discharges). In contrast, the T α fraction in a sequence of six discharges fueled with tritium puff,s increased to 44%. Larger transient increases (up to 75% T α ) were observed during subsequent tritium gas puffs. Analysis of the Doppler broadened spectral profiles revealed overall agreement with the dissociation, charge exchange, sputtering and reflection velocities predicted by the neutral Monte-Carlo code DEGAS with some deficiency in the treatment of dissociation products in the 10--100 eV range

Health physics and radioactive waste considerations for the TFTR

International Nuclear Information System (INIS)

Rule, K.; Gilbert, J.; Scott, J.

1995-01-01

The Tokamak Fusion Test Reactor (TFTR) began high power fusion operations, with tritium, in November of 1993. The operational health physics program involves maintenance on activated materials and tritium contaminated systems. Survey data and findings are collected on routine and special maintenance situations ranging from work on small volume piping to large volume neutral beam systems. The results of radiological measurements are described in relation to the differentiation of elemental tritium to tritium oxide in worker's breathing zones and the associated general work area. The contamination levels, airborne radioactivity, and oil concentrations are also compared. Measurements for gamma radiation are performed to determine personnel access requirements and for comparison to activation and decay models as a planning tools. TFTR presents many unusual challenges with regard to dismantling, packaging and disposal of its components and ancillary systems. A functional time phased radioactive waste generation schedule was developed to enhance project planning. This project will be the first demonstration of the decommissioning of a tritium fueled fusion test reactor

New Monte Carlo results for the TFTR/Lithium Blanket Module system

International Nuclear Information System (INIS)

Engholm, B.A.

1985-01-01

Neutronics analysis results from Phase II of the TFTR Lithium Blanket Module (LBM) program are reported. Principal activities were analyses of new coverplate and protective plate designs; updating of the MCNP Monte Carlo model of TFTR/LBM; and performing new reference calculations for D-D and D-T plasmas. The new protective plate was found to reduce LBM responses by 20%. Updating the model included a new tally structure in which the LBM is divided into 92 volume elements corresponding to foil locations. A new version of the MCNP surface-source routine was used, along with the latest pointwise cross sections. All flux, tritium and foil responses are stored at NMFECC and are available for comparison with measurements, when the experimental program gets underway

Pneumatic pellet injectors for TFTR and JET

International Nuclear Information System (INIS)

Combs, S.K.; Milora, S.L.

1986-01-01

This paper describes the development of pneumatic hydrogen pellet injectors for plasma fueling applications on the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET). The performance parameters of these injectors represent an extension of previous experience and include pellet sizes in the range 2-6 mm in diameter and speeds approaching 2 km/s. Design features and operating characteristics of these pneumatic injectors are presented

Feasability study of using the TFTR Thomson scattering system for q profile measurements

International Nuclear Information System (INIS)

Brizard, A.; Grewk, B.; Johnson, D.; LeBlanc, B.

1986-01-01

The results of a study made to determine the possibility of using the TFTR 76 channels Thomson scattering system to measure the direction of local magnetic fields in a tokamak plasma are presented. As this is a local measurement, this technique can in principle yield q profiles without the need of any de-convolution. The effect of the TFTR geometrical configuration and its various components on the expected measurement accuracy is discussed. The authors find that the measurement of q values within the inner half of the plasma should be possible, with only minor modification to the present TVTS system

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.)

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.

Engineering analysis of ITER In-Vessel Viewing System guide tube

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Design and performance of main vacuum pumping system of SST-1 Tokamak

Energy Technology Data Exchange (ETDEWEB)

Khan, Ziauddin, E-mail: ziauddin@ipr.res.in; Pathan, Firozkhan; George, Siju; Dhanani, Kalpesh; Paravastu, Yuvakiran; Semwal, Pratibha; Pradhan, Subrata

2014-01-15

Highlights: •SST-1 Tokamak was successfully commissioned. •Vacuum vessel and cryostat were pumped down to 6.3 × 10{sup −7} mbar and 1.3 × 10{sup −5} mbar. •Leaks developed during baking were detected in-situ by RGA and confirmed later on. •Cryo-pumping effect was observed when LN2 thermal shields reached below 273 K. •Non-standard aluminum wire-seals have shown leak tightness < 1.0 × 10{sup −9} mbar l/s. -- Abstract: Steady-state Superconducting Tokamak (SST-1) was installed and it is commissioning for overall vacuum integrity, magnet systems functionality in terms of successful cool down to 4.5 K and charging up to 10 kA current was started from August 2012. Plasma operation of 100 kA current for more than 100 ms was also envisaged. It is comprised of vacuum vessel (VV) and cryostat (CST). Vacuum vessel, an ultra-high (UHV) vacuum chamber with net volume of 23 m{sup 3} was maintained at the base pressure of 6.3 × 10{sup −7} mbar for plasma confinement. Cryostat, a high-vacuum (HV) chamber with empty volume 39 m{sup 3} housing superconducting magnet system, bubble thermal shields and hydraulics for these circuits, maintained at 1.3 × 10{sup −5} mbar in order to provide suitable environment for these components. In order to achieve these ultimate vacuums, two numbers of turbo-molecular pumps (TMP) are installed in vacuum vessel while three numbers of turbo-molecular pumps are installed in cryostat. Initial pumping of both the chambers was carried out by using suitable Roots pumps. PXI based real time controlled system is used for remote operation of the complete pumping operation. In order to achieve UHV inside the vacuum vessel, it was baked at 150 °C for longer duration. Aluminum wire-seals were used for all non-circular demountable ports and a leak tightness < 1.0 × 10{sup −9} mbar l/s were achieved.

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Vacuum system design and tritium inventory for the charge exchange diagnostic on the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Medley, S.S.

1986-01-01

The application of charge exchange analyzers for the measurement of ion temperature in fusion plasma experiments requires a direct connection between the diagnostic and plasma-discharge vacuum chambers. Differential pumping of the gas load from the diagnostic stripping cell operated at > or approx. = 10 -3 Torr is required to maintain the analyzer chamber at a pressure of -6 Torr. The migration of gases between the diagnostic and plasma vacuum chambers must be minimized. In particular, introduction of the analyzer stripping cell gas into the plasma chamber having a base pressure of -8 Torr must be suppressed. The charge exchange diagnostic for the Tokamak Fusion Test Reactor (TFTR) is comprised of two analyzer systems designed to contain a total of 18 independent mass/energy analyzers and one diagnostic neutral beam rated at 80 keV, 15 A. The associated arrays of multiple, interconnected vacuum systems were analyzed using the Vacuum System Transient Simulator (Vsts) computer program which models the transient transport of multigas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced costs, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus

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

TFTR data management system

International Nuclear Information System (INIS)

Randerson, L.; Chu, J.; Ludescher, C.; Malsbury, J.; Stark, W.

1986-01-01

Developments in the tokamak fusion test reactor (TFTR) data management system supporting data management system supporting data acquisition and off-line physics data reduction are described. Data from monitor points, timing channels, and transient recorder channels and other devices are acquired and stored for use by on-line tasks. Files are transferred off-line automatically. A configuration utility determines data acquired and files transferred. An event system driven by file arrival activates off-line reduction processes. A post-run process transfers files not shipped during runs. Files are archived to tape and are retrievable by digraph and shot number. Automatic skimming based on most recent access, file type, shot numbers, and user-set protection maintains the files required for post-run data reduction

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)

Simulations of DT experiments in TFTR

International Nuclear Information System (INIS)

Budny, R.; Bell, M.G.; Biglari, H.; Bitter, M.; Bush, C.; Cheng, C.Z.; Fredrickson, E.; Grek, B.; Hill, K.W.; Hsuan, H.; Janos, A.; Jassby, D.L.; Johnson, D.; Johnson, L.C.; LeBlanc, B.; McCune, D.C.; Mikkelsen, D.R.; Park, H.; Ramsey, A.T.; Sabbagh, S.A.; Scott, S.; Schivell, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.; Taylor, G.; Zarnstorff, M.C.; Zweben, S.J.

1991-12-01

A transport code (TRANSP) is used to simulate future deuterium-tritium experiments (DT) in TFTR. The simulations are derived from 14 TFTR DD discharges, and the modeling of one supershot is discussed in detail to indicate the degree of accuracy of the TRANSP modeling. Fusion energy yields and α-particle parameters are calculated, including profiles of the α slowing down time, average energy, and of the Alfven speed and frequency. Two types of simulations are discussed. The main emphasis is on the DT equivalent, where an equal mix of D and T is substituted for the D in the initial target plasma, and for the D O in the neutral-beam injection, but the other measured beam and plasma parameters are unchanged. This simulation does not assume that α heating will enhance the plasma parameters, or that confinement will increase with T. The maximum relative fusion yield calculated for these simulations is Q DT ∼ 0.3, and the maximum α contribution to the central toroidal β is β α (0) ∼ 0.5%. The stability of toroidicity-induced Alfven eigenmodes (TAE) and kinetic ballooning modes (KBM) is discussed. The TAE mode is predicted to become unstable for some of the equivalent simulations, particularly after the termination of neutral beam injection. In the second type of simulation, empirical supershot scaling relations are used to project the performance at the maximum expected beam power. The MHD stability of the simulations is discussed

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.

Ion cyclotron emission due to collective instability of fusion products and beam ions in TFTR and JET

International Nuclear Information System (INIS)

Dendy, R.O.; Clements, K.G.; Lashmore-Davies, C.N.; Cottrell, G.A.; Majeski, R.; Cauffman, S.

1995-06-01

Ion cyclotron emission (ICE) has been observed from neutral beam-heated TFTR and JET tritium experiments at sequential cyclotron harmonics of both fusion products and beam ions. The emission originates from the outer mid-plane plasma, where fusion products and beam ions are likely to have a drifting ring-type velocity-space distribution which is anisotropic and sharply peaked. Fusion product-driven ICE in both TFTR and JET can be attributed to the magnetoacoustic cyclotron instability, which involves the excitation of obliquely propagating waves on the fast Alfven/ion Bernstein branch at cyclotron harmonics of the fusion products. Differences between ICE observations in JET and TFTR appear to reflect the sensitivity of the instability growth rate to the ratio υ birth /c A , where υ birth is the fusion product birth speed and c A is the local Alfven speed:for fusion products in the outer midplane edge of TFTR, υ birth A ; for alpha-particles in the outer midplane edge of JET, the opposite inequality applies. If sub-Alfvenic fusion products are isotropic or have undergone even a moderate degree of thermalization, the magnetoacoustic instability cannot occur. In contrast, the super-Alfvenic alpha-particles which are present in the outer mid-plane of JET can drive the magnetoacoustic cyclotron instability even if they are isotropic or have a relatively broad distribution of speeds. These conclusions may account for the observation that fusion product-driven ICE in JET persists for longer than fusion product-driven ICE in TFTR. (Author)