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Sample records for tokamak vacuum vessel

  1. 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.

  2. 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)

  3. 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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  5. 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

  6. 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.

  7. 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.

  8. 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

  9. 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.)

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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)

  17. 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.)

  18. 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)

  19. 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)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  1. 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.)

  2. 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

  3. 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.)

  4. 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

  5. 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)

  6. 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.

  7. 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

  8. 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.

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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.)

  15. Initial conditioning of the TFTR vacuum vessel

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  16. 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.)

  17. 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

  18. 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.)

  19. 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.

  20. 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.)

  1. 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.)

  2. Trajectory planning of tokamak flexible in-vessel inspection robot

    International Nuclear Information System (INIS)

    Wang, Hesheng; Chen, Weidong; Lai, Yinping; He, Tao

    2015-01-01

    Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

  3. Trajectory planning of tokamak flexible in-vessel inspection robot

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

  4. Vessel eddy current characteristics in SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

  5. 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.)

  6. 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.)

  7. 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.

  8. 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.)

  9. In-vessel remote maintenance of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref

  10. 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

  11. 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.

  12. 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

  13. 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

  14. 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)

  15. 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

  16. Experiences on vacuum conditioning in the cryostat of KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwang Pyo, E-mail: kpkim@nfri.er.ke [National Fusion Research Institute, Daejeon (Korea, Republic of); Woo, I.S.; Chang, Y.B.; Kwag, S.W.; Song, N.H.; Bang, E.N.; Hong, J.S.; Chu, Y.; Park, K.R. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    Highlights: ► The vacuum of the cryostat has been stably maintained during the KSTAR operation. ► The detected cold leak at the PF/CS coils and CS structure. ► The present helium leak makes no issue for the cryostat operation. -- Abstract: Korea Superconducting Tokamak Advanced Research (KSTAR) device has been successfully operated for the plasma experiment from KSTAR 1st campaign to 4th campaign. The main pumping system for the cryostat has to maintain the target pressure below 1.0 × 10{sup −4} mbar at room temperature and 1.0 × 10{sup −5} mbar at extremely low temperature for the plasma experiment against the air leak coming from ports of vessel and/or the helium leak from cooling loops in the cryostat. No leak has been detected at room temperature. Unexpectedly, the cold-leak appeared in the cryostat at temperature around 50 K during the cool-down in the KSTAR 2nd campaign. We carefully analyzed the characteristics of detected cold leak because it can cause the increase of the base pressure in the cryostat. After the cool-down, the leak detection was performed to locate the position and size of the leak by the pressurizing the loops. As a result, it is found that the cold leak was located at cooling loops for PF/CS coils and CS structure. Nevertheless, the vacuum inside the cryostat was well maintained below 6.0 × 10{sup −8} mbar during the entire operation period. The impact of the He-leak in present status on the plasma operation is negligible. However, we have found that the leak rate increases as a function of time. Therefore careful monitoring on cold-leak is an important technical issue for the operation of superconducting tokamak.

  17. Hand-eye coordinative remote maintenance in a tokamak vessel

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Qiang, E-mail: qiu6401@sjtu.edu.cn; Gu, Kai, E-mail: gukai0707@sjtu.edu.cn; Wang, Pengfei, E-mail: wpf790714@163.com; Bai, Weibang, E-mail: 654253204@qq.com; Cao, Qixin, E-mail: qxcao@sjtu.edu.cn

    2016-03-15

    Highlights: • If there is not rotation between the visual coordinate frame (O{sub e}X{sub e}Y{sub e}) and hand coordinate frame (O{sub h}X{sub h}Y{sub h}), a person can coordinate the movement between hand and eye easily. • We establish an alignment between the movement of the operator's hand and the visual scene of the end-effector as displayed on the monitor. • A potential function is set up in a simplified vacuum vessel model to provide a fast collision checking, and the alignment between repulsive force and Omega 7 feedback force is accomplished. • We carry out an experiment to evaluate its performance in a remote handling task. - Abstract: The reliability is vitally important for the remote maintenance in a tokamak vessel. In order to establish a more accurate and safer remote handling system, a hand-eye coordination method and an artificial potential function based collision avoidance method were proposed in this paper. At the end of this paper, these methods were implemented to a bolts tightening maintenance task, which was carried out in our 1/10 scale tokamak model. Experiment results have verified the value of the hand-eye coordination method and the collision avoidance method.

  18. Hand-eye coordinative remote maintenance in a tokamak vessel

    International Nuclear Information System (INIS)

    Qiu, Qiang; Gu, Kai; Wang, Pengfei; Bai, Weibang; Cao, Qixin

    2016-01-01

    Highlights: • If there is not rotation between the visual coordinate frame (O_eX_eY_e) and hand coordinate frame (O_hX_hY_h), a person can coordinate the movement between hand and eye easily. • We establish an alignment between the movement of the operator's hand and the visual scene of the end-effector as displayed on the monitor. • A potential function is set up in a simplified vacuum vessel model to provide a fast collision checking, and the alignment between repulsive force and Omega 7 feedback force is accomplished. • We carry out an experiment to evaluate its performance in a remote handling task. - Abstract: The reliability is vitally important for the remote maintenance in a tokamak vessel. In order to establish a more accurate and safer remote handling system, a hand-eye coordination method and an artificial potential function based collision avoidance method were proposed in this paper. At the end of this paper, these methods were implemented to a bolts tightening maintenance task, which was carried out in our 1/10 scale tokamak model. Experiment results have verified the value of the hand-eye coordination method and the collision avoidance method.

  19. 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.)

  20. 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.)

  1. 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

  2. Active cooling system for Tokamak in-vessel operation manipulator

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Jianjun, E-mail: yuanjj@sjtu.edu.cn; Chen, Tan; Li, Fashe; Zhang, Weijun; Du, Liang

    2015-10-15

    Highlights: • We summarized most of the challenges of fusion devices to robot systems. • Propose an active cooling system to protect all of the necessary components. • Trial design test and theoretical analysis were conducted. • Overall implementation of the active cooling system was demonstrated. - Abstract: In-vessel operation/inspection is an indispensable task for Tokamak experimental reactor, for a robot/manipulator is more capable in doing this than human being with more precise motion and less risk of damaging the ambient equipment. Considering the demanding conditions of Tokamak, the manipulator should be adaptable to rapid response in the extreme conditions such as high temperature, vacuum and so on. In this paper, we propose an active cooling system embedded into such manipulator. Cameras, motors, gearboxes, sensors, and other mechanical/electrical components could then be designed under ordinary conditions. The cooling system cannot only be a thermal shield since the components are also heat sources in dynamics. We carry out a trial test to verify our proposal, and analyze the active cooling system theoretically, which gives a direction on the optimization by varying design parameters, components and distribution. And based on thermal sensors monitoring and water flow adjusting a closed-loop feedback control of temperature is added to the system. With the preliminary results, we believe that the proposal gives a way to robust and inexpensive design in extreme environment. Further work will concentrate on overall implementation and evaluation of this cooling system with the whole inspection manipulator.

  3. 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.

  4. 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

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

    International Nuclear Information System (INIS)

    DeLucia, J.

    1985-12-01

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

  6. 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.)

  7. 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.

  8. 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

  9. Estimation of the radial force on the tokamak vessel wall during fast transient events

    Energy Technology Data Exchange (ETDEWEB)

    Pustovitov, V. D., E-mail: pustovitov-vd@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2016-11-15

    The radial force balance in a tokamak during fast transient events with a duration much shorter than the resistive time of the vacuum vessel wall is analyzed. The aim of the work is to analytically estimate the resulting integral radial force on the wall. In contrast to the preceding study [Plasma Phys. Rep. 41, 952 (2015)], where a similar problem was considered for thermal quench, simultaneous changes in the profiles and values of the pressure and plasma current are allowed here. Thereby, the current quench and various methods of disruption mitigation used in the existing tokamaks and considered for future applications are also covered. General formulas for the force at an arbitrary sequence or combination of events are derived, and estimates for the standard tokamak model are made. The earlier results and conclusions are confirmed, and it is shown that, in the disruption mitigation scenarios accepted for ITER, the radial forces can be as high as in uncontrolled disruptions.

  10. Vacuum system of the Tokamak Novillo

    International Nuclear Information System (INIS)

    Valencia Alvarado, R.; Lopez Callejas, R.; Melendez Lugo, L.; Chavez Alarcon, E.

    1990-01-01

    A toroidal vacuum chamber of 28 access ports was constructed from four stainless steel 316L elbows joined together with dielectric seals to provide voltage break in the toroidal direction. All vacuum seals were viton O-rings. A 500 1/s turbomolecular pump provide a base pressure of 4 x 10 -8 mbar with light backing (∼ 60 deg C). In the regime of cleaning discharge the toroidal chamber is filled with H 2 to a pressure of 0.2 + 0.4 mbar. The ohmic heating coils are pulsed with a AF oscillator (10 kW, 17.5 kHz) for 60-80 msec, at a repetition rate of about 2 Hz. A toroidal magnetic field of roughly 600 G is added to generate the cleaning discharge. According to a residual gas analyzer installed in the system, 80% of the base pressure was given by water; the other components were hydrogen and mon- and dioxides of carbon. (Author)

  11. 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)

  12. Conditioning of the vacuum chamber of the Tokamak Novillo

    International Nuclear Information System (INIS)

    Valencia A, R.; Lopez C, R.; Melendez L, L.; Chavez A, E.; Colunga S, S.; Gaytan G, E.

    1992-03-01

    The obtained experimental results of the implementation of two techniques of present time for the conditioning of the internal wall of the chamber of discharges of the Tokamak Novillo are presented, which has been designed, built and put in operation in the Laboratory of Plasma Physics of the National Institute of Nuclear Research (ININ). These techniques are: the vacuum baking and the low energy pulsed discharges, which were applied after having reached an initial pressure of the order of 10 -7 Torr. with a system of turbomolecular pumping previous preparation of surfaces and vacuum seals. The analysis of residual gases was carried out with a mass spectrometer before and after conditioning. The obtained results show that the vacuum baking it was of great effectiveness to reduce the value of the initial pressure in short time, in more of a magnitude order and the low energy discharges reduced the oxygen at worthless levels with regard to the initial values. (Author)

  13. 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

  14. In-vessel maintenance concepts for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Yount, J.A.

    1983-01-01

    Concepts for rail-mounted and guided in-vessel handling machines (IVM) for remote maintenance inside tokamak fusion reactors are described. The IVM designs are based on concepts for tethered remotely operated vehicles and feature the use of multiple manipulator arms for remote handling and remote-controlled TV cameras for remote viewing. The concepts include IVMs for both single or dual rail systems located in the top or bottom of the reactor vessel

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

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Bedakihale, Vijay; Ranganath, Tata

    2009-01-01

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

  16. Long- and short-term trends in vessel conditioning of TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    LaMarche, P.H.; Dylla, H.F.; Bell, M.G.

    1986-10-01

    We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150 0 C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long-term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short-term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area

  17. 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.

  18. 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.

  19. 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

  20. 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)

  1. 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)

  2. 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.

  3. 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.

  4. 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.)

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

    International Nuclear Information System (INIS)

    Sato, Hiroshi.

    1980-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  8. Alternatives of ITER vacuum vessel support system

    International Nuclear Information System (INIS)

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

    2002-07-01

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

  9. Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

    International Nuclear Information System (INIS)

    Wang, Hesheng; Xu, Lifei; Chen, Weidong

    2016-01-01

    In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

  10. Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

    In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

  11. 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.)

  12. 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)

  13. 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

  14. Dynamic navigation simulation of an articulated multi-link arm for in-vessel inspection tasks in a Tokamak

    International Nuclear Information System (INIS)

    Rastogi, Naveen; Prakash, Ravi; Dutta, Pramit; Virpara, Nirav

    2013-01-01

    This paper is part of the remote handling (RH) activities towards preparedness for the future fusion machines. The aim of the R and D program performed is to demonstrate the feasibility of the inspection tasks inside the Vacuum Vessel. Due to the toroidal geometry and huge dimensions of the Vacuum Vessel, there is an inevitable need of a precise and fast automated articulated inspection system that can perform the required inspection tasks without damaging the surface and to maintain the machine availability for the maximum time. When considering generic Tokamak relevant conditions, the set of major challenges for the Remote Equipment is to sustain the severe operating conditions: ultra high vacuum, temperature and tritium level. The limited number of machine access ports and the very constrained environment complicate the introduction of a robot into the machine. The Multi Link inspection arm is required to be deployed in the bounded environment inside the Tokamak Vessel. This paper presents the development of software for implementation of autonomous navigation motion control algorithms based design simulations for the inspection Arm for routine inspection tasks, navigation to the targeted coordinates inside the Vessel, clash avoidance and to perform other auxiliary mechanical tasks. The software is also capable to store the joints and frame locations at every interval which can be used for the real time control application. The developed software has the flexibility to work with any number of links and joints. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  17. Vacuum vessel of thermonuclear device and manufacturing method thereof

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  18. Tests of vacuum interrupters for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Warren, R.; Parsons, M.; Honig, E.; Lindsay, J.

    1979-04-01

    The Tokamak Fusion Test Reactor (TFTR) project at Princeton University requires the insertion of a resistor in an excited ohmic-heating coil circuit to produce a plasma initiation pulse (PIP). It is expected that the maximum duty for the switching system will be an interruption of 24 kA with an associated recovery voltage of 25 kV. Vacuum interrupters were selected as the most economical means to satisfy these requirements. However, it was felt that some testing of available systems should be performed to determine their reliability under these conditions. Two interrupter systems were tested for over 1000 interruptions each at 24 kA and 25 kV. One system employed special Westinghouse type WL-33552 interrupters in a circuit designed by LASL. This circuit used a commercially available actuator and a minimum size counterpulse bank and saturable reactor. The other used Toshiba type VGB2-D20 interrupters actuated by a Toshiba mechanism in a Toshiba circuit using a larger counterpulse bank and saturable reactor

  19. 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.)

  20. 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...

  1. Limiter/vacuum system for plasma impurity control and exhaust in tokamaks

    International Nuclear Information System (INIS)

    Abdou, M.; Brooks, J.; Mattas, R.

    1980-01-01

    A detailed design of a limiter/vacuum system for plasma impurity control and exhaust has been developed for the STARFIRE tokamak power plant. It is shown that the limiter/vacuum concept is a very attractive option for power reactors. It is relatively simple and inexpensive and deserves serious experimental verification

  2. 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.

  3. 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.)

  4. 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.)

  5. 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

  6. 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.

  7. Locking mechanism for in-vessel components of tokamak reactor

    International Nuclear Information System (INIS)

    Nishio, S.; Shimizu, K.; Koizumi, K.; Tada, E.

    1992-01-01

    The locking and unlocking mechanism for in-vessel replaceable components such as blanket modules, is one of the most critical issues of the tokamak fusion reactor, since the sufficient stiffness against the enormous electromagnetic loads and the easy replaceability are required. In this paper, the authors decide that a caulking cotter joint is worth initiating the R and D from veiwpoints of an effective use of space, a replaceability, a removability of nuclear heating, and a reliability. In this approach, the cotter driving (thrusting and plucking) mechanism is a critical technology. A flexible tube concept has been developed as the driving mechanism, where the stroke and driving force are obtained by a fat shape by the hydraulic pressure. The original normal tube is subjected to the working percentage of more than several hundreds percentage (from thickness of 1.2 mm to 0.2 mm) for plastically forming the flexible tube

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-01-01

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

  9. 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)

  10. 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.

  11. 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).

  12. 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...

  13. 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).

  14. 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

  15. Remote maintenance of Compact Ignition Tokamak ex-vessel systems

    International Nuclear Information System (INIS)

    DePew, R.E.; Macdonald, D.

    1989-01-01

    The use of deuterium-tritium (D-T) fuel in the Compact Ignition Tokamak (CIT) will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion device's auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. Throughout the CIT remote maintenance (RM) studies conducted to date, computer modeling has been used extensively to investigate manipulator access in these complex, tightly packed, and cluttered surroundings. A recent refinement of computer modeling involves the use of an intelligent engineering work station for realtime interactive display of task simulations. This paper discusses the use of three-dimensional (3-D) kinematic computer models of the CIT machines that are proving to be powerful tools in our efforts to evaluate RM requirements. This presentation includes a video-taped simulation of remote replacement of a plasma viewing assembly. The simulation illustrates some of the constraints associated with typical RM activities and the ways in which computer modeling enhances the design process. 1 ref., 3 figs

  16. 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

  17. Tokamak DEMO-FNS: Concept of magnet system and vacuum chamber

    Energy Technology Data Exchange (ETDEWEB)

    Azizov, E. A., E-mail: Azizov-EA@nrcki.ru; Ananyev, S. S. [National Research Center Kurchatov Institute (Russian Federation); Belyakov, V. A.; Bondarchuk, E. N.; Voronova, A. A. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Golikov, A. A. [National Research Center Kurchatov Institute (Russian Federation); Goncharov, P. R. [Peter the Great St. Petersburg Polytechnic University (Russian Federation); Dnestrovskij, A. Yu. [National Research Center Kurchatov Institute (Russian Federation); Zapretilina, E. R. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Ivanov, D. P. [National Research Center Kurchatov Institute (Russian Federation); Kavin, A. A.; Kedrov, I. V. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Klischenko, A. V.; Kolbasov, B. N. [National Research Center Kurchatov Institute (Russian Federation); Krasnov, S. V. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Krylov, A. I. [National Research Center Kurchatov Institute (Russian Federation); Krylov, V. A.; Kuzmin, E. G. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Kuteev, B. V. [National Research Center Kurchatov Institute (Russian Federation); Labusov, A. N. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); and others

    2016-12-15

    The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion—fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet system and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k{sub 95} = 2. The fusion power is P{sub FUS} = 40 MW. The toroidal magnetic field on the plasma-filament axis is B{sub t0} = 5 T. The plasma current is I{sub p} = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb{sub 3}Sn, NbTi and Nb{sub 3}Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron—water radiation shield (70% of stainless steel and 30% of water).

  18. Vacuum physics analysis of HT-7 superconducting tokamak pump limiter

    International Nuclear Information System (INIS)

    Hu Jiansheng; Li Chengfu; He Yexi

    1998-10-01

    The pump limiter is analysed with HT-7 superconducting tokamak parameter and the pump limiter construction. The particle exhaust of the pump limiter can be to achieve about 7.7%. So the pump limiter can be applied in the HT-7 device and will make good affection in plasma discharge

  19. Conditioning of the vacuum chamber of the Tokamak Novillo; Acondicionamiento de la camara de vacio del Tokamak Novillo

    Energy Technology Data Exchange (ETDEWEB)

    Valencia A, R; Lopez C, R; Melendez L, L; Chavez A, E; Colunga S, S; Gaytan G, E

    1992-03-15

    The obtained experimental results of the implementation of two techniques of present time for the conditioning of the internal wall of the chamber of discharges of the Tokamak Novillo are presented, which has been designed, built and put in operation in the Laboratory of Plasma Physics of the National Institute of Nuclear Research (ININ). These techniques are: the vacuum baking and the low energy pulsed discharges, which were applied after having reached an initial pressure of the order of 10{sup -7} Torr. with a system of turbomolecular pumping previous preparation of surfaces and vacuum seals. The analysis of residual gases was carried out with a mass spectrometer before and after conditioning. The obtained results show that the vacuum baking it was of great effectiveness to reduce the value of the initial pressure in short time, in more of a magnitude order and the low energy discharges reduced the oxygen at worthless levels with regard to the initial values. (Author)

  20. 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.

  1. 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)

  2. 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

  3. 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.

  4. CFD modelling of the wall friction velocity field in the ITER tokamak resulting from airflow during a loss of vacuum accident—Consequences for particle resuspension

    Energy Technology Data Exchange (ETDEWEB)

    Gélain, T., E-mail: thomas.gelain@irsn.fr; Rondeau, A.; Peillon, S.; Sabroux, J.C.; Gensdarmes, F.

    2015-11-15

    During a loss of vacuum accident (LOVA), dusts that will be present in the future tokamak ITER are likely to be re-suspended. Such aerosols formed may present a risk for explosion and airborne contamination. This article presents parameters that govern the forces affecting particles deposited on a wall and subject to airflow. It is shown the influence of three parameters in the dust mobilization mechanism, i.e.: the particles diameter, the fluid density and the friction velocity. From numerical simulations, it is determined the evolution of wall friction velocities in the vacuum vessel (VV) of ITER during a LOVA. The numerical calculations performed with ANSYS CFX code provide average friction velocities in the lower part of the tokamak between 12 m s{sup −1} at a pressure of 150 Pa, and 0.5 m s{sup −1} at a pressure of 10{sup 5} Pa.

  5. Vacuum system for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Lange, W.J.; Green, D.; Sink, D.A.

    1976-01-01

    The vacuum system for TFTR is described. Insofar as possible, conventional and ultrahigh vacuum (UHV) components and technology will be employed. Subassemblies will be prebaked in vacuum to reduce subsequent outgassing, and assembly will employ TIG welding and metal gaskets. It is not anticipated that the totally assembled torus with its numerous diagnostic appendages will be baked in situ to a high temperature, however a lower bakeout temperature (approximately 250 0 C) is under consideration. Final vacuum conditioning will be performed using discharge cleaning to obtain a specific outgassing rate of less than or = to 10 -10 Torr liter/sec cm 2 hydrogen isotopes and less than or = to 10 -12 Torr liter/sec cm 2 of other gases, and a base pressure of less than or = to 5 x 10 -8 Torr

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  14. 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

  15. 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.

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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)

  3. 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)

  4. 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.

  5. Space-resolved vacuum ultra-violet spectroscopy on T.F.R. Tokamak plasmas

    International Nuclear Information System (INIS)

    1978-01-01

    Results are reported of space-resolved vacuum-ultraviolet spectroscopy (between 100 A and 2000A) on T.F.R. Tokamak plasmas and examples are given of profiles for both heavy and light impurity ions. The experimental method and the associated uncertainties and problems are stressed. The great importance of numerical calculations in the interpretation of the impurity profiles is pointed out. (author)

  6. 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

  7. 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

  8. 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.

  9. Modular tokamak magnetic system

    International Nuclear Information System (INIS)

    Yang, T.F.

    1988-01-01

    This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-02

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

  12. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  13. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    International Nuclear Information System (INIS)

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

    1999-03-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-01-01

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

  16. Dynamic simulation of a planar flexible boom for tokamak in-vessel operations

    International Nuclear Information System (INIS)

    Ambrosino, G.; Celentano, G.; Garofalo, F.; Maisonnier, D.

    1991-01-01

    In this paper we present a dynamic model for the analysis of the vibrations of a planar articulated flexible boom to be used for tokamak in-vessel maintenance operations. The peculiarity of the mechanical structure of the boom enables us to consider separately the oscillations in the horizontal and vertical planes so that two separate models can be constructed for describing these phenomena. The results of simulations based on booms like that proposed for NET in-vessel operations are presented. (orig.)

  17. Distribution of the In-Vessel Diagnostics in ITER Tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  19. Results from ITER Vacuum Vessel Sector Manufacturing Development in Europe

    International Nuclear Information System (INIS)

    Jones, L.

    2006-01-01

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

  20. Manufacturing preparations for the European Vacuum Vessel Sector for ITER

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  1. Structural analysis of support structure for ITER vacuum vessel

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka

    2004-12-01

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

  2. Design and development of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  3. Integration test of ITER full-scale vacuum vessel sector

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. Integration test of ITER full-scale vacuum vessel sector

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  7. 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

  8. Resistive requirements for the vacuum wall of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Brooks, J.N.; Ehat, D.; Harkness, S.D.; Norem, J.; Stevens, H.; Turner, L.

    1978-01-01

    Most conceptual designs of tokamak power reactors have incorporated a ceramic insulator in the vacuum wall to make the wall electrically non-conducting. Such a material will have to be highly resistant to radiation damage at doses up to at least 10 MW-yr/m 2 while being compatible with a coolant and a first wall whose dimensions change due to thermal cycling and radiation damage. Thus there is considerable incentive to assess the consequences of eliminating the flux breaker from the design and having a conducting boundary instead. In this initial study the question of having a finite wall resistance has been examined in terms of its major implications on both the normal and abnormal operation of a tokamak reactor. This study has been conducted within the framework of the ANL-EPR-77 design although the results should provide some guidance for future reactors as well. The EPR design referred to is a 5 m major radius tokamak with an aspect ratio of 3.5, and with an equilibrium plasma current of 7.3 MA. The vacuum chamber is designed to accommodate a non-circular plasma with a height to width ratio of up to 1.65. The basic vacuum wall design is shown in Fig. 1. It is located about 0.4 M from the plasma boundary and has an irregular polygon shape made of sixteen sections, one per TF coil interval. Variations of this design having a range of resistance values have been used in the analysis

  9. Safety Analysis in Large Volume Vacuum Systems Like Tokamak: Experiments and Numerical Simulation to Analyze Vacuum Ruptures Consequences

    Directory of Open Access Journals (Sweden)

    A. Malizia

    2014-01-01

    Full Text Available The large volume vacuum systems are used in many industrial operations and research laboratories. Accidents in these systems should have a relevant economical and safety impact. A loss of vacuum accident (LOVA due to a failure of the main vacuum vessel can result in a fast pressurization of the vessel and consequent mobilization dispersion of hazardous internal material through the braches. It is clear that the influence of flow fields, consequence of accidents like LOVA, on dust resuspension is a key safety issue. In order to develop this analysis an experimental facility is been developed: STARDUST. This last facility has been used to improve the knowledge about LOVA to replicate a condition more similar to appropriate operative condition like to kamaks. By the experimental data the boundary conditions have been extrapolated to give the proper input for the 2D thermofluid-dynamics numerical simulations, developed by the commercial CFD numerical code. The benchmark of numerical simulation results with the experimental ones has been used to validate and tune the 2D thermofluid-dynamics numerical model that has been developed by the authors to replicate the LOVA conditions inside STARDUST. In present work, the facility, materials, numerical model, and relevant results will be presented.

  10. 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.

  11. 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

  12. Radial force on the vacuum chamber wall during thermal quench in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Pustovitov, V. D., E-mail: pustovitov-vd@nrcki.ru [National Research Centre Kurchatov Institute (Russian Federation)

    2015-12-15

    The radial force balance during a thermal quench in tokamaks is analyzed. As a rule, the duration τ{sub tp} of such events is much shorter than the resistive time τ{sub w} of the vacuum chamber wall. Therefore, the perturbations of the magnetic field B produced by the evolving plasma cannot penetrate the wall, which makes different the magnetic pressures on its inner and outer sides. The goal of this work is the analytical estimation of the resulting integral radial force on the wall. The plasma is considered axially symmetric; for the description of radial forces on the wall, the results of V.D. Shafranov’s classical work [J. Nucl. Energy C 5, 251 (1963)] are used. Developed for tokamaks, the standard equilibrium theory considers three interacting systems: plasma, poloidal field coils, and toroidal field coils. Here, the wall is additionally incorporated with currents driven by ∂B/∂t≠0 accompanying the fast loss of the plasma thermal energy. It is shown that they essentially affect the force redistribution, thereby leading to large loads on the wall. The estimates prove that these loads have to be accounted for in the disruptive scenarios in large tokamaks.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Sugihara, Masayoshi; Shimomura, Yasuo

    1993-01-01

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

  15. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    Im, K.H.; Cho, S.; Her, N.I.

    2001-01-01

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  16. Ex-vessel remote maintenance for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist in removing and repairing such components as diagnostic modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the plasma chamber includes a bridge-mounted manipulator system for test cell operations, decontamination (decon) equipment, hot cell equipment, and solid-radiation-waste-handling equipment. Wherever possible, the project will use commercially available equipment. Several areas of the maintenance system design were addressed in fiscal year (FY) 1987, including conceptual designs of manipulator systems, the start of a remote equipment research and development (RandD) program, and definition of the hot cell, decon, and equipment repair facility requirements. R and D work included preliminary demonstrations of remote handling operations on full-size, partial mock-ups of the CIT machine at the Oak Ridge National Laboratory (ORNL) Remote Operations and Maintenance Development (ROMD) Facility. 1 ref., 6 figs

  17. 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)

  18. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

  19. Characterization of photo-multiplier tube as ex-vessel radiation detector in tokamak

    Science.gov (United States)

    Jo, Jungmin; Cheon, MunSeong; Kim, Junghee; An, YoungHwa; Park, Seungil; Chung, Kyoung-Jae; Hwang, Y. S.

    2017-09-01

    Feasibility of using conventional photo-multiplier tubes (PMTs) without a scintillator as an ex-vessel radiation detector in a tokamak environment is studied. Basic irradiation tests using standard gamma ray sources and a d-d neutron generator showed that the PMT is responding both to gamma photons and neutrons, possibly due to the direct generation of secondary electrons inside the PMT by the impingement of high energy photons. Because of the selective sensitivity of the PMT to hard x-ray and neutrons in ohmic and neutral beam injected plasmas, respectively, it is shown that the PMT with certain configuration can be utilized either to monitor the fluctuation in the fusion neutron generation rate or to study the behavior of runaway electrons in tokamaks.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  1. Plasma transport in stochastic magnetic field caused by vacuum resonant magnetic perturbations at diverted tokamak edge

    International Nuclear Information System (INIS)

    Park, G.; Chang, C. S.; Joseph, I.; Moyer, R. A.

    2010-01-01

    A kinetic transport simulation for the first 4 ms of the vacuum resonant magnetic perturbations (RMPs) application has been performed for the first time in realistic diverted DIII-D tokamak geometry [J. Luxon, Nucl. Fusion 42, 614 (2002)], with the self-consistent evaluation of the radial electric field and the plasma rotation. It is found that, due to the kinetic effects, the stochastic parallel thermal transport is significantly reduced when compared to the standard analytic model [A. B. Rechester and M. N. Rosenbluth, Phys. Rev. Lett. 40, 38 (1978)] and the nonaxisymmetric perpendicular radial particle transport is significantly enhanced from the axisymmetric level. These trends agree with recent experimental result trends [T. E. Evans, R. A. Moyer, K. H. Burrell et al., Nat. Phys. 2, 419 (2006)]. It is also found, as a side product, that an artificial local reduction of the vacuum RMP fields in the vicinity of the magnetic separatrix can bring the kinetic simulation results to a more detailed agreement with experimental plasma profiles.

  2. 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

  3. 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.)

  4. 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.

  5. 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)

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. 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.

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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

  3. He leak detection in the presence of deuterium background in tokamak vacuum systems

    International Nuclear Information System (INIS)

    Blanchard, W.R.; Krawchuk, R.B.; Dylla, H.F.

    1982-01-01

    Helium leak detection systems for magnetic fusion devices present several unique design problems because of the large dynamic range required and the high partial pressures of D 2 encountered. We describe the design and operation of a He leak detector system for the PDX tokamak. The system consists of a differentially-pumped, low resolution, He mass spectrometer which is interfaced to the foreline of one of the torus turbomolecular pump lines. The He detector has a minimum throughput sensitivity of 10 -10 Torrxl/s, and the torus-integrated system has been designed for detection of torus leaks over the range of 10 -7 to 10 Torrxl/s. Minimum leak-rates on the 38 m 3 PDX vessel which have been quantified using this system are approx.3 x 10 -8 Torrxl/s. When PDX is operated with D 2 plasmas it is necessary to reduce the partial pressure of D 2 by a factor of 100 within the mass spectrometer to maintain this sensitivity in the presence of the torus D 2 outgassing. We have designed and incorporated a D 2 filter which employs a Zr--Al getter assembly to affect the required D 2 pressure reduction

  4. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. 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

  7. Calculations in the tokamak vacuum region: δWυ and diagnostic capabilities

    International Nuclear Information System (INIS)

    Chance, M.S.

    1993-01-01

    The calculational methods used for the magnetic energy, δW v , in the vacuum region surrounding tokamak plasmas permit a natural extension to more general toroidal configurations, and also for calculating useful ancillary quantities like the simulation of the Mirnov-loop measurements, and the eddy currents generated in the conducting materials outside the plasma discharge. The generalizations include the capabilities for topologically spherical shells, and toroidally symmetric unlinked (with respect to the plasma discharge) solid conductors. The former was motivated by spheromak-like configurations, and the latter for simulating the segmented shell of PBX-M. The collocation method and Green's second identity is used for solving ∇ 2 χ = 0 for the magnetic scalar potential. This lends itself readily to incorporating the necessary boundary conditions, and once χ is known on the surfaces, a second application of Green's identity enables the simulation of the Mirnov-loop measurements. The eddy currents on the surface of the conductors, K c , is given by n x ∇χ. This is useful for visualizing the proposed coil currents needed for stabilizing the external kink modes by feedback mechanisms, and also for estimating the stresses in the shell due to disruptions, etc. The code is modular and can be easily interfaced to various stability codes like the PEST, NOVA, GATO, etc. system of codes, or can be used in a stand-alone mode for diagnostic purposes. Details of the numerical method and results are presented. These include the efficient treatment of the singularities inherently present in the Green's function method

  8. 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

  9. 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.)

  10. Conceptual design of an in-vessel inspection robotic system for Tokamak environment

    International Nuclear Information System (INIS)

    Kumar, Prabhat; Raju, Daniel; Ranjan, Vaibhav; Patel, Prateek; Dave, Jatinkumar; Naik, Mehul

    2013-01-01

    An in-vessel inspection robotic system has been conceptualized for operation inside a tokamak vessel. The robotic system is envisaged to comprise of a robotic arm, end-effector, microcontroller and wireless communication system. The end-effector is envisaged to be a special purpose camera for in-situ inspection between plasma shots. The three-link robotic arm, designed for ITER-like environment, has 4 revolute joints- 3 providing manipulation in poloidal plane and the fourth one providing limited movement in adjacent toroidal planes. This paper provides the conceptual design of the system along with kinematic analysis of robotic arm. Solutions have been derived for forward and inverse kinematic models and the Jacobian matrix for the robotic arm linkage. In forward kinematic model, given a set of joint-link parameters, the position and orientation of end-effector are determined with respect to a reference frame. In inverse kinematic model, given the specified position and orientation of end-effector with respect to a reference frame, a set of joint variables are derived that would bring the end-effector into the required posture. Using Jacobian matrix, the relation between the end-effector velocity and the joint velocity of a manipulator is obtained i.e. given the individual joint velocity; the end-effector velocity is obtained. A CAD model has been generated using CATIA to simulate the kinematic model and carry out computational stress analysis. (author)

  11. Ex-vessel remote maintenance design for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) imposes a requirement for remote handling technology for ex-vessel maintenance operations on auxiliary machine components. These operations consist of repairing and replacing components such as diagnostic, radio-frequency (rf) heating, and fueling systems using remotely operated maintenance equipment in the test cell. In addition, ex-vessel maintenance design also includes developing hot cell facilities for equipment decontamination, repair, and solid radioactive waste handling. The test cell maintenance philosophy is markedly influenced by the neutron/gamma shield surrounding the machine that allows personal access into the test cell one day after shutdown. Hence, maintenance operations can be performed hands-on in the test cell with the shield intact and must be remotely performed when the shield is disassembled for machine access. The constricted access to the auxiliary components of the machine affect the design requirements for the maintenance equipment and impose major spatial constraints. Several major areas of the maintenance system design are being addressed in fiscal year 1987. These include conceptual design of the manipulator system, preliminary remote equipment research and development, and definition of the hot cell, decontamination, and equipment repair facility requirements. The manipulator work includes investigating transporters and viewing/lighting subsystems. 2 figs

  12. 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

  13. Development of a Laser Ablation System Kit (LASK) for Tokamak in vessel tritium and dust inventory control

    International Nuclear Information System (INIS)

    Hernandez, C.; Roche, H.; Pocheau, C.; Grisolia, C.; Gargiulo, L.; Semerok, A.; Vatry, A.; Delaporte, P.; Mercadier, L.

    2009-01-01

    During Tokamak operation, Plasma Facing Components (PFCs) are subjected to severe interaction with plasma. As a consequence and independently of the PFCs composition, materials eroded and then re-deposited in the form of layers on the surfaces, can flake and produce dusts. These fragile structures are able to trap part of the hydrogenated species (tritium for example) in vessel inventory. For safety reasons, it is mandatory to measure and to control vessel dust and tritium inventory. Up to now, laser techniques are a part of the most promising methods able to solve these ITER open issues. Of special interest are laser systems loaded on a miniature tool that can be attached to a Multi Purpose Deployer (MPD) and used for laser treatments (detritiation and other), for PFCs chemical analysis as well as for micro particles recovery of dust produced during laser ablation. Such a system (Laser Ablation System Kit: LASK) is currently under development at IRFM and the following presentation will describe the current achievements of this project and the perspectives. In this paper, we will present an innovative compact system, which, loaded on a Multi Purpose Deployer, could allow operation in a harsh environment (pressure range from atmospheric to Ultra High Vacuum and temperature up to 120 deg. C). According to the process conditions, different treatments can be performed: at low laser fluence, PFCs thermal treatment will be expected, while at high laser fluence material will be ablated allowing Dust (and T) recovery as well as chemical analysis of material. This 'in-line' chemical analysis based on Laser Induced Breakdown Spectroscopy (LIBS) enables the ablation process to be controlled and preserves the substrate integrity. The paper will be focussed on the methodology followed during the LASK development and the method used to determine a laser process window able to remove co-deposited film without damaging the bulk material and taking into account external parameter

  14. Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR

    Science.gov (United States)

    Gonzalez, Jeferson; Soto, Christian; Carvajal, Johan; Ribeiro, Celso

    2013-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

  15. 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.

  16. 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

  17. 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)

  18. 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)

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

    International Nuclear Information System (INIS)

    Driesen, G.

    1975-10-01

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

  20. Comparison between 3D eddy current patterns in tokamak in-vessel components generated by disruptions

    International Nuclear Information System (INIS)

    Sakellaris, J.; Crutzen, Y.

    1996-01-01

    During plasma disruption events in Tokamaks, a large amount of magnetic energy is associated to the transfer of plasma current into eddy currents in the passive structures. In the ITER program two design concepts have been proposed. One approach (ITER CDA design) is based on copper stabilization loops (i.e., twin loops) attached to box-shaped blanket segments, electrically and mechanically separated along the toroidal direction. For another design concept (ITER EDA design) based on lower plasma elongation there is no need for specific stabilization loops. The passive stabilization is obtained by toroidally continuous components (i.e., the plasma facing wall of the blanket segments allows a continuity along the toroidal direction). Consequently, toroidal currents flow, when electromagnetic transients occur. Electromagnetic loads appear in the blanket structures in case of plasma disruptions and/or vertical displacement events either for the ITER CDA design concept or for the ITER EDA design concept. In this paper the influence of the in-vessel design configuration concepts--insulated segments or electrically continuous structures--in terms of magnetic shielding and electric insulation on the magnitude and the flow pattern of the eddy currents is investigated. This investigation will allow a performance evaluation of the two proposed design concepts

  1. The development of an in-vessel cryopump system for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Schaubel, K.M.; Baxi, C.B.; Campbell, G.L.; Laughon, G.J.; Mahdavi, M.A.; Makariou, C.C.; Smith, J.P.; Schaffer, M.J.; Menon, M.M.

    1993-07-01

    The design, testing and initial operation of the DIII-D advanced divertor cryocondensation pumping system is presented. The pump resides inside the tokamak plasma containment vessel where it provides particle exhaust pumping, and it is subjected to Joule heating and hot particle heat loads during each 10 second discharge. In addition, the pump must withstand plasma disruption induced electromagnetic forces and 400 degrees C bake-out temperatures. Cooling is accomplished by forced flow liquid helium with the two-phase helium exhaust passing through a reliquefier for thermal efficiency. A prototype pump was constructed to study surface temperature rise as a function of flow geometry, applied heat load, helium mass flow rate, and pump outlet conditions. Prototype testing led to the development of a special geometry which was demonstrated to enhance two-phase flow stability and overall heat transfer. During initial operation, deuterium pumping speeds of 32,000 L/s at 2 mTorr pressure were achieved with a helium flow rate of 5 g/s. This speed was maintained during 300 W, 8 s long test beat pulses which meets operational goals

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma

  9. 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)

  10. 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.

  11. 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

  12. 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)

  13. 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.

  14. 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.)

  15. 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

  16. 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)

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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.)

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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].

  18. Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Minato, Akio; Tone, Tatsuzo

    1985-12-01

    A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

  19. 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...

  20. 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

  1. Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

    International Nuclear Information System (INIS)

    Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka

    2014-03-01

    The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

  2. Vacuum system and cleaning techniques in the FTU machines

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  3. Tritium handling and vacuum considerations for the STARFIRE commercial tokamak reactor

    International Nuclear Information System (INIS)

    Finn, P.A.; Clemmer, R.G.; Maroni, V.A.; Dillow, C.

    1979-01-01

    Tritium processing and vacuum pumping requirements were analyzed for the STARFIRE commercial fusion reactor design. It was found that vacuum pumps having a helium capture probability of 0.5 (total helium pump speed 1.2 x 10 4 m 3 /s) in combination with the proposed STARFIRE limiter-vacuum concept is sufficient to achieve plasma impurity control and, simultaneously, high fractional burnup (11%). The high fractional burnup and minimum fuel recycle time result in a very low fuel cycle tritium inventory, approx. 1300 g. A Lean-T burn method that can further reduce the fuel cycle inventory by 30 to 50% is discussed. D 2 O is proposed as a first wall coolant from considerations of plasma contamination (due to hydrogen isotope permeation through coolant tubes) and enrichment of recycled tritium from the coolant circuit. Tritium recovery from solid breeders, under realistic structural and breeder materials constraints, appears to represent a formidable task. The tritium inventory in the solid breeder is estimated to be as high as 10 kg, which would make the blanket the largest single hold-up point for tritium in the plant

  4. 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

  5. EAST-AIA deployment under vacuum: Calibration of laser diagnostic system using computer vision

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yang, E-mail: yangyang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, 350 Shushanhu Rd, Hefei, Anhui (China); Song, Yuntao; Cheng, Yong; Feng, Hansheng; Wu, Zhenwei; Li, Yingying; Sun, Yongjun; Zheng, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, 350 Shushanhu Rd, Hefei, Anhui (China); Bruno, Vincent; Eric, Villedieu [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France)

    2016-11-15

    Highlights: • The first deployment of the EAST articulated inspection arm robot under vacuum is presented. • A computer vision based approach to measure the laser spot displacement is proposed. • An experiment on the real EAST tokamak is performed to validate the proposed measure approach, and the results shows that the measurement accuracy satisfies the requirement. - Abstract: For the operation of EAST tokamak, it is crucial to ensure that all the diagnostic systems are in the good condition in order to reflect the plasma status properly. However, most of the diagnostic systems are mounted inside the tokamak vacuum vessel, which makes them extremely difficult to maintain under high vacuum condition during the tokamak operation. Thanks to a system called EAST articulated inspection arm robot (EAST-AIA), the examination of these in-vessel diagnostic systems can be performed by an embedded camera carried by the robot. In this paper, a computer vision algorithm has been developed to calibrate a laser diagnostic system with the help of a monocular camera at the robot end. In order to estimate the displacement of the laser diagnostic system with respect to the vacuum vessel, several visual markers were attached to the inner wall. This experiment was conducted both on the EAST vacuum vessel mock-up and the real EAST tokamak under vacuum condition. As a result, the accuracy of the displacement measurement was within 3 mm under the current camera resolution, which satisfied the laser diagnostic system calibration.

  6. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  12. 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)

  13. 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.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  17. 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

  18. 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

  19. User's manual for DSTAR MOD1: A comprehensive tokamak disruption code

    International Nuclear Information System (INIS)

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

    1986-01-01

    A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces that can occur during major tokamak plasma disruptions. The DSTAR code development effort has been accomplished by coupling a recently developed free boundary tokamak plasma transport computational model with other models developed to predict impurity transport and radiation, and the electromagnetic and thermal dynamic response of vacuum vessel components. The combined model, DSTAR, is a unique tool for predicting the consequences of tokamak disruptions. This informal report discusses the sequence of events of a resistive disruption, models developed to predict plasma transport and electromagnetic field evolution, the growth of the stochastic region of the plasma, the transport and nonequilibrium ionization/emitted radiation of the ablated vacuum vessel material, the vacuum vessel thermal and magnetic response, and user input and code output

  20. Experimental measurement of magnetic field null in the vacuum chamber of KTM tokamak based on matrix of 2D Hall sensors

    Energy Technology Data Exchange (ETDEWEB)

    Shapovalov, G.; Chektybayev, B., E-mail: chektybaev@nnc.kz; Sadykov, A.; Skakov, M.; Kupishev, E.

    2016-11-15

    Experimental technique of measurement of magnetic field null region inside of the KTM tokamak vacuum chamber has been developed. Square matrix of 36 2D Hall sensors, which used in the technique, allows carrying out direct measurements of poloidal magnetic field dynamics in the vacuum chamber. To better measuring accuracy, Hall sensor’s matrix was calibrated with commercial Helmholtz coils and in situ measurement of defined magnetic field from poloidal and toroidal coils. Standard KTM Data-Acquisition System has been used to collect data from Hall sensors. Experimental results of measurement of magnetic field null in the vacuum chamber of KTM are shown in the paper. Additionally results of the magnetic field null reconstruction from signals of inductive total flux loops are shown in the paper.

  1. 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

  2. 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)

  3. 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)

  4. Particle exhaust scheme using an in-vessel cryocondensation pump in the advanced divertor configuration of the DIII-D tokamak

    International Nuclear Information System (INIS)

    Menon, M.M.; Mioduszewski, P.K.; Owen, L.W.; Anderson, P.M.; Baxi, C.B.; Langhorn, A.; Luxon, J.L.; Mahdavi, M.A.; Schaffer, M.J.; Schaubel, K.M.; "" class="author-name" title=" (General Atomics Co., San Diego, CA (United States))" data-affiliation=" (General Atomics Co., San Diego, CA (United States))" >Smith, J.P>

    1992-01-01

    In this paper, a particle exhaust scheme using a cryocondensation pump in the advanced divertor configuration of the DIII-D tokamak is described. In this configuration, the pump is located inside a baffle chamber within the tokamak, designed to receive particles reflected off the divertor strike region. A concentric coaxial loop with forced-convection flow of two-phase helium is selected as the cryocondensation surface. The pumping configuration is optimized by Monte Carlo techniques to provide maximum exhaust efficiency while minimizing the deleterious effects of impingement of energetic plasma particles on cryogenic surfaces. Heat loading contributions from various sources on the cryogenic surfaces are estimated, based on which the cryogenic surfaces are estimated, based on which the cryogenic flow loop for the pump is designed. The mechanical aspects of the pump, designed to meet the many challenging requirements of operating the cryopump internal to the tokamak vacuum and in close proximity with the high-temperature plasma, are also outlined

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-09-01

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

  10. 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)

  11. 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

  12. 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

  13. 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.

  14. 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.

  15. Current drive by spheromak injection into a tokamak

    International Nuclear Information System (INIS)

    Brown, M.R.; Bellan, P.M.

    1990-01-01

    The authors report the first observation of current drive by spheromak injection into a tokamak due to the process of helicity injection. Current drive is observed in Caltech's ENCORE tokamak (30% increase, ΔI > 1 kA) only when both the tokamak and injected spheromak have the same sign of helicity (where helicity is defined as positive if current flows parallel to magnetic field lines and negative if anti-parallel). The initial increase (decrease) in current is accompanied by a sharp decrease (increase) in loop voltage and the increase in tokamak helicity is consistent with the helicity content of the injected spheromak. In addition, the injection of the spheromak raises the tokamak central density by a factor of six. The introduction of cold spheromak plasma causes sudden cooling of the tokamak discharge from 12 eV to 4 eV which results in a gradual decline in tokamak plasma current by a factor of three. In a second experiment, the authors inject spheromaks into the magnetized toroidal vacuum vessel (with no tokamak plasma). An m = 1 magnetic structure forms in the vessel after the spheromak undergoes a double tilt; once in the cylindrical entrance between gun and tokamak, then again in the tokamak vessel. A horizontal shift of the spheromak equilibrium is observed in the direction opposite that of the static toroidal field. In the absence of net toroidal flux, the structure develops a helical pitch as predicted by theory. Experiments with a number of refractory metal coatings have shown that tungsten and chrome coatings provide some improvement in spheromak parameters. They have also designed and will soon construct a larger, higher current spheromak gun with a new accelerator section for injection experiments on the Phaedrus-T tokamak

  16. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

    Aymar, R.; Bareyt, B.; Bon Mardion, G.

    1980-10-01

    This document describes the basic design for the main components of the Tokamak system of Tora Supra. As such, it focuses on the engineering problems, and refers to last year report on Tora Supra (EUR-CEA-1021) for objectives and experimental programme of the apparatus on one hand, and for qualifying tests of the main technical solutions on the other hand. Superconducting toroidal field coil system, vacuum vessels and radiation shields, poloidal field system and cryogenic system are described

  17. Development and testing of bumper limiter of aluminum alloy vacuum vessel for reacting plasma experiment

    International Nuclear Information System (INIS)

    Uchikawa, T.; Fujiwara, M.; Ioki, K.; Irie, T.; Nayama, R.; Nishikawa, M.; Onozuka, M.; Tomita, M.

    1985-01-01

    Two types of graphite bumper limiters were designed and trially fabricated for a reacting plasma device, R-tokamak. High heat load tests were conducted to examine thermal behavior and thermal shock resistance of the limiters by using a 100kW electron beam facility. The experimental data were compared with the results of 3-D thermal analysis

  18. Development and testing of bumper limiter of aluminum alloy vacuum vessel for reacting plasma experiment

    Energy Technology Data Exchange (ETDEWEB)

    Uchikawa, T.; Fujiwara, M.; Ioki, K.; Irie, T.; Nayama, R.; Nishikawa, M.; Onozuka, M.; Tomita, M.

    1985-07-01

    Two types of graphite bumper limiters were designed and trially fabricated for a reacting plasma device, R-tokamak. High heat load tests were conducted to examine thermal behavior and thermal shock resistance of the limiters by using a 100kW electron beam facility. The experimental data were compared with the results of 3-D thermal analysis.

  19. Detailed Design and Fabrication Method of the ITER Vacuum Vessel Ports

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Rozov, V.

    2006-01-01

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

  1. ITER diagnostic system: Vacuum interface

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  2. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  3. Ex-vessel remote maintenance design for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) imposes a requirement for remote handling technology to carry out maintenance operations on auxiliary machine components. These operations consist of removing and repairing components such as diagnostics and radio frequency (rf) heating modules using remotely operated maintenance equipment. The major equipment that is being developed to accomplish maintenance external to the plasma chamber includes the bridge-mounted manipulator system for test cell operations, decontamination (decon) equipment, hot cell equipment, and solid rad-waste handling equipment. Wherever possible, the project will use commercially available equipment. Several areas of the maintenance system design have been addressed in fiscal year (FY) 1987. These included conceptual designs of manipulator systems, the start of a remote equipment research and development (R and D) program, and definition of the hot cell, decon, and equipment repair facility requirements. The manipulator work included investigating transporters and viewing/lighting subsystems. In each case, existing commercial units are being assessed initially, along with viable alternative approaches. R and D work also included demonstrations of remote handling operations on full-size, partial mock-ups of the CIT machine at the Oak Ridge National Laboratory (ORNL) Remote Operations and Maintenance Development Facility

  4. Control, pressure perturbations, displacements, and disruptions in highly elongated tokamak plasmas

    International Nuclear Information System (INIS)

    Marcus, F.B.; Hofmann, F.; Tonetti, G.; Jardin, S.C.; Noll, P.

    1989-06-01

    The control and evolution of highly elongated tokamak plasmas with large growth rates are simulated with the axisymmetric, resistive MHD code TSC in the geometry of the TCV tokamak. Pressure perturbations such as sawteeth and externally programmed displacements create initial velocity perturbations which may be stabilized by low power, rapid response coils inside the passively stabilizing vacuum vessel, together with slower shaping coils outside the vessel. Vertical disruption induced voltages and forces on the rapid coils and vessel are investigated, and a model is proposed for an additional vertical force due to poloidal currents. (author) 6 figs., 1 tab., 26 refs

  5. 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

  6. 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)

  7. An efficient modeling of fine air-gaps in tokamak in-vessel components for electromagnetic analyses

    International Nuclear Information System (INIS)

    Oh, Dong Keun; Pak, Sunil; Jhang, Hogun

    2012-01-01

    Highlights: ► A simple and efficient modeling technique is introduced to avoid undesirable massive air mesh which is usually encountered at the modeling of fine structures in tokamak in-vessel component. ► This modeling method is based on the decoupled nodes at the boundary element mocking the air gaps. ► We demonstrated the viability and efficacy, comparing this method with brute force modeling of air-gaps and effective resistivity approximation instead of detail modeling. ► Application of the method to the ITER machine was successfully carried out without sacrificing computational resources and speed. - Abstract: A simple and efficient modeling technique is presented for a proper analysis of complicated eddy current flows in conducting structures with fine air gaps. It is based on the idea of replacing a slit with the decoupled boundary of finite elements. The viability and efficacy of the technique is demonstrated in a simple problem. Application of the method to electromagnetic load analyses during plasma disruptions in ITER has been successfully carried out without sacrificing computational resources and speed. This shows the proposed method is applicable to a practical system with complicated geometrical structures.

  8. In-vessel maintenance remote manipulator system

    International Nuclear Information System (INIS)

    Jimenez, E.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Suzuki, Satoshi; Akiba, Masato; Saito, Masakatsu

    2006-01-01

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

  10. 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)

  11. Spherical tokamak power plant design issues

    International Nuclear Information System (INIS)

    Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.

    2000-01-01

    The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed

  12. Time - resolved thermography at Tokamak T-10

    International Nuclear Information System (INIS)

    Grunow, C.; Guenther, K.; Lingertat, J.; Chicherov, V.M.; Evstigneev, S.A.; Zvonkov, S.N.

    1987-01-01

    Thermographic experiments were performed at T-10 tokamak to investigate the thermic coupling of plasma and the limiter. The limiter is an internal equipment of the vacuum vessel of tokamak-type fusion devices and the interaction of plasma with limiter results a high thermal load of limiter for short time. In according to improve the limiter design the temperature distribution on the limiter surface was measured by a time-resolved thermographic method. Typical isotherms and temperature increment curves are presented. This measurement can be used as a systematic plasma diagnostic method because the limiter is installed in the tokamak whereas special additional probes often disturb the plasma discharge. (D.Gy.) 3 refs.; 7 figs

  13. Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

    International Nuclear Information System (INIS)

    Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.

    2015-01-01

    Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

  14. Vessel thermal map real-time system for the JET tokamak

    Directory of Open Access Journals (Sweden)

    D. Alves

    2012-05-01

    Full Text Available The installation of international thermonuclear experimental reactor-relevant materials for the plasma facing components (PFCs in the Joint European Torus (JET is expected to have a strong impact on the operation and protection of the experiment. In particular, the use of all-beryllium tiles, which deteriorate at a substantially lower temperature than the formerly installed carbon fiber composite tiles, imposes strict thermal restrictions on the PFCs during operation. Prompt and precise responses are therefore required whenever anomalous temperatures are detected. The new vessel thermal map real-time application collects the temperature measurements provided by dedicated pyrometers and infrared cameras, groups them according to spatial location and probable offending heat source, and raises alarms that will trigger appropriate protective responses. In the context of the JET global scheme for the protection of the new wall, the system is required to run on a 10 ms cycle communicating with other systems through the real-time data network. In order to meet these requirements a commercial off-the-shelf solution has been adopted based on standard x86 multicore technology. Linux and the multithreaded application real-time executor (MARTe software framework were respectively the operating system of choice and the real-time framework used to build the application. This paper presents an overview of the system with particular technical focus on the configuration of its real-time capability and the benefits of the modular development approach and advanced tools provided by the MARTe framework.

  15. 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

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

    International Nuclear Information System (INIS)

    Vepa, K.; Sterbentz, W.H.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  18. 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

  19. Design and construction of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, G.S.

    2001-01-01

    The extensive design effort has been focused on two major aspects of the KSTAR project mission, steady-state operation capability and 'advanced tokamak' physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long-pulse current-drive and heating systems. The 'advanced tokamak' aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizers, internal control coils , and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR experimental facility with cryogenic system and de-ionized water-cooling and main power systems has been designed, and the construction work has been on-going for completion in year 2004. (author)

  20. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-12-15

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  5. Merging startup experiments on the UTST spherical tokamak

    International Nuclear Information System (INIS)

    Yamada, Takuma; Kamio, Shuji; Imazawa, Ryota

    2010-01-01

    The University of Tokyo Spherical Tokamak (UTST) was constructed to explore the formation of ultrahigh-beta spherical tokamak (ST) plasmas using double null plasma merging. The main feature of the UTST is that the poloidal field coils are located outside the vacuum vessel to demonstrate startup in a reactor-relevant situation. Initial operations used partially completed power supplies to investigate the appropriate conditions for plasma merging. The plasma current of the merged ST reached 100 kA when the central solenoid coil was used to assist plasma formation. Merging of two ST plasmas through magnetic reconnection was successfully observed using two-dimensional pickup coil arrays, which directly measure the toroidal and axial magnetic fields inside the UTST vacuum vessel. The resistivity of the current sheet was found to be anomalously high during merging. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2010-06-01

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

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

    CERN Multimedia

    CERN. Geneva

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Shibanuma, Kiyoshi

    2007-05-01

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

  9. Development of high field superconducting Tokamak 'TRIAM-1M'

    International Nuclear Information System (INIS)

    Ito, Satoshi; Suzuki, Takao; Suzuki, Shohei; Nishi, Masatsugu; Kawasaki, Takahide.

    1984-01-01

    The tokamak nuclear fusion apparatus ''TRIAM-1M'' which is constructed in the Research Institute for Applied Mechanics, Kyushu University, has a number of distinctive features as compared with other tokamak projects, that is, the toroidal field coils are made of superconductors for the first time in Japan, and the apparatus is small and has strong magnetic field. Hitachi Ltd. designed and has forwarded the manufacture of the TRIAM-1M. In this paper, the total constitution of the apparatus and the design and manufacture of the plasma vacuum vessel, superconducting toroidal coils and others are reported. The objectives of research are the containment of strong field tokamak plasma and the establishment of the law of proportion, the development of turbulent flow heating method, the adoption of mixed wave current driving method and the practical use of Nb 3 Sn superconducting coils. The apparatus is composed of the vacuum vessel containing plasma, toroidal field coils, poloidal field coils, current transformer coils and turbulent flow heating coils for plasma heating, heat insulating vacuum vessel and supporting structures. The evacuating facility, helium liquefying refrigerator and cooling water facility are installed around the main body. (Kako, I.)

  10. ITER tokamak device

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  15. Design of the ITER Tokamak Assembly Tools

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  16. Design of the ITER tokamak assembly tools

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyunki [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)], E-mail: hkpark@nfri.re.kr; Lee, Jaehyuk; Kim, Taehyung [SFA Engineering Corp., 42-7 Palyong-dong, Changwon-si, Gyeongsangnam-do 641-847 (Korea, Republic of); Song, Yunju [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of); Im, Kihak [ITER Organization, CEA Cadarasche, 13108 Saint Paul-lez-Durance (France); Kim, Byungchul; Lee, Hyeongon; Jung, Ki-Jung [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)

    2008-12-15

    ITER tokamak assembly is mainly composed of lower cryostat activities, sector sub-assembly, sector assembly, in-vessel activities and ex-vessel activities. The main tools for sector sub-assembly procedures consists of upending tool, sector lifting tool, vacuum vessel support and bracing tool and sector sub-assembly tool. Conceptual design of assembly tools for sector sub-assembly procedures is described herein. The basic structure for upending tool has been developed under the assumption that upending is performed with crane which will be installed in Tokamak building. Sector lifting tool is designed to adjust the position of a sector to minimize the difference between the center of the tokamak building crane and the center of gravity of the sector. Sector sub-assembly tool is composed of special frame for the fine adjustment of position control with 6 degrees of freedom. The design of VV support and bracing tool for four kinds of VV 40 deg. sectors has been developed. Also, structural analysis for upending tool, sector sub-assembly tool has been studied using ANSYS for the situation of an applied load with the same dead weight multiplied by 3/4. The results of structural analyses for these tools were below the allowable values.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

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

    Science.gov (United States)

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

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  2. Numerical simulation and analysis for the baking out system of the HT-7U super-conducting tokamak device

    International Nuclear Information System (INIS)

    Song Yuntao

    2004-01-01

    It can provide an ultrahigh vacuum location for the plasma operation. In order to improve its vacuum degree and attain a high quality operation environment for plasma, it is very important to proceed 250 degree C baking out to clear the wall before the plasma operation. The paper firstly gives two kinds of structures for the baking of the vacuum vessel, in which one is the baking by electricity and another is baking by the nitrogen gas. Secondly based on the numerical simulation and analysis, some results have been attained such as the baking power, temperature field distribution and thermal stress for the vacuum vessel, which can provide some valuable theory basis for the engineering design and optimization of the baking system of the HT-7U vacuum vessel or other similar super-conducting tokamak devices

  3. Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

    Science.gov (United States)

    Pathan, F. S.; Khan, Z.; Semwal, P.; Raval, D. C.; Joshi, K. S.; Thankey, P. L.; Dhanani, K. R.

    2008-05-01

    Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN2 panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN2 panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN2 panels during sniffer test and pressure drop test using helium gas.

  4. Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pathan, F S; Khan, Z; Semwal, P; Raval, D C; Joshi, K S; Thankey, P L; Dhanani, K R [Institute for Plasma Research, Bhat, Gandhinagar - 382 428, Gujarat (India)], E-mail: firose@ipr.res.in

    2008-05-01

    Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN{sub 2} panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN{sub 2} panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN{sub 2} panels during sniffer test and pressure drop test using helium gas.

  5. Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

    International Nuclear Information System (INIS)

    Pathan, F S; Khan, Z; Semwal, P; Raval, D C; Joshi, K S; Thankey, P L; Dhanani, K R

    2008-01-01

    Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN 2 panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN 2 panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN 2 panels during sniffer test and pressure drop test using helium gas

  6. Thermal analysis of a coaxial helium panel of a cryogenic vacuum pump for advanced divertor of DIII-D tokamak

    International Nuclear Information System (INIS)

    Baxi, C.B.; Langhorn, A.; Schaubel, K.; Smith, J.

    1991-08-01

    It is planned to install a 50,000 1/s cryogenic pump for particle removal in the D3-D tokamak. A critical component of this cryogenic pump will be a helium panel which has to be maintained at a liquid helium temperature. The outer surface area of the helium panel has an area of 1 m 2 and consists of a 2.5 cm diameter, 10 m long tube. From design considerations, a coaxial geometry is preferable since it requires a minimum number of welds. However, the coaxial geometry also results in a counter flow heat exchanger arrangement, where the outgoing warm fluid will exchange heat with incoming cold fluid. This is of concern since the helium panel must be cooled from liquid nitrogen temperature to liquid helium temperature in less than 5 minutes for successful operation of the cryogenic pump. In order to analyze the thermal performance of the coaxial helium panel, a finite difference computer model of the geometry was prepared. The governing equations took into account axial as well as radial conduction through the tube walls. The variation of thermal properties was modeled. The results of the analysis showed that although the coaxial geometry behaves like a counter flow heat exchanger, within the operating range of the cryogenic pump a rapid cooldown of the helium panel from liquid nitrogen temperature to the operating temperature is feasible. A prototypical experiment was also performed at General Atomics (GA) which verified the concept and the analysis. 4 refs., 8 figs

  7. Measurement of the hydrogen recombination coefficient in the TEXT tokamak as a function of outgassing and power radiated during tokamak discharges

    International Nuclear Information System (INIS)

    Langley, R.A.; Rowan, W.L.; Bravenec, R.V.; Nelin, K.

    1986-10-01

    The global recombination rate coefficient k/sub r/ for hydrogen has been measured in the TEXT tokamak vacuum vessel for various surface conditions. An attempt was made to correlate the measured values of k/sub r/ with residual gas analyzer (RGA) data taken before each measurement of k/sub r/ and with the power radiated during tokamak discharges produced after each measurement of k/sub r/. The results show that k/sub r/ increases during a series of tokamak discharges, k/sub r/ is relatively insensitive to power radiated during tokamak discharges, and k/sub r/ increases with the RGA measurements of mass 28 and 40 but not with those of mass 18. In addition, it was found that the mass 18 (H 2 O) signal decreases as glow discharge experiments with hydrogen were performed

  8. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  9. Plasma facing components design of KT-2 tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

    1997-04-01

    The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

  10. 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

  11. 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.

  12. Assembly study for JT-60SA tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Shibanuma, K., E-mail: shibanuma.kiyoshi@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Arai, T.; Hasegawa, K.; Hoshi, R.; Kamiya, K.; Kawashima, H.; Kubo, H.; Masaki, K.; Saeki, H.; Sakurai, S.; Sakata, S.; Sakasai, A.; Sawai, H.; Shibama, Y.K.; Tsuchiya, K.; Tsukao, N.; Yagyu, J.; Yoshida, K.; Kamada, Y. [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Mizumaki, S. [Toshiba Corporation, Minato-ku, Tokyo 105-8001 (Japan); and others

    2013-10-15

    The assembly scenarios and assembly tools of the major tokamak components for JT-60SA are studied in the following. (1) The assembly frame (with a dedicated 30-tonne crane), which is located around the JT-60SA tokamak, is adopted for effective assembly works in the torus hall and the temporary support of the components during assembly. (2) Metrology for precise positioning of the components is also studied by defining the metrology points on the components. (3) The sector segmentation for weld joints and positioning of the vacuum vessel (VV), the assembly scenario and tools for VV thermal shield (TS), the connection of the outer intercoil structure (OIS) and the installation of the final toroidal field coil (TFC) are studied, as typical examples of the assembly scenarios and tools for JT-60SA.

  13. Tokamak Fusion Core Experiment maintenance study

    International Nuclear Information System (INIS)

    Snyder, A.M.; Watts, K.D.

    1985-01-01

    The recently completed Tokamak Fusion Core Experiment (TFCX) design project was carried out to investigate potential next generation tokamak concepts. An important aspect of this project was the early development and incorporation of remote maintainability throughout the design process. This early coordination and incorporation of maintenance aspects to the design of the device and facilities would assure that the machine could ultimately be maintained and repaired in an efficient and cost effective manner. To meet this end, a rigorously formatted engineering trade study was performed to determine the preferred configuration for the TFCX reactor based primarily on maintenance requirements. The study indicated that the preferred design was one with an external vacuum vessel and torrodial field coils that could be removed via a simple radial motion. The trade study is presented and the preferred TFCX configuration is described

  14. 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.

  15. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2005-01-01

    SST-1, a steady state superconducting tokamak, is undergoing commissioning tests at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in a tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. The auxiliary current drive is based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. Detailed commissioning tests on the cryogenic system and experiments on the hydraulic characters and cool down features of single TF coils have been completed prior to the cool down of the entire superconducting system. Results of the single TF magnet cool down, and testing of the magnet system are presented. First experiments related to the breakdown and the current ramp up will subsequently be carried out. (author)

  16. Reconstruction of vacuum magnetic flux in QUEST

    International Nuclear Information System (INIS)

    Ishiguro, Masaki; Hanada, Kazuaki; Nakamura, Kazuo

    2010-01-01

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

  17. Theory of nonaxisymmetric vertical displacement events in tokamaks

    International Nuclear Information System (INIS)

    Fitzpatrick, R.

    2011-01-01

    A semi-analytic sharp-boundary model of a nonaxisymmetric vertical displacement event (VDE) in a large aspect-ratio, high-beta (i.e. β ∼ ε), vertically elongated tokamak plasma is developed. The model is used to simulate nonaxisymmetric VDEs with a wide range of different plasma equilibrium and vacuum vessel parameters. These simulations yield poloidal halo current fractions and toroidal peaking factors whose magnitudes are similar to those seen in experiments, and also reproduce the characteristic inverse scaling between the halo current fraction and the toroidal peaking factor. Moreover, the peak poloidal halo current density in the vacuum vessel is found to correlate strongly with the reciprocal of the minimum edge safety factor attained during the VDE. In addition, under certain circumstances, the ratio of the net sideways force acting on the vacuum vessel to the net vertical force is observed to approach unity. Finally, the peak vertical force per unit area acting on the vessel is found to have a strong correlation with the equilibrium toroidal plasma current at the start of the VDE, but is also found to increase with increasing vacuum vessel resistivity relative to the scrape-off layer plasma.

  18. Characteristics of current quenches during disruptions in the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Zhang, Y; Chen, Z Y; Fang, D; Jin, W; Huang, Y H; Wang, Z J; Yang, Z J; Chen, Z P; Ding, Y H; Zhang, M; Zhuang, G

    2012-01-01

    Characteristics of tokamak current quenches are an important issue for the determination of electro-magnetic forces that act on the in-vessel components and vacuum vessel during major disruptions. The characteristics of current quenches in spontaneous disruptions in the J-TEXT tokamak have been investigated. It is shown that the waveforms for the fastest current quenches are more accurately fitted by linear current decays than exponential, although neither is a good fit in many slower cases. The minimum current quench time is about 2.4 ms for the J-TEXT tokamak. The maximum instantaneous current quench rate is more than seven times the average current quench rate in J-TEXT. (paper)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  1. Development and Operational Experiences of the JT-60U Tokamak and Power Supplies

    International Nuclear Information System (INIS)

    Hosogane, N.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Neyatani, Y.; Horiike, H.; Sakurai, S.; Masaki, K.; Yamamoto, M.; Kodama, K.; Sasajima, T.; Terakado, T.; Ohmori, S.; Ohmori, Y.; Okano, J.

    2002-01-01

    The design of the JT-60U tokamak, the configuration of the coil power supplies, and the operational experiences gained to date are reviewed. JT-60U is a large tokamak upgraded from the original JT-60 in order to obtain high plasma current, large plasma volume, and highly elongated divertor configurations. All components inside the toroidal magnetic field coils, such as vacuum vessel, poloidal magnetic field coils, divertor, etc., were modified. Various technologies and ideas were introduced to develop these components; for example, a multi-arc double skin wall structure for the vacuum vessel and a functional poloidal magnetic field coil system with taps for obtaining various plasma configurations. Furthermore, boron-carbide coated carbon fiber composite (CFC) tiles were used as divertor tiles to reduce erosion of carbon-base tiles. Later, a semiclosed divertor with pumps, for which cryo-panels originally used for NBI units were converted, was installed in the replacement of the open divertor. These development and operational results provide data for future tokamaks. Major failures experienced in the long operational period of JT-60U, such as water leakage from the toroidal magnetic field coil, fracture of carbon tiles, and breakdown of a filter capacitor, are described. As a maintenance issue for tokamaks using deuterium fueling gas, a method for reducing radiation exposure of in-vessel workers is described

  2. Tokamak Fusion Core Experiment (TFCX) special-purpose remote maintenance systems

    International Nuclear Information System (INIS)

    Masson, L.S.; Welland, H.J.

    1985-01-01

    A key element in the preconceptual design of the Tokamak Fusion Core Experiment (TFCX) was the development of design concepts for special-purpose remote maintenance systems. Included were systems for shield sector replacement, vacuum vessel sector and toroidal field coil replacement, limiter blade replacement, protective tile replacement, and general-purpose maintenance. This paper addresses these systems as they apply to the copper toroidal field (TF) coil version of the TFCX

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  4. Tokamak Simulation Code modeling of NSTX

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.

    2000-01-01

    The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption

  5. Nitrogen Gas Heating and Supply System for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Khan, Ziauddin; Pathan, Firozkhan; Paravastu, Yuvakiran; George, Siju; Ramesh, Gattu; Bindu, Hima; Raval, Dilip C.; Thankey, Prashant; Dhanani, Kalpesh; Pradhan, Subrata

    2013-01-01

    Steady State Tokamak (SST-1) vacuum vessel baking as well as baking of the first wall components of SST-1 are essential to plasma physics experiments. Under a refurbishment spectrum of SST-1, the nitrogen gas heating and supply system has been fully refurbished. The SST-1 vacuum vessel consists of ultra-high vacuum (UHV) compatible eight modules and eight sectors. Rectangular baking channels are embedded on each of them. Similarly, the SST-1 plasma facing components (PFC) are comprised of modular graphite diverters and movable graphite based limiters. The nitrogen gas heating and supply system would bake the plasma facing components at 350°C and the SST-1 vacuum vessel at 150°C over an extended duration so as to remove water vapour and other absorbed gases. An efficient PLC based baking facility has been developed and implemented for monitoring and control purposes. This paper presents functional and operational aspects of a SST-1 nitrogen gas heating and supply system. Some of the experimental results obtained during the baking of SST-1 vacuum modules and sectors are also presented here. (fusion engineering)

  6. 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

  7. 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

  8. Study of runaway electrons using dosimetry of hard x-ray radiations in Damavand tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Rasouli, C.; Pourshahab, B.; Rasouli, H. [Plasma Physics and Nuclear Fusion Research School, Nuclear Science and Technology Research Institute, AEOI, PO Box 14155-1339, Tehran (Iran, Islamic Republic of); Hosseini Pooya, S. M.; Orouji, T. [Radiation Application Research School, Nuclear Science and Technology Research Institute, AEOI, PO Box 14155-1339, Tehran (Iran, Islamic Republic of)

    2014-05-15

    In this work several studies have been conducted on hard x-ray emissions of Damavand tokamak based on radiation dosimetry using the Thermoluminescence method. The goal was to understand interactions of runaway electrons with plasma particles, vessel wall, and plasma facing components. Total of 354 GR-200 (LiF:Mg,Cu,P) thermoluminescence dosimeter (TLD) crystals have been placed on 118 points – three TLDs per point – to map hard x-ray radiation doses on the exterior of the vacuum vessel. Results show two distinctive levels of x-ray radiations doses on the exterior of the vessel. The low-dose area on which measured dose is about 0.5 mSv/shot. In the low-dose area there is no particular component inside the vessel. On the contrary, on high-dose area of the vessel, x-ray radiations dose exceeds 30 mSv/shot. The high-dose area coincides with the position of limiters, magnetic probe ducts, and vacuum vessel intersections. Among the high-dose areas, the highest level of dose is measured in the position of the limiter, which could be due to its direct contact with the plasma column and with runaway electrons. Direct collisions of runaway electrons with the vessel wall and plasma facing components make a major contribution for production of hard x-ray photons in Damavand tokamak.

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

    Science.gov (United States)

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

    2015-01-01

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

  10. Design of the Cryostat for HT-7U Superconducting Tokamak

    Science.gov (United States)

    Yu, Jie; Wu, Song-tao; Song, Yun-tao; Weng, Pei-de

    2002-06-01

    The cryostat of HT-7U tokamak is a large vacuum vessel surrounding the entire basic machine with a cylindrical shell, a dished top and a flat bottom. The main function of HT-7U cryostat is to provide a thermal barrier between an ambient temperature test hall and a liquid helium-cooled superconducting magnet. The loads applied to the cryostat are from sources of vacuum pressure, dead weight, seismic events and electromagnetic forces originated by eddy currents. It also provides feed-through penetrations for all the connecting elements inside and outside the cryostat. The main material selected for the cryostat is stainless steel 304L. The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out by using a finite element code. Stress analysis results show that the maximum stress intensity was below the allowable value. In this paper, the structural analyses and design of HT-7U cryostat are emphasized.

  11. Study on assembly techniques and procedures for ITER tokamak device

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi; Ue, Koichi; Shimizu, Katsusuke; Onozuka, Masanori

    2006-06-01

    The International Thermonuclear Experimental Reactor (ITER) tokamak is mainly composed of a doughnut-shaped vacuum vessel (VV), four types of superconducting coils such as toroidal field coils (TF coils) arranged around the VV, and in-vessel components, such as blanket and divertor. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of the VV and the TF coil are required to be a high accuracy of ±3 mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements as well as the configuration of the tokamak with large size and heavy weight. Based on the above backgrounds, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The tokamak assembly operations are categorized into six work break down structures (WBS), i.e., (1) preparation for assembly operations, (2) sub-assembly of the 40deg sector composed of 40deg VV sector, two TF coils and thermal shield between VV and TF coil at the assembly hall, (3) completion of the doughnut-shaped tokamak assembly composed of nine 40deg sectors in the cryostat at the tokamak pit, (4) measurement of positioning and accuracy after the completion of the tokamak assembly, (5) installation of the ex-vessel components, and (6) installation of in-vessel components. In the present report, two assembly operations of (2) and (3) in the above six WBS, which are the most critical in the tokamak assembly, are mainly described. The report describes the following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology

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

    Energy Technology Data Exchange (ETDEWEB)

    Martín-Menéndez, Cristina, E-mail: cristina@natec-ingenieros.com [Numerical Analysis Technologies, S.L. Marqués de San Esteban No. 52, 33206 Gijón (Spain); Rodríguez, Eduardo [Department of Mechanical Engineering, University of Oviedo, Campus de Gijón, 33203 Gijón (Spain); Ottolini, Marco [Ansaldo Nucleare S.p.A., Corso Perrone 25, 16152 Genova (Italy); Caixas, Joan [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Guirao, Julio [Numerical Analysis Technologies, S.L. Marqués de San Esteban No. 52, 33206 Gijón (Spain)

    2016-03-15

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

  13. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

  14. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs

  15. Initial plasma production by induction electric field on QUEST tokamak

    International Nuclear Information System (INIS)

    Hasegawa, Makoto; Nakamura, Kazuo; Sato, Kohnosuke

    2007-01-01

    Induction electric field by center solenoid coil plays a roll to produce initial plasma. According to Townsend avalanche theory, minimum electric field for plasma breakdown depends on neutral gas pressure and connection length. On QUEST spherical tokamak, a connection length is evaluated as 966m on null point neighborhood with coil current ratio I PF26 /I CS =0.1, and induction electric field considering eddy current of vacuum vessel is evaluated as about 0.1 V/m on null point neighborhood. With Townsend avalanche theory, these values manage to produce initial plasma on QUEST. (author)

  16. Eddy current calculations for the tore supra tokamak

    International Nuclear Information System (INIS)

    Blum, J.; Dupas, L.; Leloup, C.; Thooris, B.

    1983-01-01

    This paper deals with the calculation of the eddy currents in the structures of a Tokamak, which can be assimilated to thin conductors, so that the three-dimensional problem can be reduced mathematically to a two-dimensional one, the variables being two orthogonal coordinates of the considered surface. A variational formulation of the problem in terms of the electric vector potential is then given and a finite element method has been used, which enables to treat the complicated geometry of the toroidal field magnet, the mechanical structures and the vacuum vessels of Tore Supra

  17. A DESIGN RETROSPECTIVE OF THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    LUXON, J.L

    2001-06-01

    OAK-B135 The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and rf heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research program. This paper gives an integrated picture of the facility and its capabilities

  18. Optimized Baking of the DIII-D Vessel

    International Nuclear Information System (INIS)

    Anderson, P.M.; Kellman, A.G.

    1999-01-01

    The DIII-D tokamak vacuum vessel baking system is used to heat the vessel walls and internal hardware to an average temperature of 350 C to allow rapid conditioning of the vacuum surfaces. The system combines inductive heating and a circulating hot air system to provide rapid heating with temperature uniformity required by stress considerations. In recent years, the time to reach 350 C had increased from 9 hrs to 14 hrs. To understand and remedy this sluggish heating rate, an evaluation of the baking system was recently performed. The evaluation indicated that the mass of additional in-vessel hardware (50% increase in mass) was primarily responsible. This paper reports on this analysis and the results of the addition of an electric air heater and procedural changes that have been implemented. Preliminary results indicate that the time to 350 C has been decreased to 4.5 hours and the temperature uniformity has improved

  19. 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

  20. Maintenance concept development for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Macdonald, D.

    1988-01-01

    The Compact Ignition Tokamak (CIT), located at the Princeton Plasma Physics Laboratory, will be the next major experimental machine in the US Fusion Program. Its use of deuterium-tritium (D-T) fuel requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist of removing and repairing such components as diagnostic equipment modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the vacuum vessel includes both bridge-mounted and floor-mounted manipulator systems. Additionally, decontamination (decon) equipment, hot cell repair facilities, and equipment for handling and packaging solid radioactive waste (rad-waste) are being developed. Recent design activities have focused on establishing maintenance system interfaces with the facility design, developing manipulator system requirements, and using mock-ups to support the tokamak configuration design. 3 refs., 8 figs

  1. Statistical analysis of first period of operation of FTU Tokamak

    International Nuclear Information System (INIS)

    Crisanti, F.; Apruzzese, G.; Frigione, D.; Kroegler, H.; Lovisetto, L.; Mazzitelli, G.; Podda, S.

    1996-09-01

    On the FTU Tokamak the plasma physics operations started on the 20/4/90. The first plasma had a plasma current Ip=0.75 MA for about a second. The experimental phase lasted until 7/7/94, when a long shut-down begun for installing the toroidal limiter in the inner side of the vacuum vessel. In these four years of operations plasma experiments have been successfully exploited, e.g. experiments of single and multiple pellet injections; full current drive up to Ip=300 KA was obtained by using waves at the frequency of the Lower Hybrid; analysis of ohmic plasma parameters with different materials (from the low Z silicon to high Z tungsten) as plasma facing element was performed. In this work a statistical analysis of the full period of operation is presented. Moreover, a comparison with the statistical data from other Tokamaks is attempted

  2. New results from Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.

    2002-01-01

    New results from Globus-M spherical tokamak (ST) are presented. Reported are the achievements of high plasma current of 0.36 MA and high toroidal magnetic field of 0.55 T. Plasma column stability in Globus-M is conserved at low edge safety factors and high plasma densities. Achieved lowest safety factor was q(cyl) 19 m -3 . New methods of density increase are discussed. Low-density boarder of operational space is investigated. Runaway electrons properties and conditions of their generation are investigated. Results look promising for STs. Plasma-wall interaction study was performed. Silicon probes were installed into vacuum vessel. They were exposed to boronization, first, and then deposited film interacted with plasma. Discussed are film properties. Briefly described are new diagnostic tools installed on tokamak. Status and preliminary results obtained with auxiliary heating systems are shown. (author)

  3. Migration of a radioactive beryllium in the ISX-B tokamak

    International Nuclear Information System (INIS)

    England, A.C.; Hillis, D.L.; Edmonds, P.H.

    1985-10-01

    One of 12 beryllium tiles on a top rail limiter in the Impurity Study Experiment (ISX-B) tokamak was intentionally made radioactive. The migration of the radioactivity due to melting, ablation, and general excoriation of the radioactive tile was studied. Several hundred milligrams of material from the radioactive tile were spread to the other tiles over the course of the experiment, which consisted of more than 3000 tokamak shots. A small amount of activity was found on the bottom of the tokamak vacuum vessel. The distribution of activity on the other limiter tiles showed a marked radially inward directivity, although all of the tiles received some activity. The possibility that some of the activity was the result of photo- and electro-excitation by runaway electrons cannot be ruled out but probably cannot account for the bulk of the effect

  4. JT-60 power tests from mechanical and thermal viewpoints of tokamak machine

    International Nuclear Information System (INIS)

    Takatsu, H.; Yamamoto, M.; Ohkubo, M.

    1986-01-01

    JT-60 power tests were carried out, to demonstrate, in advance of actual plasma operation, satisfactory performance of the tokamak machine, power suppliers and control system in combination. The tests began with low power ones of individual coil systems, progressed to full power ones and concluded successfully. The present paper describes the principal results of JT-60 power tests from mechanical and thermal viewpoints of tokamak machine. All of the coil systems were raised up to full power operation in combination and system performance was verified including thermal and mechanical integrity of tokamak machine. Measured strain and displacement showed good agreements with those predicted in the design, which was an evidence that electromagnetic loads were supported adequately as expected in the design. Vibration of the vacuum vessel was found to be large up to 48 m/s/sup 2/ and caused excessive vibration of the lateral port gate-valves. A few limitations to machine operation were also made clear quantatively

  5. Optimization of magnetic field system for glass spherical tokamak GLAST-III

    International Nuclear Information System (INIS)

    Ahmad, Zahoor; Ahmad, S; Naveed, M A; Deeba, F; Javeed, M Aqib; Batool, S; Hussain, S; Vorobyov, G M

    2017-01-01

    GLAST-III (Glass Spherical Tokamak) is a spherical tokamak with aspect ratio A = 2. The mapping of its magnetic system is performed to optimize the GLAST-III tokamak for plasma initiation using a Hall probe. Magnetic field from toroidal coils shows 1/ R dependence which is typical with spherical tokamaks. Toroidal field (TF) coils can produce 875 Gauss field, an essential requirement for electron cyclotron resonance assisted discharge. The central solenoid (CS) of GLAST-III is an air core solenoid and requires compensation coils to reduce unnecessary magnetic flux inside the vessel region. The vertical component of magnetic field from the CS in the vacuum vessel region is reduced to 1.15 Gauss kA −1 with the help of a differential loop. The CS of GLAST can produce flux change up to 68 mVs. Theoretical and experimental results are compared for the current waveform of TF coils using a combination of fast and slow capacitor banks. Also the magnetic field produced by poloidal field (PF) coils is compared with theoretically predicted values. It is found that calculated results are in good agreement with experimental measurement. Consequently magnetic field measurements are validated. A tokamak discharge with 2 kA plasma current and pulse length 1 ms is successfully produced using different sets of coils. (paper)

  6. Statistical analysis of first period of operation of FTU Tokamak; Analisi statistica del primo periodo di operazioni del Tokamak FTU

    Energy Technology Data Exchange (ETDEWEB)

    Crisanti, F; Apruzzese, G; Frigione, D; Kroegler, H; Lovisetto, L; Mazzitelli, G; Podda, S [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Energia

    1996-09-01

    On the FTU Tokamak the plasma physics operations started on the 20/4/90. The first plasma had a plasma current Ip=0.75 MA for about a second. The experimental phase lasted until 7/7/94, when a long shut-down begun for installing the toroidal limiter in the inner side of the vacuum vessel. In these four years of operations plasma experiments have been successfully exploited, e.g. experiments of single and multiple pellet injections; full current drive up to Ip=300 KA was obtained by using waves at the frequency of the Lower Hybrid; analysis of ohmic plasma parameters with different materials (from the low Z silicon to high Z tungsten) as plasma facing element was performed. In this work a statistical analysis of the full period of operation is presented. Moreover, a comparison with the statistical data from other Tokamaks is attempted.

  7. Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

    Energy Technology Data Exchange (ETDEWEB)

    Matsukawa, M. E-mail: matsukaw@naka.jaeri.go.jp; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T

    2003-09-01

    The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control.

  8. Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

    International Nuclear Information System (INIS)

    Matsukawa, M.; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T.

    2003-01-01

    The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  10. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

    The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

  11. A simple in-vessel/FW component viewing system for SST-1

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Biswas, Prabal; Vasava, Kirit R.; Jaiswal, Snehal; Parekh, Tejas; Chauhan, Pradeep; Patel, Hiteshkumar; Pradhan, Subrata

    2015-01-01

    A simple compact system is being proposed for in-situ visual inspection of around 3800 First Wall (FW) graphite (armour) tiles in the vacuum vessel of SST-1 tokamak. The 2 DOF, manual driven system (permanently stationed inside vacuum vessel behind outer passive stabilizer) at top and bottom mid-plane locations consist of a rack and pinion mechanism operating a arm with a CCD camera/LED mounted on it, moving over a cam profile to cover approximately 1/8 th of the toroidal span of the vacuum vessel both at interior top/bottom locations with in the FW modules. The camera and LED light should withstand the ultrahigh vacuum conditions, prolonged baking temperatures of around 200°C along with high electromagnetic forces inside the vessel. This system can be operated remotely in-between shots from outside the VV through a linear motion feed through providing linear moment to a rack and pinion mechanism connected to the arm. This mechanism provides a better viewing of the inside FW components and vessel wall surface of tokamak with simple engineering and operational effort. Any information can be acquired from system regarding damages to FWC due to interaction with plasma as well as damage of other support structures inside VV. In comparison to more complicated and complex inspection system used in other tokamaks, this mechanism can be used for frequent in vessel visual inspection, which limits the system to be small, simple, occupying less space and custom made. This system is cheap with a minimum time for realization of the concept. The paper will present the conceptual and engineering design aspect of the in-viewing system, CAD images, its advantages and limitations, camera and LED details, data acquisition and the present status of realization of the project. (author)

  12. Study of impurities in Aditya Tokamak during different conditions using quadrupole mass analyzer

    International Nuclear Information System (INIS)

    Bhatt, S.B.; Jadeja, K.A.; Patel, K.M.; Patel, N.D.; Raval, M.K.; Ghosh, J.

    2015-01-01

    In fusion devices, e.g., Tokamak, the presence of the impurities, i.e. gas species other than the fuel gas, deteriorates plasma and makes confinement difficult. The gas molecules tend to get adsorbed on the surfaces of the solid state materials of the vessel wall during discharges. A Residual Gas Analyzer (RGA) is the most commonly useful instrument to measure the presence and quantity of the various gases in a vacuum system. Quadrupole Mass Analyzer (QMA) is installed on Aditya Tokamak to measure the concentrations of various gas species present in Aditya vacuum system. It is also used to monitor impurities generated during various phases of discharges in Aditya Tokamak. The impurities are reduced by various types of discharge cleaning and in-situ coatings. Presence of residual gas concentration in vacuum system creates limitation for achievement of ultrahigh vacuum and also affects plasma performance. The presence of residual gases is due to different reasons like atmospheric concentration, contamination of the wall materials, outgassing from the exposed materials, permeation, real and virtual leaks

  13. Conceptual Design of Alborz Tokamak Poloidal Coils System

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.

    2013-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.

  14. Midplane Faraday rotation: A densitometer for large tokamaks

    International Nuclear Information System (INIS)

    Jobes, F.C.; Mansfield, D.K.

    1992-01-01

    The density in a large tokamak such as International Thermonuclear Experimental Reactor (ITER), or any of the proposed future US machines, can be determined by measuring the Faraday rotation of a 10.6 μm laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then n e (R) can be readily obtained with a simple Abel inversion about the center line of the tokamak. For a large machine, operated at a full field of 30 T m and a density of 2x10 20 /m 3 , the rotation angle would be quite large-about 60 degree for two passes. A layout in which a single laser beam is fanned out in the horizontal midplane of the tokamak, with a set of retroreflectors on the far side of the vacuum vessel, would provide good spatial resolution, depending only upon the number of reflectors. With this proposed layout, only one window would be needed. Because the rotation angle is never more than 1 ''fringe,'' the data is always good, and it is also a continuous measurement in time. Faraday rotation is dependent only upon the plasma itself, and thus is not sensitive to vibration of the optical components. Simulations of the expected results show that ITER, or any large tokamak, existing or proposed, would be well served even at low densities by a midplane Faraday rotation densitometer of ∼64 channels

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  16. A preliminary design and structural analysis on the central column for supporting the full 40 .deg. Sectors at tokamak in pit

    International Nuclear Information System (INIS)

    Nam, Kyoungo; Park, Hyun Ki; Kim, Dong Jin; Moon, Jaeh Wan; Kim, Byung Seok; Watson, Emma; Shaw, Robert

    2012-01-01

    The ITER Tokamak machine is composed of nine 40 .deg. sectors shaped of torus. Each 40.deg. sector is made up of one 40 .deg. vacuum vessel (VV), two 20 .deg. toroidal filed coils and associated vacuum vessel thermal shield (VVTS) segments which consist of one inboard and two outboard VVTS. The VV/TFC/VVTS 40 .deg. sectors are sub assembled at assembly building respectively at sector sub assembly tool and then nine sub assembled 40 .deg. sectors are finally assembled at in-pit of Tokamak building. ITER sector assembly tools are the purpose built assembly tools to assemble nine 40 .deg. sectors into the complete ITER Tokamak machine at Tokamak in pit. Based on the design description document, final report prepared by the ITER organization (IO) and tooling requirements, Korea has carried out the conceptual and preliminary design of these assembly tools. Especially, the central column is the main tool, which is composed of some hollow cylinders, to support full nine 40 .deg. sectors at Tokamak in pit. Configuration and structural analysis of the central column are presented

  17. Compact Ignition Tokamak Program: R and D needs

    International Nuclear Information System (INIS)

    Flanagan, C.A.

    1985-01-01

    This report on the Compact Ignition Tokamak Program supplies information concerning: segmented vacuum vessel joint development; first wall tile attachments; first wall/tile development - composite materials; vacuum leak detection; high frequency rf sources; Faraday shield development; design and testing of rf launchers for high power, ling pulse operation; radiation hardened, low loss, dielectric windows for rf, IR, visible, UV and X-rays, mirrors for changing direction and focusing IR, visible and UV radiation; radiation resistant optical dielectric wave guides; radiation resistant HV insulation for diagnostic magnetic pickup coils; compact radiation and/or magnetic shielding for in-vault diagnostics that need some attenuation to reduce S/N ratio; radiation hardened line-of-sight sensors such as bolometers, UV and soft X-ray detectors, neutral particle analyzers, torus pressure gauges; special maintenance fixtures and tools; material properties - design data base - all materials; and insulation - electrical/thermal and mechanical properties

  18. Tokamak building-design considerations for a large tokamak device

    International Nuclear Information System (INIS)

    Barrett, R.J.; Thomson, S.L.

    1981-01-01

    Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

  19. 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

  20. Structural analysis of the JT-60SA cryostat vessel body

    Energy Technology Data Exchange (ETDEWEB)

    Botija, José, E-mail: jose.botija@ciemat.es [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Alonso, Javier; Fernández, Pilar; Medrano, Mercedes; Ramos, Francisco; Rincon, Esther; Soleto, Alfonso [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Davis, Sam; Di Pietro, Enrico; Tomarchio, Valerio [Fusion for Energy, JT-60SA European Home Team, 85748 Garching bei Munchen (Germany); Masaki, Kei; Sakasai, Akira; Shibama, Yusuke [JAEA – Japan Atomic Energy Agency, Naka Fusion Institute, Ibaraki 311-0193 (Japan)

    2013-10-15

    Highlights: ► Structural analysis to validate the JT-60SA cryostat vessel body design. ► Design code ASME 2007 “Boiler and Pressure Vessel Code. Section VIII”. ► First buckling mode: load multiplier of 10.644, higher than the minimum factor 4.7. ► Elastic and elastic–plastic stress analysis meets ASME against plastic collapse. ► Bolted fasteners have been analyzed showing small gaps closed by strong welding. -- Abstract: The JT-60SA cryostat is a stainless steel vacuum vessel (14 m diameter, 16 m height) which encloses the Tokamak providing the vacuum environment (10{sup −3} Pa) necessary to limit the transmission of thermal loads to the components at cryogenic temperature. It must withstand both external atmospheric pressure during normal operation and internal overpressure in case of an accident. The paper summarizes the structural analyses performed in order to validate the JT-60SA cryostat vessel body design. It comprises several analyses: a buckling analysis to demonstrate stability under the external pressure; an elastic and an elastic–plastic stress analysis according to ASME VIII rules, to evaluate resistance to plastic collapse including localized stress concentrations; and, finally, a detailed analysis with bolted fasteners in order to evaluate the behavior of the flanges, assuring the integrity of the vacuum sealing welds of the cryostat vessel body.

  1. Remote replacement of TF [toroidal field] and PF [poloidal field] coils for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Macdonald, D.; Watkin, D.C.; Hollis, M.J.; DePew, R.E.; Kuban, D.P.

    1990-01-01

    The use of deuterium-tritium fuel in the Compact Ignition Tokamak will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion devices auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. In-vessel remote maintenance included replacement of divertor and first wall hardware, faraday shields, and for an in-vessel inspection system. Provision for remote replacement of a vacuum vessel sector, toroidal field coil or poloidal field ring coil was not included in the project baseline. As a result of recent coil failures experienced at a number of facilities, the CIT project decided to reconsider the question of remote recovery from a coil failure and, in January of 1990, initiated a coil replacement study. This study focused on the technical requirements and impact on fusion machine design associated with remote recovery from any coil failure

  2. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

  3. Conversion of magnetic energy to runaway kinetic energy during the termination of runaway current on the J-TEXT tokamak

    Science.gov (United States)

    Dai, A. J.; Chen, Z. Y.; Huang, D. W.; Tong, R. H.; Zhang, J.; Wei, Y. N.; Ma, T. K.; Wang, X. L.; Yang, H. Y.; Gao, H. L.; Pan, Y.; the J-TEXT Team

    2018-05-01

    A large number of runaway electrons (REs) with energies as high as several tens of mega-electron volt (MeV) may be generated during disruptions on a large-scale tokamak. The kinetic energy carried by REs is eventually deposited on the plasma-facing components, causing damage and posing a threat on the operation of the tokamak. The remaining magnetic energy following a thermal quench is significant on a large-scale tokamak. The conversion of magnetic energy to runaway kinetic energy will increase the threat of runaway electrons on the first wall. The magnetic energy dissipated inside the vacuum vessel (VV) equals the decrease of initial magnetic energy inside the VV plus the magnetic energy flowing into the VV during a disruption. Based on the estimated magnetic energy, the evolution of magnetic-kinetic energy conversion are analyzed through three periods in disruptions with a runaway current plateau.

  4. Quality control of FWC during assembly and commissioning in SST-1 Tokamak

    Science.gov (United States)

    Patel, Hitesh; Santra, Prosenjit; Parekh, Tejas; Biswas, Prabal; Jayswal, Snehal; Chauhan, Pradeep; Paravastu, Yuvakiran; George, Siju; Semwal, Pratibha; Thankey, Prashant; Ramesh, Gattu; Prakash, Arun; Dhanani, Kalpesh; Raval, D. C.; Khan, Ziauddin; Pradhan, Subrata

    2017-04-01

    First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma, comprises of limiters, divertors, baffles, passive stabilizers designed to operate long duration (∼1000 s) discharges of elongated plasma. All FWC consist of copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at inter-connected ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a rigorous quality control and checks at every stage of the assembly process. This paper will present the quality control aspects and checks of FWC from commencement of assembly procedure, namely material test reports, leak testing of high temperature baked components, assembled dimensional tolerances, leak testing of all welded joints, graphite tile tightening torques, electrical continuity and electrical isolation of passive stabilizers from vacuum vessel, baking and cooling hydraulic connections inside vacuum vessel.

  5. Automation of Aditya vacuum control system based on CODAC Core System

    Energy Technology Data Exchange (ETDEWEB)

    Raulji, Vismaysinh D., E-mail: vismay@ipr.res.in; Pujara, Harshad; Arambhadiya, Bharat; Jadeja, Kumarpalsinh; Bhatt, Shailesh; Rajpal, Rachana

    2016-11-15

    Highlights: • Monitor and control of vacuum control system based on CODAC Core System. • Communication between SIEMENS PLC and open source software EPICS. • With CODAC Core easy to configure and programming of slow controller. - Abstract: The main objective of vacuum control system is to provide ultrahigh vacuum for Aditya Tokamak operations. Aditya Vacuum vessel is having four vacuum pumping lines. To demonstrate implementation of automation; a study case is under taken by automating single Pumping Line of the Aditya vacuum system using CODAC Core System (CCS). Currently, vacuum system is operated manually. The CCS based control system allows remote control, monitoring, alarm handling of vacuum parameters. The CODAC Core System is the Linux based software package that is distributed by ITER Organization for the development of Plant System I&C software. CODAC Core System includes EPICS, CSS (Control System Studio) etc. CSS is used for HMI (Human Machine Interface), alarms and archives. SDD (Self Description Data) tool is used to configure plant system I&C. SDD Editor is an Eclipse based application to define the plant system, interface, I&C component, interfaced signals, configure variable. SCADA (Supervisory Control and Data Acquisition) system is developed in CSS. Data is transferred between PLC and CSS through EPICS. The complete system is tested with Aditya Vacuum Control System with process interlocks. Operator interface is also developed using Lab VIEW as a choice of the user. This paper will describe the salient features of the developed control system in detail.

  6. Analysis of images from videocameras in the Frascati Tokamak Upgrade tokamak

    International Nuclear Information System (INIS)

    De Angelis, R.; Migliori, S.; Borioni, S.; Bracco, G.; Pierattini, S.; Perozziello, A.

    2004-01-01

    The plasma edge interaction in FTU tokamak is monitored by wide angle videocameras. Data are acquired as movies or single frames at a rate of 50 frames/s. The images show interesting features of the plasma such as the presence of Marfes or runaways and give useful information on the status of large parts of the vacuum vessel and toroidal limiter. Due to the large number of data available visual inspection of the movies is often insufficient to correlate the images to the experimental findings. This article illustrates a number of applications developed in order to correlate the images with plasma signals and to search the image database for specific features relevant to the discharge

  7. Safety implications of a graphite oxidation accident in the compact ignition tokamak device

    International Nuclear Information System (INIS)

    Merrill, B.J.; O'Brien, M.H.

    1989-01-01

    This paper addresses the possible safety consequences of an air ingress accident for the Compact Ignition Tokamak (CIT) device. An experimental program was undertaken to determine oxidation rates of four nuclear grade graphites in air at temperatures ranging from 800 to 1800 C and flow velocities from 3 to 7 m/s. On the basis of these test results, an analytic model was developed to assess the extent of first wall/divertor protective tile oxidation and the amount of energy released from this oxidation. For CIT, a significant restriction to vacuum vessel air inflow will be provided by the air seals and walls of the surrounding test cells. Under these conditions, the graphite oxidation reaction inside the vacuum vessel will become oxygen starved within minutes of the onset of this event. Since significant oxidation rates were not achieved, the heat release did not elevate structural temperatures to levels of concern with regard to activated material release. 7 refs., 9 figs

  8. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

    For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

  9. Conceptual design of EAST flexible in-vessel inspection system

    International Nuclear Information System (INIS)

    Peng, X.B.; Song, Y.T.; Li, C.C.; Lei, M.Z.; Li, G.

    2010-01-01

    Remote handling technology, especially the flexible in-vessel inspection system (FIVIS) without breaking the working condition of the vacuum vessel, has been identified as one major challenge on the maintenance for the future tokamak fusion reactor. The FIVIS introduced here is specially developed for EAST superconducting tokamak that has actively cooled plasma facing components (PFCs). It aims flexible close-up inspection of EAST PFCs to help the understanding of operation issues that could occur in the vacuum vessel. This paper resumes the preliminary work of the FIVIS project, including the requirement analysis and the development of the conceptual design. The FIVIS consists out of a long reach multi-articulated manipulator and a process tool. The manipulator has a modular design for its subsystems and can reach all areas of the first wall in the distance of 15 mm and in the range of ±90 o along toroidal direction. It will be folded and hidden in the designated horizontal port during plasma discharge period.

  10. Overall behaviour of PFC integrated SST-1 vacuum system

    Science.gov (United States)

    Khan, Ziauddin; Raval, Dilip C.; Paravasu, Yuvakiran; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; George, Siju; Shoaib, Mohammad; Prakash, Arun; Babu, Gattu R.; Thankey, Prashant; Pathan, Firozkhan S.; Pradhan, Subrata

    2017-04-01

    As a part of phase-I up-gradation of Steady-state Superconducting Tokamak (SST-1), Graphite Plasma Facing Components (PFCs) have been integrated inside SST-1 vacuum vessel as a first wall (FW) during Nov 14 and May 2015. The SST-1 FW has a total surface area of the installed PFCs exposed to plasma is ∼ 40 m2 which is nearly 50% of the total surface area of stainless steel vacuum chamber (∼75 m2). The volume of the vessel within the PFCs is ∼ 16 m3. After the integration of PFCs, the entire vessel as well as the PFC cooling/baking circuits has been qualified with an integrated helium leak tightness of baked at 250 °C for nearly 20 hours employing hot nitrogen gas to remove the absorbed water vapours. Thereafter, Helium glow discharges cleaning were carried out towards the removal of surface impurities. The pump down characteristics of SST-1 vacuum chamber and the changes in the residual gaseous impurities after the installation of the PFCs will be discussed in this paper.

  11. FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak

    International Nuclear Information System (INIS)

    Suratia, Pooja; Patel, Jigneshkumar; Rajpal, Rachana; Kotia, Sorum; Govindarajan, J.

    2012-01-01

    Highlights: ► Evaluation and comparison of the working performance of FLC is done with that of PID Controller. ► FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. ► FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. ► Developed FLC controller is able to maintain the plasma column within required range of ±0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional–Integral–Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).

  12. Design features of the KSTAR in-vessel control coils

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H.K. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)], E-mail: hkkim@nfri.re.kr; Yang, H.L.; Kim, G.H.; Kim, Jin-Yong; Jhang, Hogun; Bak, J.S.; Lee, G.S. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)

    2009-06-15

    In-vessel control coils (IVCCs) are to be used for the fast plasma position control, field error correction (FEC), and resistive wall mode (RWM) stabilization for the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The IVCC system comprises 16 segments to be unified into a single set to achieve following remarkable engineering advantages; (1) enhancement of the coil system reliability with no welding or brazing works inside the vacuum vessel, (2) simplification in fabrication and installation owing to coils being fabricated outside the vacuum vessel and installed after device assembly, and (3) easy repair and maintenance of the coil system. Each segment is designed in 8 turns coil of 32 mm x 15 mm rectangular oxygen free high conductive copper with a 7 mm diameter internal coolant hole. The conductors are enclosed in 2 mm thick Inconel 625 rectangular welded vacuum jacket with epoxy/glass insulation. Structural analyses were implemented to evaluate structural safety against electromagnetic loads acting on the IVCC for the various operation scenarios using finite element analysis. This paper describes the design features and structural analysis results of the KSTAR in-vessel control coils.

  13. 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

  14. Outgassing rates before, during and after bake-out for various vacuum and first wall candidate materials of a large tokamak device

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Gomay, J.; Sugiyama, Y.; Mizuno, M.; Komiya, S.; Tazima, T.

    1977-01-01

    Outgassing rates of vacuum wall candidate materials; stainless steel SS-304L and YUS-170, Inconel-625 and Hastelloy-X, and first wall materials; molybdenum, pyrolytic graphite and silicon carbide are measured before, during and after a bake-out at 500 0 C. The outgassing rate from the inside wall of the cylinder made of each material is estimated from the pressure difference between before and after a calibrated orifice. The ultimate outgassing rates of SS-304L and pyrolytic graphite, and YUS-170 Inconel-625, Hastelloy-X and molybdenum are the orders of 10 -10 and 10 -11 Pa.l.s -1 cm -2 , respectively

  15. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  16. Disruption-induced poloidal currents in the tokamak wall

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2017-01-01

    Highlights: • Induction effects during disruptions and rapid transient events in tokamaks. • Plasma-wall electromagnetic interaction. • Flux-conserving evolution of plasma equilibrium. • Poloidal current induced in the vacuum vessel wall in a tokamak. • Complete analytical derivations and estimates. - Abstract: The poloidal current induced in the tokamak wall during fast transient events is analytically evaluated. The analysis is based on the electromagnetic relations coupled with plasma equilibrium equations. The derived formulas describe the consequences of both thermal and current quenches. In the final form, they give explicit dependence of the wall current on the plasma pressure and current. A comparison with numerical results of Villone et al. [F. Villone, G. Ramogida, G. Rubinacci, Fusion Eng. Des. 93, 57 (2015)] for IGNITOR is performed. Our analysis confirms the importance of the effects described there. The estimates show that the disruption-induced poloidal currents in the wall should be necessarily taken into account in the studies of disruptions and disruption mitigation in ITER.

  17. Disruption-induced poloidal currents in the tokamak wall

    Energy Technology Data Exchange (ETDEWEB)

    Pustovitov, V.D., E-mail: Pustovitov_VD@nrcki.ru [National Research Centre ‘Kurchatov Institute’, Pl. Kurchatova 1, Moscow 123182 (Russian Federation); National Research Nuclear University MEPhI, Kashirskoe sh. 31, Moscow 115409, Russia (Russian Federation)

    2017-04-15

    Highlights: • Induction effects during disruptions and rapid transient events in tokamaks. • Plasma-wall electromagnetic interaction. • Flux-conserving evolution of plasma equilibrium. • Poloidal current induced in the vacuum vessel wall in a tokamak. • Complete analytical derivations and estimates. - Abstract: The poloidal current induced in the tokamak wall during fast transient events is analytically evaluated. The analysis is based on the electromagnetic relations coupled with plasma equilibrium equations. The derived formulas describe the consequences of both thermal and current quenches. In the final form, they give explicit dependence of the wall current on the plasma pressure and current. A comparison with numerical results of Villone et al. [F. Villone, G. Ramogida, G. Rubinacci, Fusion Eng. Des. 93, 57 (2015)] for IGNITOR is performed. Our analysis confirms the importance of the effects described there. The estimates show that the disruption-induced poloidal currents in the wall should be necessarily taken into account in the studies of disruptions and disruption mitigation in ITER.

  18. Vertical poloidal asymmetries of low-Z element radiation in the PDX tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Brau, K.; Suckewer, S.; Wong, S.K.

    1983-06-01

    Vertical poloidal asymmetries of hydrogen isotopes and low-Z impurity radiation in the PDX tokamak may be caused by poloidally asymmetric sources of these elements at gas inlet valves, limiters or vacuum vessel walls, asymmetric magnetic field geometry in the region beyond the plasma boundary, or by ion curvature drifts. Low ionization states of carbon (C II- C IV) are more easily influenced by edge conditions than is CV. Vertical poloidal asymmetries of CV are correlated with the direction of the toroidal field. The magnitude of the asymmetry agrees with the predictions of a quasifluid neoclassical model. Experimental data and numerical simulations are presented to investigate different models of impurity poloidal asymmetries.

  19. Vertical poloidal asymmetries of low-Z element radiation in the PDX tokamak

    International Nuclear Information System (INIS)

    Brau, K.; Suckewer, S.; Wong, S.K.

    1983-06-01

    Vertical poloidal asymmetries of hydrogen isotopes and low-Z impurity radiation in the PDX tokamak may be caused by poloidally asymmetric sources of these elements at gas inlet valves, limiters or vacuum vessel walls, asymmetric magnetic field geometry in the region beyond the plasma boundary, or by ion curvature drifts. Low ionization states of carbon (C II- C IV) are more easily influenced by edge conditions than is CV. Vertical poloidal asymmetries of CV are correlated with the direction of the toroidal field. The magnitude of the asymmetry agrees with the predictions of a quasifluid neoclassical model. Experimental data and numerical simulations are presented to investigate different models of impurity poloidal asymmetries

  20. A conceptual design and structural stabilities of in-pit assembly tools for the completion of final sector assembly at tokamak hall

    International Nuclear Information System (INIS)

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

    2010-01-01

    The final assembly of main components of the International Thermonuclear Experimental Reactor (ITER) tokamak, Vacuum Vessel (VV) and Toroidal Field Coils (TFCs), is achieved by the sequential assembly of the nine sub-assembled 40 o sectors in tokamak pit. Each sub-assembled 40 o sector is composed of one VV 40 o sector, two TFCs, and in-between Vacuum Vessel Thermal Shield (VVTS) segments. Sub-assembly is carried out in the assembly building and then the sub-assembled sectors are transferred into tokamak pit, in sequence, to complete sector assembly. The role of in-pit assembly tool is to support and align the sub-assembled sectors in tokamak pit. It also plays the role of reference datum during assembly until the completion of main components assembly. Korea Domestic Agency (KO DA) has developed the conceptual design of most ITER purpose-built assembly tools under the collaboration with the ITER Organization. Among the conceptual designs carried out, this paper describes the function, the structure, the selected material and the design results of the in-pit assembly tools comprising central column, radial beams and their supports, TF inner supports and in-pit working floor. The results of structural analysis using ANSYS for the various loading cases are given as well. The resultant stresses and deflections turned out to fall within the allowable ranges.

  1. Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y. L.; Xu, G. S.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Tritz, K. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, Maryland 21218 (United States); Zhu, Y. B. [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States)

    2015-12-15

    A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.

  2. Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Li, Y. L.; Xu, G. S.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L.; Tritz, K.; Zhu, Y. B.

    2015-01-01

    A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks

  3. 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.

  4. Photo nuclear data needed to understand multi-MeV electrons behaviour on the ITER Tokamak

    International Nuclear Information System (INIS)

    Joyer, P.; Martin, G.

    1994-01-01

    During early operation of the future tokamak ITER, electrons will be accelerated up to hundreds of MeV during unavoidable disruptions of the plasma current. As they impinge on the vacuum vessel, they will create high intensity X-ray beams, source of high activation spots by photo nuclear spallation reactions. To estimate beforehand the induced dose rates, a reliable set of cross sections is needed: reactions of X-rays (from 10 to 500 MeV) on a few usual materials. In addition, to characterize these electron beams on present day tokamaks, as Tore-Supra, additional data for some more exotic elements in a lower energy range (< 100 MeV) could be useful. (authors). 6 refs., 2 figs

  5. Determination of plasma column transverse section in the TBR-1 tokamak

    International Nuclear Information System (INIS)

    Conde, M.E.

    1986-01-01

    The temporal evolution of plasma column transverse section in the TBR-1 tokamak is determined. The experimental melhod is based on the simulation of toroidal current distribution in plasma by a set of toroidal filaments. The currents in these filaments are determined by minimization of square error between the magnetic field produced by filaments and the field measured into the tokamak vacuum vessel. For the measurement of magnetic field, twenty small magnetic coils were constructed and installated in the region protected by current limiters. The plasma column transverse cross section is determined by poloidal field produced by the currents in filaments. The multipole moments of plasma current distribution and the Λ Shafranov parameter were obtained. (M.C.K.) [pt

  6. 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

  7. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  8. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

    With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

  9. A carbon-carbon panel design concept for the inboard limiter of the Compact Ignition Tokamak (CIT)

    International Nuclear Information System (INIS)

    Mantz, H.C.; Bowers, D.A.; Williams, F.R.; Witten, M.A.

    1989-01-01

    The inboard limiter of the Compact Ignition Tokamak (CIT) must protect the vacuum vessel from the plasma energy. This limiter region must withstand nominal heat fluxes in excess of 10 MW/m 2 and in addition it must be designed to be remotely maintained. Carbon-carbon composite material was selected over bulk graphite materials for the limiter design because of its ability to meet the thermal and structural requirements. The structural design concept consists of carbon-carbon composite panels attached to the vacuum vessel by a hinged rod/retainer concept. Results of the preliminary design study to define this inboard limiter are presented. The design concept is described along with the analyses of the thermal and structural response during nominal plasma operation and during plasma disruption events. 2 refs., 8 figs

  10. Remote maintenance considerations for swimming pool tokamak reactor

    International Nuclear Information System (INIS)

    Niikura, S.; Yamada, M.; Kasai, M.

    1983-01-01

    Swimming Pool Tokamak Reactor (SPTR) is one of the candidate devices which are expected to demonstrate physical and engineering feasibility for fusion power reactors. In SPTR, water shield is adopted instead of solid shield structures. Among the advantages of SPTR are, from viewpoint of remote maintenance, small handling weight and high space availability between TF coils and a vacuum vessel. On the other hand, high dose rate during reactor repair and adverse effects on remote maintenance equipment by the shielding water might be the disadvantage of SPTR, where it is assumed that the shielding water is drained during reactor repair. Since the design of SPTR is still at the preliminary stage, for remote maintenance, much effort has been directed to clarification of design conditions such as environment and handling weight. As for the remote maintenance system concepts, studies have been focussed on those for a vacuum vessel and its internal structure (blanket, divertor and protection walls) expected to be repaired more frequently. The vacuum vessel assembly is divided into 21 sectors and number of TF coils is 14. A pair of TF coils are connected with each other by antitorque beams on the whole side surface. Vacuum vessel cassettes and associated blanket, divertor and protection walls are replaced through seven windows between TF coils pairs. Therefore each vacuum vessel cassette is required moving mechanisms in toroidal and radial directions. Options for slide mechanisms are wheels, balls, rollers and water bearings. Options for driving the cassette are self-driving by hydraulic motors and external driving by rack-pinion, wires or specific vehicles. As a result of studies, the moving mechanism with wheels and hydraulic motors has been selected for the reference design, and the system with water bearings and rack-pinion as an alternative. Furthermore typical concepts have been obtained for remote maintenance equipment such as wall-mounted manipulators, tools for

  11. Acceleration mechanism of vertical displacement event and its amelioration in tokamak disruptions

    International Nuclear Information System (INIS)

    Nakamura, Yukiharu; Yoshino, Ryuji; Pomphrey, N.; Jardin, S.C.

    1996-01-01

    Vertical displacement events (VDEs), which are frequently observed in disruptive discharges of elongated tokamaks, are investigated using the Tokamak Simulation Code. We show that disruption events such as a sudden plasma pressure drop (β p collapse) and the subsequent plasma current quench (I p quench) can accelerate VDEs due to the adverse destabilizing effect of the resistive shell, which has previously been thought to stabilize VDEs. In a tokamak with a surrounding shell which is asymmetric with respect to the geometric midplane, the I p quench also causes an additional VDE acceleration due to the vertical imbalance of the attractive force. While the shell-geometry characterizes the VDE dynamics, the growth rate of VDEs depends strongly on the magnitude of the β p collapse, the speed of the I p quench and the n-index of the plasma equilibrium just before the disruption. An amelioration of I p quench-induced VDEs was experimentally established in the JT-60U tokamak by optimizing the vertical location of the plasma just prior to the disruption. The JT-60U vacuum vessel is shown to be suitable for preventing the β p collapse-induced VDE. (author)

  12. Technical diagnosis system for EAST tokamak

    International Nuclear Information System (INIS)

    Qian Jing; Weng, P.D.; Luo, J.R.; Chen, Z.M.; Wu, Y.

    2010-01-01

    Technical diagnosis system (TDS) is one of the important subsystems of EAST (experimental advanced superconducting tokamak) device, main function of which is to monitor status parameters in EAST device. Those status parameters include temperature of different positions of main components, resistance of each superconducting (SC) coils, joint resistance of SC coils and high-temperature superconducting (HTS) current leads, strain of cold-quality components endured force, and displacement and current of toroidal field (TF) coils in EAST device, which are analog input signals. In addition there are still some analog and digital output signals. The TDS monitors all of those signals in the period of EAST experiments. TDS data monitoring is described in detail for it plays important role during EAST campaign. And how to protect the SC magnet system during each plasma discharging is presented with data of temperature of coolant inlet and outlet of SC coils and feeders and cases of the TF coils and temperature in the upper and middle and bottom of the TF coil case. During construction of the TDS primary difficulties come from installation of Lakeshore Cernox temperature sensors, strain measurement of central solenoid coils support legs and installation of co-wound voltage sensors for quench detection. While during operation since the first commissioning big challenges are from temperature measurement changes in current leads and quench detection of PF coils. Those difficulties in both stages are introduced which are key to make the TDS reliable. Meanwhile analysis of experimental data like temperature as a back up to testify quench occurrence and stress on vacuum vessel thermal shield and vacuum vessel have also been discussed.

  13. Study of plasma discharge evolution and edge turbulence with fast visible imaging in the Aditya tokamak

    International Nuclear Information System (INIS)

    Banerjee, Santanu; Manchanda, R.; Chowdhuri, M.B.

    2015-01-01

    Study of discharge evolution through the different phases of a tokamak plasma shot viz., the discharge initiation, current ramp-up, current flat-top and discharge termination, is essential to address many inherent issues of the operation of a Tokamak. Fast visible imaging of the tokamak plasma can provide valuable insight in this regard. Further, edge turbulence is considered to be one of the quintessential areas of tokamak research as the edge plasma is at the immediate vicinity of the plasma core and plays vital role in the core plasma confinement. The edge plasma also bridges the core and the scrape off layer (SOL) of the tokamak and hence has a bearing on the particle and heat flux escaping the plasma column. Two fast visible imaging systems are installed on the Aditya tokamak. One of the system is for imaging the plasma evolution with a wide angle lens covering a major portion of the vacuum vessel. The imaging fiber bundle along with the objective lens is installed inside a radial re-entrant viewport, specially designed for the purpose. Another system is intended for tangential imaging of the plasma column. Formation of the plasma column and its evolution are studied with the fast visible imaging in Aditya. Features of the ECRH and LHCD operations on Aditya will be discussed. 3D filaments can, be seen at the plasma edge all along the discharge and they get amplified in intensity at the plasma termination phase. Statistical analysis of these filaments, which are essentially plasma blobs will be presented. (author)

  14. Preliminary design study of a steady state tokamak device

    International Nuclear Information System (INIS)

    Miya, Naoyuki; Nakajima, Shinji; Ushigusa, Kenkichi; and athors)

    1992-09-01

    Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)

  15. Magnetic Diagnostics for Equilibrium Reconstructions in the Presence of Nonaxisymmetric Eddy Current Distributions in Tokamaks

    International Nuclear Information System (INIS)

    Kaita, R.; Kozub, T.; Logan, N.; Majeski, R.; Menard, J.; Zakharov, L.

    2010-01-01

    The lithium tokamak experiment (LTX) is a modest-sized spherical tokamak (R 0 = 0.4 m and a = 0.26 m) designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 C. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.

  16. Design study of toroidal magnets for tokamak experimental power reactors

    International Nuclear Information System (INIS)

    Stekly, Z.J.J.; Lucas, E.J.

    1976-12-01

    This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed

  17. An interior vessel viewing system for DIII-D

    International Nuclear Information System (INIS)

    Senior, R.

    1989-11-01

    It was anticipated that there could be damage to the interior walls of the vacuum vessel during operations of the DIII-D tokamak. A method of viewing the inside of the vessel from the outside was required, that would allow the interior walls to be inspected visually for damage and to locate any debris resulting from operations. A miniature closed circuit television color camera system was developed which could be inserted into one of several ports of the vessel during a 'clean' vent, i.e., vented to inert gas. The system has pan, tilt and zoom capability and carries its own lighting. The use of this system allows a quick assessment of the condition of the vessel to be made under 'clean' vent conditions. This precludes the need for the permit process and manned entry into the vessel which would allow air inside the vessel. A permanent record of the inspection can then be made on video tape. The design and configuration of this camera system is presented and its use as a diagnostic tool discussed. 2 refs., 5 figs

  18. Determination of the plasma column shape in the Tokamak Novillo cross section by magnetic probes

    International Nuclear Information System (INIS)

    Hernandez Omana, A.

    1993-01-01

    The determination of plasma cross section shape in Tokamaks is an important diagnostic method for equilibrium conditions analysis. In this work, it is obtained a time dependent variation of the plasma column cross section in Novillo Tokamak. The experimental method is based on using one magnetic probe, which is installed inside of the vacuum vessel in a 1 mm. wall thickness stainless steel tube, in the protected region of the limiter shadow. The plasma column cross section is determined measuring the poloidal magnetic field produced by the plasma current. This method, now running for determining the plasma column shape, requires the measurement of magnetic present field outside plasma column. The measurements are carried out from a set of small coils, which are located inside the vacuum chamber in the radial and poloidal direction, so we can measure magnetic field with no current attenuations produced by the penetration time of the stainless steel vacuum chamber. The magnetic probe detect a real time variation of magnetic flux passing through them. In order to obtain the magnetic field values, it is required that the electric signals coming from the magnetic probe be integrated, this operation is carried out by active circuits located between the probe signal and one oscilloscope. The integrated signals can be exhibited photographed on the oscilloscope display. (Author)

  19. Design of the TFTR [Tokamak Fusion Test Reactor] maintenance manipulator

    International Nuclear Information System (INIS)

    Loesser, G. D.; Heitzenroeder, P.; Bohme, G.; Selig, M.

    1987-01-01

    The Tokamak Fusion Test Reactor (TFTR) plans to generate a total of 3 x 10 21 neutrons during its deuterium-tritium run period in 1900. This will result in high levels of radiation, especially within the TFTR vacuum vessel. The maintenance manipulator's mission is to assist TFTR in meeting Princeton Plasma Physics Laboratory's personnel radiation exposure criteria and in maintaining as-low-as-reasonably-achievable principals by limiting the radiation exposure received by operating and maintenance personnel. The manipulator, which is currently being fabricated and tested by Kernforschungszentrum Karlsruhe, is designed to perform limited, but routine and necessary, functions within the TFTR vacuum torus after activation levels within the torus preclude such functions being performed by personnel. These functions include visual inspection, tile replacement, housekeeping tasks, diagnostic calibrations, and leak detection. To meet its functional objectives, the TFTR maintenance manipulator is required to be operable in TFTR's very high vacuum environment (typically 2 x 10 -8 Torr). It must also be bakeable at 150 degree C and able to withstand the radiation environment

  20. 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)

  1. Remote maintenance design activities and research and development accomplishments for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1989-01-01

    The use of deuterium-tritium (D-T) fuel for the Compact Ignition Tokamak (CIT) requires the use of remote handling technology in order to carry out maintenance operations. The remote operations consist of removing and replacing such components as first wall armor protection tiles, radio-frequency (RF) heating modules, and diagnostic modules. The major pieces of equipment being developed for maintenance operations internal to the vacuum vessel include an articulated boom manipulator (ABM), an inspection manipulator, and special tooling. For operations external to the vessel, the equipment includes a bridge-mounted manipulator system, decontamination equipment, hot cell equipment, and solid radioactive waste (rad-waste) handling and packaging equipment. The CIT Project is completing the conceptual design phase; research and development (R and D) activities, which include demonstrations of remote maintenance operations on full-size partial mock-ups are under way. (orig.)

  2. Remote maintenance design activities and research and development accomplishments for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1988-01-01

    The use of deuterium-tritium (D-T) fuel for the Compact Ignition Tokamak (CIT) requires the use of remote handling technology to carry out maintenance operations. The remote operations consist of removing and replacing such components as first wall armor protection tiles, radio-frequency (rf) heating modules, and diagnostic modules. The major pieces of equipment being developed for maintenance activities internal to the vacuum vessel include an articulated boom manipulator (ABM), an inspection manipulator, and special tooling. For activities external to the vessel, the equipment includes a bridge-mounted manipulator system, decontamination equipment, hot cell equipment, and solid radiation-waste (rad-waste) handling and packaging equipment. The CIT Project is completing the conceptual design phase; research and development (R and D) activities, which include demonstrations of remote maintenance operations on full-size partial mock-ups are under way. 5 figs

  3. A Neutral Beam for the Lithium Tokamak eXperiment Upgrade (LTX-U)

    Science.gov (United States)

    Merino, Enrique; Majeski, Richard; Kaita, Robert; Kozub, Thomas; Boyle, Dennis; Schmitt, John; Smirnov, Artem

    2015-11-01

    Neutral beam injection into tokamaks is a proven method of plasma heating and fueling. In LTX, high confinement discharges have been achieved with low-recycling lithium walls. To further improve plasma performance, a neutral beam (NB) will be installed as part of an upgrade to LTX (LTX-U). The NB will provide core plasma fueling with up to 700 kW of injected power. Requirements for accommodating the NB include the addition of injection and beam-dump ports onto the vessel and enhancement of the vacuum vessel pumping capability. Because the NB can also serve as a source of neutrals for charge-exchange recombination spectroscopy, ``active'' spectroscopic diagnostics will also be developed. An overview of these plans and other improvements for upgrading LTX to LTX-U will be presented. Supported by US DOE contracts DE-AC02-09CH11466 and DE-AC52-07NA27344.

  4. Active and passive kink mode studies in a tokamak with a movable ferromagnetic wall

    Energy Technology Data Exchange (ETDEWEB)

    Levesque, J. P.; Hughes, P. E.; Bialek, J.; Byrne, P. J.; Mauel, M. E.; Navratil, G. A.; Peng, Q.; Rhodes, D. J.; Stoafer, C. C. [Department of Applied Physics and Applied Mathematics, Columbia University, 500 W. 120th Street, New York, New York 10027 (United States)

    2015-05-15

    High-resolution active and passive kink mode studies are conducted in a tokamak with an adjustable ferromagnetic wall near the plasma surface. Ferritic tiles made from 5.6 mm thick Hiperco{sup ®} 50 alloy have been mounted on the plasma-facing side of half of the in-vessel movable wall segments in the High Beta Tokamak-Extended Pulse device [D. A. Maurer et al., Plasma Phys. Controlled Fusion 53, 074016 (2011)] in order to explore ferritic resistive wall mode stability. Low-activation ferritic steels are a candidate for structural components of a fusion reactor, and these experiments examine MHD stability of plasmas with nearby ferromagnetic material. Plasma-wall separation for alternating ferritic and non-ferritic wall segments is adjusted between discharges without opening the vacuum vessel. Amplification of applied resonant magnetic perturbations and plasma disruptivity are observed to increase when the ferromagnetic wall is close to plasma surface instead of the standard stainless steel wall. Rapidly rotating m/n=3/1 external kink modes have higher growth rates with the nearby ferritic wall. Feedback suppression of kinks is still as effective as before the installation of ferritic material in vessel, in spite of increased mode growth rates.

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

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2013-01-01

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

  6. Recrystallized graphite utilization as the first wall material in Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.; Novokhatsky, A.N.; Petrov, Y.V.; Sakharov, N.V.; Terukov, E.I.; Trapeznikova, I.N.; Denisov, E.A.; Kurdumov, A.A.; Kompaniec, T.N.; Lebedev, V.M.; Litunovstkii, N.V.; Mazul, I.

    2007-01-01

    Full text of publication follows: Globus-M spherical tokamak, built at A.F. Ioffe Physico-Technical Institute in 1999 is the first Russian spherical tokamak and has the broad area of research in controlled fusion [1]. Besides small aspect ratio (A=1.5) the distinguishing feature of the tokamak is the powerful energy supply system and auxiliary heating, which give opportunity to reach high specific power deposition up to few W/cm 3 . The utmost plasma current density and B/R ratio among spherical tokamaks allow operation in the range of high plasma densities ∼ 10 20 m -3 . This feature results in big power density loads to the first wall due to small plasma-wall spacing. The area of the first wall amour was gradually increased during few years since 2003, and nowadays reaches almost 90% of the inner vessel surface faced to plasma. Plasma facing protecting tiles are manufactured from recrystallized graphite doped by different elements (Ti, Si, B). Additionally the plasma facing surface was protected by films deposited during boronization. The tendency of short time and long time scale plasma parameters variation are discussed including the plasma performance improvement with increase of protected area. Technology of tiles preparation before installation into the tokamak vessel is briefly described, as well as technology of plasma facing armor preparation before the plasma experiments. Few protecting tiles doped by different elements which were exposed to plasma fluxes of dissimilar power densities for a long time were extracted from the vacuum vessel. The analysis of tiles material (RGT-91) to hold (accumulate) deuterium was made. The distribution of absorbed deuterium concentration along poloidal coordinate was measured. The elementary composition of the films deposited on the tiles was studied by Rutherford back scattering technique and by nuclear resonance reaction method. Other modern methods of surface and structural analysis of material exposed to prolonged

  7. Recrystallized graphite utilization as the first wall material in Globus-M spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gusev, V.; Novokhatsky, A.N.; Petrov, Y.V.; Sakharov, N.V.; Terukov, E.I.; Trapeznikova, I.N. [A.F. IOFFE Physico-technical Institute, Russian Academy of Sciences, St Petersburg (Russian Federation); Denisov, E.A.; Kurdumov, A.A.; Kompaniec, T.N. [St. Petersburg State Univ., Research Institute of Physics (Russian Federation); Lebedev, V.M. [B.P. Konstantinov Nuclear Physics Institute, Russian Academy of Science, Gatchina (Russian Federation); Litunovstkii, N.V. [D.V. Efremov Institute of Electrophysical Apparatus, St.Petersburg (Russian Federation); Mazul, I. [Development of Plasma Facing Materials and Components Laboratory, EFREMOV INSTITUTE, St Petersbourg (Russian Federation)

    2007-07-01

    Full text of publication follows: Globus-M spherical tokamak, built at A.F. Ioffe Physico-Technical Institute in 1999 is the first Russian spherical tokamak and has the broad area of research in controlled fusion [1]. Besides small aspect ratio (A=1.5) the distinguishing feature of the tokamak is the powerful energy supply system and auxiliary heating, which give opportunity to reach high specific power deposition up to few W/cm{sup 3}. The utmost plasma current density and B/R ratio among spherical tokamaks allow operation in the range of high plasma densities {approx} 10{sup 20} m{sup -3}. This feature results in big power density loads to the first wall due to small plasma-wall spacing. The area of the first wall amour was gradually increased during few years since 2003, and nowadays reaches almost 90% of the inner vessel surface faced to plasma. Plasma facing protecting tiles are manufactured from recrystallized graphite doped by different elements (Ti, Si, B). Additionally the plasma facing surface was protected by films deposited during boronization. The tendency of short time and long time scale plasma parameters variation are discussed including the plasma performance improvement with increase of protected area. Technology of tiles preparation before installation into the tokamak vessel is briefly described, as well as technology of plasma facing armor preparation before the plasma experiments. Few protecting tiles doped by different elements which were exposed to plasma fluxes of dissimilar power densities for a long time were extracted from the vacuum vessel. The analysis of tiles material (RGT-91) to hold (accumulate) deuterium was made. The distribution of absorbed deuterium concentration along poloidal coordinate was measured. The elementary composition of the films deposited on the tiles was studied by Rutherford back scattering technique and by nuclear resonance reaction method. Other modern methods of surface and structural analysis of material

  8. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

    Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  9. Design of vessel baking system and thermal radiation shields for SST-1

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, E.R.; Nagabhushana, S.; Pathak, H.A.; Panigrahi, S.; Nath, T.R.; Babu, A.V.S; Gangradey, R.; Patel, R.J.; Saxena, Y.C. [Institute for Plasma Research, Gandhinagar (India)

    1998-07-01

    SST-1 is a Steady State Tokamak with a major radius of 1.1 m, minor radius of 0.2 m and toroidal field of 3.0 T. The toroidal and poloidal field coils of SST-1 are superconducting. One of the main objectives of SST-1 is to demonstrate steady state particle removal and active plasma density control which states the necessity of wall conditioning. The vacuum vessel will be baked up to 525 K by passing hot nitrogen gas through the U - channels welded on the inner surface of vacuum vessel. The required mass flow rate at 5 bar is 0.712 Kg/s to maintain 525 K wall temperature in steady state. Superconducting coils operating at 4.5 K will be protected against thermal radiation from hot surfaces using liquid nitrogen cooled panels operating at 87 K. Maximum 1200 litres/hour liquid nitrogen is required during vessel baking. The design of vacuum vessel baking system and thermal radiation shields and related flow analysis are presented here. (authors)

  10. Design of vessel baking system and thermal radiation shields for SST-1

    International Nuclear Information System (INIS)

    Kumar, E.R.; Nagabhushana, S.; Pathak, H.A.; Panigrahi, S.; Nath, T.R.; Babu, A.V.S; Gangradey, R.; Patel, R.J.; Saxena, Y.C.

    1998-01-01

    SST-1 is a Steady State Tokamak with a major radius of 1.1 m, minor radius of 0.2 m and toroidal field of 3.0 T. The toroidal and poloidal field coils of SST-1 are superconducting. One of the main objectives of SST-1 is to demonstrate steady state particle removal and active plasma density control which states the necessity of wall conditioning. The vacuum vessel will be baked up to 525 K by passing hot nitrogen gas through the U - channels welded on the inner surface of vacuum vessel. The required mass flow rate at 5 bar is 0.712 Kg/s to maintain 525 K wall temperature in steady state. Superconducting coils operating at 4.5 K will be protected against thermal radiation from hot surfaces using liquid nitrogen cooled panels operating at 87 K. Maximum 1200 litres/hour liquid nitrogen is required during vessel baking. The design of vacuum vessel baking system and thermal radiation shields and related flow analysis are presented here. (authors)

  11. Concept study of the Steady State Tokamak Reactor (SSTR)

    International Nuclear Information System (INIS)

    1991-06-01

    The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)

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

    Directory of Open Access Journals (Sweden)

    Luiz Gustavo Gonçalves Rodrigues

    2012-03-01

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

  13. Equilibrium Reconstruction in EAST Tokamak

    International Nuclear Information System (INIS)

    Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)

  14. Dust resuspension and transport modeling for loss of vacuum accidents

    International Nuclear Information System (INIS)

    Humrickhouse, P.W.; Corradini, M.L.; Sharpe, J.P.

    2007-01-01

    Plasma surface interactions in tokamaks are known to create significant quantities of dust, which settles onto surfaces and accumulates in the vacuum vessel. In ITER, a loss of vacuum accident may result in the release of dust which will be radioactive and/or toxic, and provides increased surface area for chemical reactions or dust explosion. A new method of analysis has been developed for modeling dust resuspension and transport in loss of vacuum accidents. The aerosol dynamic equation is solved via the user defined scalar (UDS) capability in the commercial CFD code Fluent. Fluent solves up to 50 generic transport equations for user defined scalars, and allows customization of terms in these equations through user defined functions (UDF). This allows calculation of diffusion coefficients based on local flow properties, inclusion of body forces such as gravity and thermophoresis in the convection term, and user defined source terms. The code accurately reproduces analytical solutions for aerosol deposition in simple laminar flows with diffusion and gravitational settling. Models for dust resuspension are evaluated, and code results are compared to available resuspension data, including data from the Toroidal Dust Mobilization Experiment (TDMX) at the Idaho National Laboratory. Extension to polydisperse aerosols and inclusion of coagulation effects is also discussed. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Guasp, J.; Liniers, M.

    1995-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Guasp, J

    1995-07-01

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

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

    Science.gov (United States)

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

    2017-12-01

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

  18. Quantitative spectroscopy of X-ray lines and continua in tokamaks

    International Nuclear Information System (INIS)

    Peacock, N.J.; Lawson, K.D.; Patel, A.; Barnsley, R.; Melnick, I.M.; O'Mullane, M.G.; Singleton, M.A.

    1996-10-01

    Crystal and synthetic multilayer diffractors, deployed either as flat Bragg reflectors, or curved, as in the Johann configuration, are used to study the spectrum of COMPASS-D and other tokamaks in the wavelength region 1-100 A. In this paper we concentrate on the measurement of absolute photon fluxes and the derivation of volume emissivities of the lines and continua in the X-ray region. The sensitivities of these instruments to absolute photon flux have been constructed ab initio from the individual component efficiencies, including published values of the diffractor reflectivities, which have been checked or supplemented by measurements using a double-axis goniometer or from line branching ratios. For these tokamak plasmas where the elemental abundances and effective ion charge are documented, the X-ray continuum intensity itself has been used as a calibration source to derive absolute instrument sensitivity, in reasonable agreement with the ab initio method. In the COMPASS-D tokamak, changes in the effective ion charge state, Z eff , have been derived for different operating conditions, from the absolute intensity of the continuum at ∼4 A. From the irradiances of the line emission, changes in the absolute level of impurities following ''boronisation'' of the vacuum vessel have also been documented. (UK)

  19. Recent results from the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petersen, P.I.

    1998-02-01

    The DIII-D national fusion research program focuses on establishing the scientific basis for optimization of the tokamak approach to fusion energy production. The symbiotic development of research, theory, and hardware continues to fuel the success of the DIII-D program. During the last year, a radiative divertor and a second cryopump were installed in the DIII-D vacuum vessel, an array of central and boundary diagnostics were added, and more sophisticated computer models were developed. These new tools have led to substantial progress in the understanding of the plasma. The authors now have a better understanding of the divertor as a means to manage the heat, particle, and impurity transport pumping of the plasma edge using the in situ divertor cryopumps effectively controls the plasma density. The evolution of diagnostics that probe the interior of the plasma, particularly the motional Stark effect diagnostic, has led to a better understanding of the core of the plasma. This understanding, together with tools to control the profiles, including electron cyclotron waves, pellet injection, and neutral beam injection, has allowed them to progress in making plasma configurations that give rise to both low energy transport and improved stability. Most significant here is the use of transport barriers to improve ion confinement to neoclassical values. Commissioning of the first high power (890 kW) 110 GHz gyrotron validates an important tool for managing the plasma current profile, key to maintaining the transport barriers. An upgraded plasma control system, ''isoflux control,'' which exploits real time MHD equilibrium calculations to determine magnetic flux at specified locations within the tokamak vessel and provides the means for precisely controlling the plasma shape and, in conjunction with other heating and fueling systems, internal profiles

  20. Tokamak experimental power reactor conceptual design. Volume I

    International Nuclear Information System (INIS)

    1976-08-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters

  1. Dust Measurements in Tokamaks

    International Nuclear Information System (INIS)

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-01-01

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 (micro)m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics

  2. Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

    International Nuclear Information System (INIS)

    Sato, Satoshi

    2003-09-01

    In tokamak-type DT nuclear fusion reactor, there are various type slits and ducts in the blanket and the vacuum vessel. The helium production in the rewelding location of the blanket and the vacuum vessel, the nuclear properties in the super-conductive TF coil, e.g. the nuclear heating rate in the coil winding pack, are enhanced by the radiation streaming through the slits and ducts, and they are critical concern in the shielding design. The decay gamma ray dose rate around the duct penetrating the blanket and the vacuum vessel is also enhanced by the radiation streaming through the duct, and they are also critical concern from the view point of the human access to the cryostat during maintenance. In order to evaluate these nuclear properties with good accuracy, three dimensional Monte Carlo calculation is required but requires long calculation time. Therefore, the development of the effective simple design evaluation method for radiation streaming is substantially important. This study aims to establish the systematic evaluation method for the nuclear properties of the blanket, the vacuum vessel and the Toroidal Field (TF) coil taking into account the radiation streaming through various types of slits and ducts, based on three dimensional Monte Carlo calculation using the MNCP code, and for the decay gamma ray dose rates penetrated around the ducts. The present thesis describes three topics in five chapters as follows; 1) In Chapter 2, the results calculated by the Monte Carlo code, MCNP, are compared with those by the Sn code, DOT3.5, for the radiation streaming in the tokamak-type nuclear fusion reactor, for validating the results of the Sn calculation. From this comparison, the uncertainties of the Sn calculation results coming from the ray-effect and the effect due to approximation of the geometry are investigated whether the two dimensional Sn calculation can be applied instead of the Monte Carlo calculation. Through the study, it can be concluded that the

  3. Tokamak with liquid metal toroidal field coil

    International Nuclear Information System (INIS)

    Ohkawa, T.; Schaffer, M.J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof

  4. Plasma formation and first OH experiments in GLOBUS-M tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Aleksandrov, S.V.; Burtseva, T.A.

    2001-01-01

    The paper reports results of experimental campaigns on plasma ohmic heating, performed during 1999-2000 on the spherical tokamak Globus-M. Later experimental results with tokamak fed by thyristor rectifiers are presented in detail. The toroidal magnetic field and plasma pulse duration in these experiments were significantly increased. The method of stray magnetic field compensation is described. The technology of vacuum vessel conditioning, including boronization of the vessel performed at the end of the experiments, is briefly discussed. Also discussed is the influence of ECR preioniziation on the breakdown conditions. Experimental data on plasma column formation and current ramp-up in different regimes of operation with the magnetic flux of the central solenoid (CS) limited to ∼100 mVs are presented. Ramp-up of the plasma current of 0.25 MA for the time interval ∼0.03 s with about 0.02 s flat-top at the toroidal field (TF) strength of 0.35 T allows the conclusion that power supplies, control system and wall conditioning work well. The same conclusion can be drawn from observation of plasma density behavior the density is completely controlled with external gas puff and the influence of the wall is negligible after boronization. The magnetic flux consumption efficiency is discussed. The results of magnetic equilibrium simulations are presented and compared with experiment. (author)

  5. Plasma position control in SST1 tokamak

    Indian Academy of Sciences (India)

    also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...

  6. Gas blanket fueling of a tokamak reactor

    International Nuclear Information System (INIS)

    Gralnick, S.L.

    1978-01-01

    The purpose of this paper is a speculative investigation of the potential of fueling a Tokamak by introducing a sufficiently large quantity of gaseous deuterium and tritium at the vacuum wall boundary. It is motivated by two factors: current generation tokamaks are, in a manner of speaking, fueled from the edge quite successfully as is evidenced by pulse lengths that are long compared to particle recycling times, and by rapid plasma density increase produced by gas puffing, alternative, deep penetration fueling techniques that have been proposed possess severe technological problems and large costs

  7. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  8. D-D tokamak reactor assessment

    International Nuclear Information System (INIS)

    Baxter, D.C.; Dabiri, A.E.

    1983-01-01

    A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

  9. 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

  10. Operation of an ITER relevant inspection robot on Tore Supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gargiulo, Laurent [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)], E-mail: laurent.gargiulo@cea.fr; Bayetti, Pascal; Bruno, Vincent; Hatchressian, Jean-Claude; Hernandez, Caroline; Houry, Michael [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Keller, Delphine [CEA, LIST, Service de Robotique Interactive, F-92265 Fontenay aux Roses (France); Martins, Jean-Pierre [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Measson, Yvan; Perrot, Yann [CEA, LIST, Service de Robotique Interactive, F-92265 Fontenay aux Roses (France); Samaille, Frank [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)

    2009-06-15

    Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. CEA has developed a multipurpose carrier able to realize deployments in the plasma vessel without breaking the Ultra High Vacuum (UHV) and temperature conditioning. A 6 years R and D programme was jointly conducted by CEA-LIST Interactive Robotics Unit and the Institute for Magnetic Fusion Research (IRFM) in order to demonstrate the feasibility and reliability of an in-vessel inspection robot relevant to ITER requirements. The Articulated Inspection Arm robot (AIA) is an 8-m long multilink carrier with a payload up to 10 kg operable between plasma under tokamak conditioning environment; its geometry allows a complete close inspection of Plasma Facing Components (PFCs) of the Tore Supra vessel. Different tools are being developed by CEA to be plugged at the front head of the carrier. The diagnostic presently in operation consists in a viewing system offering accurate visual inspection of PFCs. Leak detection of first wall based on helium sniffing and laser compact system for carbon co-deposited layers characterizations or treatments are also considered for demonstration. In April 2008, the AIA robot equipped with its vision diagnostic has realized a complete deployment into Tore Supra and the first closed inspection of the vessel under UHV conditions. During the upcoming experimental campaign, the same operation will be performed under relevant conditions (10{sup -6} Pa and 120 deg. C) after a conditioning phase at 200 deg. C to avoid outgassing pollution of the chamber. This paper describes the different steps of the project development, robot capabilities with the present operations conducted on Tore Supra and future requirements for making the robot a tool for tokamak routine operation.

  11. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

    The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)

  12. Design of a microwave calorimeter for the microwave tokamak experiment

    International Nuclear Information System (INIS)

    Marinak, M.

    1988-01-01

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs

  13. Vulcan: A steady-state tokamak for reactor-relevant plasma–material interaction science

    International Nuclear Information System (INIS)

    Olynyk, G.M.; Hartwig, Z.S.; Whyte, D.G.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Mumgaard, R.T.; Podpaly, Y.A.

    2012-01-01

    Highlights: ► A new scaling for obtaining reactor similarity in the divertor of scaled tokamaks. ► Conceptual design for a tokamak (“Vulcan”) to implement this new scaling. ► Demountable superconducting coils and compact neutron shielding. ► Helium-cooled high-temperature vacuum vessel and first wall. ► High-field-side lower hybrid current drive for non-inductive operation. - Abstract: An economically viable magnetic-confinement fusion reactor will require steady-state operation and high areal power density for sufficient energy output, and elevated wall/blanket temperatures for efficient energy conversion. These three requirements frame, and couple to, the challenge of plasma–material interaction (PMI) for fusion energy sciences. Present and planned tokamaks are not designed to simultaneously meet these criteria. A new and expanded set of dimensionless figures of merit for PMI have been developed. The key feature of the scaling is that the power flux across the last closed flux surface P/S ≃ 1 MW m −2 is to be held constant, while scaling the core volume-averaged density weakly with major radius, n ∼ R −2/7 . While complete similarity is not possible, this new “P/S” or “PMI” scaling provides similarity for the most critical reactor PMI issues, compatible with sufficient current drive efficiency for non-inductive steady-state core scenarios. A conceptual design is developed for Vulcan, a compact steady-state deuterium main-ion tokamak which implements the P/S scaling rules. A zero-dimensional core analysis is used to determine R = 1.2 m, with a conventional reactor aspect ratio R/a = 4.0, as the minimum feasible size for Vulcan. Scoping studies of innovative fusion technologies to support the Vulcan PMI mission were carried out for three critical areas: a high-temperature, helium-cooled vacuum vessel and divertor design; a demountable superconducting toroidal field magnet system; and a steady-state lower hybrid current drive system

  14. A method for measuring plasma position in the rectangular tokamak TJ-I

    International Nuclear Information System (INIS)

    Qin, J.; Ascasibar, E.; Pastor, I.; Navarro, A.P.; Ochando, M.A.; Pedrosa, M.A.; Rodriguez, L.; Sanchez, J.

    1993-01-01

    Determinations of plasma position in tokamaks are based on measuring the external magnetic field of the plasma current. Usually, the measurements are performed with magnetic coils along a closed contour around the plasma column, or with small magnetic probes in pairs positioned diametrically. However, such measurements are constrained in TJ-I because of a severe restriction on installing the magnetic probes. It is only possible to install some small magnetic probes inside the vacuum vessel near its top and bottom walls where the limiters stand. In this paper an alternative method of measuring plasma position in TJ-I is presented. The plasma position can be derived from the measurements of those magnetic probes located off the equatorial plane, with the aid of a filament model. The method is proved to be reliable, and would be applied to those small-size devices with similar restrictions. (author) 5 refs., 3 figs., 1 tab

  15. Deuterium retention and surface modification of tungsten macrobrush samples exposed in FTU Tokamak

    Science.gov (United States)

    Maddaluno, G.; Giacomi, G.; Rufoloni, A.; Verdini, L.

    2007-06-01

    The effect of discrete structures such as macrobrush or castellated surfaces on power handling and deuterium retention of plasma facing components is to be assessed since such geometrical configurations are needed for increasing the lifetime of the armour to heat-sink joint. Four small macrobrush W and W + 1%La2O3 samples have been exposed in the Frascati Tokamak Upgrade (FTU) scrape-off layer up to the last closed flux surface by means of the Sample Introduction System. FTU is an all metal machine with no carbon source inside vacuum vessel; it exhibits ITER relevant energy and particle fluxes on the plasma facing components. Here, results on morphological surface changes (SEM), chemical composition (EDX) and deuterium retention (TDS) are reported.

  16. Analysis and design of the Alfven wave antenna system for the SUNIST spherical tokamak

    International Nuclear Information System (INIS)

    Tan Yi; Gao Zhe; He Yexi

    2009-01-01

    Analysis and design of the Alfven wave antenna system for the SUNIST spherical tokamak are presented. Two candidate antenna concepts, folded and unfolded, are analyzed and compared with each other. In the frequency range of Alfven resonance the impedance spectrums of both two concept antennas for major modes are numerically calculated in a 1-D MHD framework. The folded concept is chosen for engineering design. The antenna system is designed to be simple and requires least modification to the vacuum vessel. The definition of the antenna shape is guided by the analyses with constraints of existing hardware layouts. Each antenna unit consists of two stainless steel straps with a thickness of 1 mm. A number of boron nitride tiles are assembled together as the side limiters for plasma shielding. Estimation shows that the structure is robust enough to withstand the electromagnetic force and the heat load for typical discharge duty cycles.

  17. A simple ideal magnetohydrodynamical model of vertical disruption events in tokamaks

    International Nuclear Information System (INIS)

    Fitzpatrick, R.

    2009-01-01

    A simple model of axisymmetric vertical disruption events (VDEs) in tokamaks is presented in which the halo current force exerted on the vacuum vessel is calculated directly from linear, marginally stable, ideal-magnetohydrodynamical (MHD) stability analysis. The basic premise of the model is that the halo current force modifies pressure balance at the edge of the plasma, and therefore also modifies ideal-MHD plasma stability. In order to prevent the ideal vertical instability, responsible for the VDE, from growing on the very short Alfven time scale, the halo current force must adjust itself such that the instability is rendered marginally stable. The model predicts halo currents which are similar in magnitude to those observed experimentally. An approximate nonaxisymmetric version of the model is developed in order to calculate the toroidal peaking factor for the halo current force.

  18. Experimental survey of the L-H transition conditions in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Gohil, P.; Watkins, J.C.

    1994-01-01

    We present the global analysis of a recent survey of the H-mode power threshold in DIII-D using D o → D + NBI after boronization of the vacuum vessel. Single parameter scans of B T , I p , density, and plasma shape have been carried out on the DIII-D tokamak for neutral beam heated single-null and double-null diverted plasmas. In single-null discharges, the power threshold is found to increase approximately linearly with B T and n e but remains independent of I p . In double-null discharges, the power threshold is found to be approximately independent of both B T and n e . Various shape parameters such as plasma-wall gaps had only a weak effect on the power threshold. Imbalancing the double null configuration resulted in a large increase in the threshold power

  19. Configuration studies for a small-aspect-ratio tokamak stellarator hybrid

    International Nuclear Information System (INIS)

    Carreras, B.A.; Lynch, V.E.; Ware, A.

    1996-08-01

    The use of modulated toroidal coils offers a new path to the tokamak-stellarator hybrids. Low-aspect-ratio configurations can be found with robust vacuum flux surfaces and rotational transform close to the transform of a reverse-shear tokamak. These configurations have clear advantages in minimizing disruptions and their effect and in reducing tokamak current drive needs. They also allow the study of low-aspect-ratio effects on stellarator confinement in small devices

  20. Characteristics of disruptive plasma current decay in the HT-2 tokamak

    International Nuclear Information System (INIS)

    Abe, Mitsushi; Takeuchi, Kazuhiro; Otsuka, Michio

    1993-01-01

    Motions of plasma current channel and time evolutions of eddy current distribution on the vacuum vessel during disruptive plasma current decay were studied experimentally in the Hitachi tokamak HT-2. The plasmas are vertically elongated and circularly shaped plasmas. A disruptive plasma current decay has three phases. During the first phase, a large displacement of the plasma position without plasma current decay is observed. Rapid plasma current decay is observed during the second phase and the decay rate is roughly constant with time. The eddy current distribution is like that due to the shell effect which creates a poloidal field to reduce the plasma displacement. During the third phase, the plasma current decays exponentially. The second phase is observed in slightly elongated and high plasma current (> 20 kA) circularly shaped plasmas. The plasma current decay rates in the second phase depend on the plasma cross sectional shape, but they do not in the third phase. The magnetic axis moves from the plasma area to the vacuum vessel wall between the second and third phases. (author)

  1. Electromagneto-mechanical coupling analysis of a test module in J-TEXT Tokamak during plasma disruption

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Haijie; Yuan, Zhensheng; Yuan, Hongwei; Pei, Cuixiang [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Chen, Zhenmao, E-mail: chenzm@mail.xjtu.edu.cn [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Yang, Jinhong; Wang, Weihua [Institute of Applied Physics of AOA, Hefei 230031 (China)

    2016-11-01

    In this paper, the dynamic response during plasma disruption of a test blanket module in vacuum vessel (VV) of the Joint TEXT (J-TEXT), which is an experimental Tokamak device with iron core, was simulated by applying a program developed by authors on the ANSYS platform using its parametric design language (APDL). The moving coordinate method as well as the load transfer and sequential coupling strategy were adopted to cope with the electromagneto-mechanical coupling effect. To establish the numerical model, the influence of the iron core on the eddy current and electromagnetic (EM) force during disruption was numerically investigated at first and the influence was found not significant. Together with the geometrical features of the J-TEXT Tokamak structure, 180° sector models without magnetic core were finally established for the EM field and the structural response simulations. To obtain the source plasma current, the plasma current evolution during disruption was simulated by using the