Saint Petersburg, Russia
Saint Petersburg, Russia

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Imbeaux F.,French Atomic Energy Commission | Citrin J.,EURATOM | Hobirk J.,Max Planck Institute for Plasma Physics (Garching) | Hogeweij G.M.D.,EURATOM | And 38 more authors.
Nuclear Fusion | Year: 2011

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H96-L = 0.6 or HIPB98 = 0.4) has been validated on a multi-machine experimental dataset for predicting the li dynamics within ±0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than ±0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of Ip = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down. © 2011 IAEA, Vienna.


Riabov G.,RAS Petersburg Nuclear Physics Institute | Artamonov S.,RAS Petersburg Nuclear Physics Institute | Ivanov E.,RAS Petersburg Nuclear Physics Institute | Mikheev G.,RAS Petersburg Nuclear Physics Institute | And 4 more authors.
RuPAC 2012 Contributions to the Proceedings - 23rd Russian Particle Accelerator Conference | Year: 2012

The history of the design and costruction of the 80 MeV H- isochronous cyclotron as well as some design features are discribed. Copyright © 2012 by the respective authors.


Svistunov Y.,NIIEFA | Durkin A.,Russian Academy of Sciences | Ovsyannikov A.D.,Saint Petersburg State University
RuPAC 2012 Contributions to the Proceedings - 23rd Russian Particle Accelerator Conference | Year: 2012

Modeling results for deuteron dynamics in RFQ structure with operational frequency 433 MHz and 1 MeV output energy are presented. The results are compared with experimental data. The purpose of investigation is to find optimal input RFQ emittance parameters for offnominal values of input current and vane voltage. Copyright © 2012 by the respective authors.


Hirai T.,ITER Organization | Escourbiac F.,ITER Organization | Carpentier-Chouchana S.,Sogeti Inc. | Durocher A.,ITER Organization | And 13 more authors.
Physica Scripta | Year: 2014

The full tungsten divertor qualification program was defined for the R&D activity in domestic agencies. The qualification program consists of two steps: (i) technology development and validation and (ii) a full-scale demonstration. Small-scale mock-ups were manufactured in Japanese and European industries and delivered to the ITER divertor test facility in Russia for high heat flux testing. In parallel activity to the qualification program, both domestic agencies demonstrated that W monoblock technologies withstanding up to 20 MW m-2 were available. © 2014 The Royal Swedish Academy of Sciences.


Svistunov Y.A.,Saint Petersburg State University | Kudinovich I.V.,Saint Petersburg State University | Golovkina A.G.,Saint Petersburg State University | Ovsyannikov D.A.,Saint Petersburg State University | And 2 more authors.
Problems of Atomic Science and Technology | Year: 2014

The problems of target choice for compact ADS with reactor thermal power 200... 400 MW and 200... 400 MeV proton beam are considered. Simulation results of neutron yield from fissile and non-fissile targets are presented and the optimal target sizes are calculated. The principal target design characteristics and its thermal condition are also considered.


Udintsev V.S.,ITER Organization | Maquet P.,ITER Organization | Alexandrov E.,Russian Federation Domestic Agency | Casal N.,ITER Organization | And 24 more authors.
Fusion Engineering and Design | Year: 2015

The Diagnostic Generic Equatorial Port Plug (GEPP) is designed to be common to all equatorial port-based diagnostic systems. It is designed to survive throughout the lifetime of ITER for 20 years, 30,000 discharges, and 3000 disruptions. The EPP structure dimensions (without Diagnostic First Walls and Diagnostic Shield Modules) are L2.9 × W1.9 × H2.4 m3. The length of the fully integrated EPP is 3174 mm. The weight of the EPP structure is about 15 t, whereas the total weight of the integrated EPP may be up to 45 t. The EPP structure provides a flexible platform for a variety of diagnostics. The Diagnostic Shield Module assemblies, or drawers, allow a modular approach with respect to diagnostic integration and maintenance. In the nuclear phase of ITER operations, they will be remotely inserted into the EPP structure in the Hot Cell Facility. The port plug structure must also contribute to the nuclear shielding, or plugging, of the port and further contain circulated water to allow cooling during operation and heating during bake-out. The Final Design of the GEPP has been successfully passed in late 2013 and is now heading toward manufacturing. The final design of the GEPP includes interfaces, manufacturing, R&D, operation and maintenance, load cases and analysis of failure modes. © 2015 Elsevier B.V.


Leonov V.M.,RAS Research Center Kurchatov Institute | Gribov Yu.V.,ITER Organization | Kavin A.A.,NIIEFA | Khayrutdinov R.R.,RAS Research Center Kurchatov Institute | And 2 more authors.
37th EPS Conference on Plasma Physics 2010, EPS 2010 | Year: 2010

Development of the operational scenarios and analysis of conditions influenced the plasma performance in different discharge stages are the important aspects of the ITER design. Results of previous studies of plasma termination in ITER 15 MA DT inductive scenario are presented in [1-3], where the main attention has been focused on the analysis of operation of the Poloidal Field (PF) system. This paper presents results of further complex study of conditions in the termination stage of the reference ITER 15 MA inductive scenario to consider the most important peculiarities of this stage and to optimize plasma parameters behaviour during this stage.


Kocan M.,ITER Organization | Pitts R.A.,ITER Organization | Gribov Y.,ITER Organization | Bruno R.,ITER Organization | And 6 more authors.
40th EPS Conference on Plasma Physics, EPS 2013 | Year: 2013

The study here demonstrates that the expected surface heat fluxes in the nominal current ramp down scenario for ITER Baseline QDT = 10 inductive operation will be well within the power handling margins of the upper FWPs. The heat fluxes are a factor 2 lower than the power handling margin in the early full-bore ramp down phase and a factor 3 lower when the secondary strike point crosses the gap between the FWPs 8 and 9. Copyright © (2013) by the European Physical Society (EPS).


Ezato K.,Japan Atomic Energy Agency | Suzuki S.,Japan Atomic Energy Agency | Seki Y.,Japan Atomic Energy Agency | Mohri K.,Japan Atomic Energy Agency | And 4 more authors.
Fusion Engineering and Design | Year: 2015

Japan Atomic Energy Agency (JAEA) is in progress for technology qualification toward full-tungsten (W) ITER divertor outer vertical target (OVT), especially, tungsten monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m2. To demonstrate the armor heat sink bonding technology and heat removal capability, 6 small-scale W monoblock mock-ups manufactured by different bonding technologies using different W materials in addition to 4 full-scale prototype plasma-facing units (PFUs). After non-destructive test, the W components were tested under high heat flux (HHF) in ITER Divertor Test Facility (IDTF) at NIIEFA. Consequently, all of the W monoblocks endured the repetitive heat load at 20 MW/m2 for 1000 cycles (requirements 20 MW/m2 for 300 cycles) without any failure. In addition to the armor to heat sink joints, the load carrying capability test on the W monoblock with a leg attachment was carried out. In uniaxial tensile test, all of the W monoblock attachments with different bonding technologies such as brazing and HIPping withstand the tensile load exceeding 20 kN that is the value more than twice the design value. The failures occurred at the leg attachments or the W monoblocks, rather than the bonding interface of the W monoblocks to the leg attachment. © 2015 Elsevier B.V. All rights reserved.


Brusova N.I.,Russian Academy of Sciences | Feschenko A.,Russian Academy of Sciences | Grekhov O.,Russian Academy of Sciences | Kalinin Yu.,Russian Academy of Sciences | And 5 more authors.
RuPAC 2012 Contributions to the Proceedings - 23rd Russian Particle Accelerator Conference | Year: 2012

The activity for beam intensity increasing and beam use efficiency improvement is under progress in INR linac. An important stage is the development and implementation of the Beam Pulse Separation System in the accelerator intermediate extraction area (160 MeV). The system is intended for distribution the beam pulses between Isotope Production Facility (up to 160 MeV) and the Experimental Facility located downstream of the accelerator exit. The report describes the upgrade of intermediate extraction area as well as the first results of experiments with the beam. Copyright © 2012 by the respective authors.

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