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Sample records for international tokamak reactor

  1. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1981-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and USSR. The Zero-Phase of the INTOR Workshop, which was conducted during 1979, assessed the technical data base that would support the construction of the next major device in the tokamak program to operate in the early 1990s and defined the objectives and characteristics of this device. The INTOR workshop was extended into phase-1, the Definition Phase, in early 1980. The objective of the Phase-1 Workshop was to develop a conceptual design of the INTOR experiment. The purpose of this paper is to give an overview of the work of the Phase-1 INTOR Workshop (January 1980-June 1981, with emphasis upon the conceptual design

  2. TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Gilleland, J.R.

    1988-01-01

    The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

  3. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

  4. Prospects for Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.

    1995-01-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

  5. Cost of generating tritium internal and external to a tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Crotzer, M.E.; Heck, F.M.; Steinke, K.C.

    1981-01-01

    The costs associated with producing tritium internal and external to a thorium-based tokamak hybrid are estimated for a number of scenarios and the resulting impact on the symbiotic system cost of electricity calculated. For tritium generation within the hybrid, both continuous and batch production is analyzed. For external production, the lithium-bearing blanket is replaced with thorium and the tritium is generated in the client fission reactors. Continuous tritium production within the hybrid is found to increase the cost of electricity from 1.4 to 4.0 mills/kW-h. Batch tritium production can increase the cost of electricity by 10 mills/kW-h. Producing tritium outside the hybrid, and thereby enhancing client support, increases the cost of electricity from 1.8 to 4.1 mills/kW-h

  6. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

    1981-08-01

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  7. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  8. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  9. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  10. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  11. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    1975-06-01

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  12. Starfire: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

    1979-01-01

    The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

  13. Channels in tokamak reactor shields

    International Nuclear Information System (INIS)

    Shchipakin, O.L.

    1981-01-01

    The results of calculations of neutron transport through the channels in the tokamak reactor radiation shields, obtained by the Monte Carlo method and by the method of discrete ordinates, are considered. The given data show that the structural materials of the channel and that of the blanket and shields in the regions close to it are subjected to almost the same irradiation as the first wall and therefore they should satisfy the technical requirements. The radiation energy release in the injector channel wall, caused by neutron shooting, substantially depends on the channel dimensions. At the channel large diameter (0.7-10 m) this dependence noticeably decreases. The investigation of the effect of the injector channel cross section form on the neutron flux density through the channel, testifies to weak dependence of shooting radiation intensity on the form of the channel cross section. It is concluded that measures to decrease unfavourable effect of the channels on the safety of the power tokamak reactor operation and maintenance cause substantial changes in reactor design due to which the channel protection must be developed at first stages. The Monte Carlo method is recommended to be used for variant calculations and when calculating the neutron flux functionals in specific points of the system the discrete ordinate method is preferred [ru

  14. Tokamak fusion reactor

    International Nuclear Information System (INIS)

    Nohara, Kiyohiko

    2009-01-01

    The structural material is one of key issues for the development of reliable superconducting magnets and peripheral equipments of fusion reactors. Standard stainless steels like SUS 304 and 316 steels available at present do not meet requirements. We are developing a new austenitic steel that has proposed target properties named 'JAERI BOX'. Additions of N and V at different amounts were tested to improve strength and fracture toughness of a base alloy SUS316LN at 4.2 K. Mechanical properties of the developed steel were examined. It is found that the charpy absorbed energy and the fracture toughness of the developed steel at 4.2 K are within JAERI BOX. (T.I.)

  15. Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

    International Nuclear Information System (INIS)

    Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

    2009-01-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

  16. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R.S. [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  17. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  18. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  19. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  20. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

    2016-01-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO 2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO 2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO 2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO 2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO 2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  1. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  2. Near-wall plasma in a tokamak thermonuclear reactor

    International Nuclear Information System (INIS)

    Pistunovich, V.I.

    1991-01-01

    The development of thermonuclear reactor projects such as the national ones GTRT and OTR (USSR), UWMAK-NUMAK and TIBER (USA), NET (Euratom), FER (Japan), as well as the international ones INTOR and JET allows one to determine in detail the problems arising in the creation of a reactor. One of these problems is extracting heat and particles from the reactor and removing thermonuclear reaction ash, which in the case of a DT reactor is helium. The solution of this problem is connected with the investigation of interaction processes of the plasma with the materials from which the reactor is constructed. Progress in studying the interaction of plasma with a material surface has given rise to a new division of plasma physics, which includes not only reactor problems, but also plasma-physical processes occurring in the region between the hot plasma and the first wall. Experiments on tokamaks have shown that the phenomena occurring in the wall plasma and in the hot zone of the tokamak once strongly correlated. In and number of regimes the confinement of the hot plasma is determined by the conditions in the peripheral region. A survey of the near-well plasma problem in a tokamak thermonuclear reactor is given. The results of recent investigations are presented, and the status of the problem as applied to the ITER reactor project is described

  3. Optimization of steady-state beam-driven tokamak reactors

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; Singer, C.E.

    1983-01-01

    Recent developments in neutral beam technology prompt us to reconsider the prospects for steady-state tokamak reactors. A mathematical reactor model is developed that includes the physics of beam-driven currents and reactor power balance, as well as reactor and beam system costs. This model is used to find the plasma temperatures that minimize the reactor cost per unit of net electrical output. The optimum plasma temperatures are nearly independent of β and are roughly twice as high as the optimum temperatures for ignited reactors. If beams of neutral deuterium atoms with near-optimum energies of 1 to 2 MeV are used to drive the current in a reactor the size of the International Tokamak Reactor, then the optimum temperatures are typically T /SUB e/ approx. = 12 to 15 keV and T /SUB i/ approx. = 17 to 21 keV for a wide range of model parameters. Net electrical output rises rapidly with increasing deuterium beam energy for E /SUB b/ less than or equal to 400 keV, but rises only slowly above E /SUB b/ about 1 MeV. We estimate that beam-driven steady-state reactors could be economically competitive with pulsed-ignition reactors if cyclic-loading problems limit the toroidal magnetic field strength of pulsed reactors to less than or equal to 85% of that allowed in steady-state reactors

  4. The ARIES tokamak fusion reactor study

    International Nuclear Information System (INIS)

    Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

    1989-01-01

    The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

  5. The ICRH tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Perkins, F.W.

    1976-01-01

    A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

  6. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-01-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)

  7. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  8. Contour analysis of steady state tokamak reactor performance

    International Nuclear Information System (INIS)

    Devoto, R.S.; Fenstermacher, M.E.

    1990-01-01

    A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

  9. Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

    International Nuclear Information System (INIS)

    Mitari, O.; Hirose, A.; Skarsgard, H.M.

    1989-01-01

    In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

  10. Burn cycle requirements comparison of pulsed and steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Brooks, J.N.; Ehst, D.A.

    1983-12-01

    Burn cycle parameters and energy transfer system requirements were analyzed for an 8-m commercial tokamak reactor using four types of cycles: conventional, hybrid, internal transformer, and steady state. Not surprisingly, steady state is the best burn mode if it can be achieved. The hybrid cycle is a promising alternative to the conventional. In contrast, the internal transformer cycle does not appear attractive for the size tokamak in question

  11. Compact tokamak reactors part 2 (numerical results)

    International Nuclear Information System (INIS)

    Wiley, J.C.; Wootton, A.J.; Ross, D.W.

    1996-01-01

    The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced

  12. Compact tokamak reactors part 2 (numerical results)

    Energy Technology Data Exchange (ETDEWEB)

    Wiley, J.C.; Wootton, A.J.; Ross, D.W.

    1996-10-21

    The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced.

  13. How much does a tokamak reactor cost?

    Science.gov (United States)

    Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

    2017-10-01

    The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

  14. Tokamak reactor cost model based on STARFIRE/WILDCAT costing

    International Nuclear Information System (INIS)

    Evans, K. Jr.

    1983-03-01

    A cost model is presented which is useful for survey and comparative studies of tokamak reactors. The model is heavily based on STARFIRE and WILDCAT costing guidelines, philosophies, and procedures and reproduces the costing for these devices quite accurately

  15. Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor

    Science.gov (United States)

    Singh, M. J.; De Esch, H. P. L.

    2010-01-01

    This paper describes the physics design of a 100 keV, 60 A H- accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

  16. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1994-01-01

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  17. The 'FINTOR 1' design. A minimum size Tokamak Experimental Reactor

    International Nuclear Information System (INIS)

    Bertolini, E.

    1978-01-01

    The work carried on by the FINTOR (Frascati, Ispra, Napoli, Tokamak Reactor) group has now reached a stage where the effects of the main physics and engineering constraints, for a minimum size Tokamak Experimental Reactor have been clearly identified. This phase, now completed, has allowed to produce a self-consistent design of each basic component of the reactor, FINTOR 1, and to identify the more relevant interface problems toward a further optimisation of the reactor dimensions and characteristics to be performed in the future (FINTOR 2)

  18. Plasma startup patterns in tokamak reactors

    International Nuclear Information System (INIS)

    Maki, Koichi; Tone, Tatsuzo.

    1983-01-01

    Plasma startup patterns are studied from the viewpoint of net power loss represented by the total power loss less the α-particle heating power. The existence is shown of a critical temperature of plasma at which the net power loss becomes independent of plasma density. Observations are made which indicate that the net power loss decreases with lowering plasma density in the range below the critical temperature and vice versa, whether governed by empirical or trapped-ion scaling laws. A startup pattern is presented which minimizes the net power loss during startup, and which prescribes that: (1) The plasma density should be kept as low as possible until the plasma is heated up to the critical temperature; (2) thereafter, the plasma density should be increased to its steady state value while retaining the critical temperature; and (3) finally, with the density kept constant, the temperature should be further raised to its steady state value. The net power loss at critical temperature represents the lower limit of heating power required to bring the plasma to steady state in tokamak reactors. (author)

  19. A systems analysis of the ARIES tokamak reactors

    International Nuclear Information System (INIS)

    Bathke, C.G.

    1992-01-01

    The multi-institutional ARIES study has completed a series of cost-of-electricity optimized conceptual designs of commercial tokamak fusion reactors that vary the assumed advances in technology and physics. A comparison of these designs indicates the cost benefit of various design options. A parametric systems analysis suggests a possible means to obtain a marginally competitive fusion reactor

  20. ITER [International Thermonuclear Experimental Reactor] reactor building design study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

  1. Conceptual design of a commercial tokamak reactor using resistive magnets

    International Nuclear Information System (INIS)

    LeClaire, R.J. Jr.

    1988-01-01

    The future of the tokamak approach to controlled thermonuclear fusion depends in part on its potential as a commercial electricity-producing device. This potential is continually being evaluated in the fusion community using parametric, system, and conceptual studies of various approaches to improving tokamak reactor design. The potential of tokamaks using resistive magnets as commercial electricity-producing reactors is explored. Parametric studies have been performed to examine the major trade-offs of the system and to identify the most promising configurations for a tokamak using resistive magnets. In addition, a number of engineering issues have been examined including magnet design, blanket/first-wall design, and maintenance. The study indicates that attractive design space does exist and presents a conceptual design for the Resistive Magnet Commercial Tokamak Reactor (RCTR). No issue has been identified, including recirculating power, that would make the overall cost of electricity of RCTR significantly different from that of a comparably sized superconducting tokamak. However, RCTR may have reliability and maintenance advantages over commercial superconducting magnet devices

  2. Impurity control in near-term tokamak reactors

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Smith, D.L.; Brooks, J.N.

    1976-10-01

    Several methods for reducing impurity contamination in near-term tokamak reactors by modifying the first-wall surface with a low-Z or low-sputter material are examined. A review of the sputtering data and an assessment of the technological feasibility of various wall modification schemes are presented. The power performance of a near-term tokamak reactor is simulated for various first-wall surface materials, with and without a divertor, in order to evaluate the likely effect of plasma contamination associated with these surface materials

  3. Conceptual analysis of a tokamak reactor with lithium dust jet

    International Nuclear Information System (INIS)

    Kuteev, B.V.; Krylov, S.V.; Sergeev, V.Yu.; Skokov, V.G.; Timokhin, V.M.

    2010-01-01

    The steady-state operation of tokamak reactors requires radiating a substantial part of the fusion energy dissipated in plasma to make more uniform the heat loads onto the first wall and to reduce the erosion of the divertor plates. One of the approaches to realize this goal uses injection of lithium dust jet into the scrape-off layer (SOL). A quantitative conceptual analysis of the reactor parameters with lithium dust jet injection is presented here. The effects of the lithium on the core and SOL plasma are considered. The first results of developing the lithium jet injection technology and its application to the T-10 tokamak are also presented.

  4. Start of the international tokamak physics activity

    International Nuclear Information System (INIS)

    Campbell, D.

    2001-01-01

    This newsletter comprises a summary on the start of the International Tokamak Physics activity (ITPA) by Dr. D. Campbell, Chair of the ITPA Co-ordinating Committee. As the ITER EDA drew to a close, it became clear that it was desirable to establish a new mechanism in order to promote the continued development of the physics basis for burning plasma experiments and to preserve the invaluable collaborations between the major international fusion communities which had been established through the ITER physics expert groups. As a result of the discussions of the representatives of the European Union, Japan, the Russian Federation and the United States the agreed principles for conducting the International Tokamak Physics Activity (ITPA) were elaborated and ITPA topical physics groups were organized

  5. Comparative study of cost models for tokamak DEMO fusion reactors

    International Nuclear Information System (INIS)

    Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Ban, Kanae; Kondo, Takuya; Tobita, Kenji; Goto, Takuya

    2012-01-01

    Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study. (author)

  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. Revised design for the Tokamak experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1977-03-01

    A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost.

  8. 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

  9. 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

  10. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.

    2001-01-01

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  11. Commercial tokamak reactors with resistive toroidal field magnets

    International Nuclear Information System (INIS)

    Bombery, L.; Cohn, D.R.; Jassby, D.L.

    1984-01-01

    Scaling relations and design concepts are developed for commercial tokamak reactors that use watercooled copper toroidal field (TF) magnets. Illustrative parameters are developed for reactors that are scaled up in size from LITE test reactor designs, which use quasi-continuous copper plate magnets. Acceptably low magnet power requirements may be attainable in a moderate beta (β = 0.065) commercial reactor with a major radius of 6.2 m. The shielding thickness and magnet size are substantially reduced relative to values in commercial reactors with superconducting magnets. Operation at high beta (β = 0.14) leads to a reduction in reactor size, magnet-stored energy, and recirculating power. Reactors using resistive TF magnets could provide advantages of physically smaller devices, improved maintenance features, and increased ruggedness and reliability

  12. Technology issues for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

    The approach for decommissioning the Tokamak Fusion Test Reactor has evolved from a conservative plan based on cutting up and burying all of the systems, to one that considers the impact tritium contamination will have on waste disposal, how large size components may be used as their own shipping containers, and even the possibility of recycling the materials of components such as the toroidal field coils and the tokamak structure. In addition, the project is more carefully assessing the requirements for using remotely operated equipment. Finally, valuable cost database is being developed for future use by the fusion community

  13. Conceptual design of a Tokamak hybrid power reactor (THPR)

    International Nuclear Information System (INIS)

    Matsuoka, F.; Imamura, Y.; Inoue, M.; Asami, N.; Kasai, M.; Yanagisawa, I.; Ida, T.; Takuma, T.; Yamaji, K.; Akita, S.

    1987-01-01

    A conceptual design of a fusion-fission hybrid tokamak reactor has been carried out to investigate the engineering feasibility and promising scale of a commercial hybrid reactor power plant. A tokamak fusion driver based on the recent plasma scaling law is introduced in this design study. The major parameters and features of the reactor are R=6.06 m, a=1.66 m, Ip=11.8 MA, Pf=668 MW, double null divertor plasma and steady state burning with RF current drive. The fusion power has been determined with medium energy multiplication in the blanket so as to relieve thermal design problems and produce electric power around 1000 MW. Uranium silicide is used for the fast fission blanket material to promise good nuclear performance. The coolant of the blanket is FLIBE and the tritium breeding blanket material is Li 2 O ceramics providing breeding ratio above unity

  14. Life cycle assessment for energy payback of spherical tokamak reactors

    International Nuclear Information System (INIS)

    Ban, Kanae; Yamazaki, Kozo; Arimoto, Hideki; Oishi, Tetsutarou; Shoji, Tatsuo

    2012-01-01

    The energy payback ratio (EPR) is the index evaluating how a power plant produces the higher output energy effectively from the lower input energy. We used the system code and evaluated the EPRs of three reactor models; spherical tokamak (NC-ST), low aspect ratio tokamak reactor (LATR), and standard tokamak reactor (TR). The NC-ST and LATR systems have no center solenoid at the inboard side of plasma. In the case of NC-ST, normal conducting coils (NC) are used for toroidal field system. In LATR and TR superconducting coils (SC) are adopted. After we studied the relationship between the EPR of each fusion reactor and the aspect ratio in the range from 1.2 to 4.0, we evaluated the EPR of three typical models; ARIES-ST like, VECTOR like, and ARIES-RS like. The input parameters of aspect ratio, elongation, and normalized beta are as same as those of each original reactor. It was found that the EPR of LATR is the highest. Since LATR has no power loss due to ohmic-heating of normal conducting coils, the input energy of construction, operation and replacement is the lowest. (author)

  15. Factors affecting the minimum capital cost of a tokamak reactor

    International Nuclear Information System (INIS)

    Hancox, R.

    1981-01-01

    The Mk IIA Culham conceptual tokamak reactor design is a 2500 MWe steady-state reactor developed on the basis of a cost optimisation. A revised 1200 MWe conceptual design, the Mk IIB, used a lower wall loading and lower thermodynamic efficiency. A detailed costing of the Mk IIB design, however, showed it to have an unacceptably high capital cost. Since this high cost is a common characteristic of many fusion reactor designs, the cost optimisation of the Mk II design has been reconsidered. (author)

  16. Tritium experience in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A. [Princeton Univ., NJ (United States). Princeton Plasma Physics Lab.; Brooks, J.N. [Argonne National Lab., IL (United States); Hogan, J. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  17. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  18. Minerals resource implications of a tokamak fusion reactor economy

    International Nuclear Information System (INIS)

    Cameron, E.; Conn, R.W.; Kulcinski, G.L.; Sviatoslavsky, I.

    1979-09-01

    The mineral resource implications of an economy of tokamak-type fusion reactors are assessed based upon the recent conceptual reactor design study, NUWMAK, developed at the University of Wisconsin. For comparative purposes, various structural alloys of vanadium and steel are assumed to be usable in the NUWMAK design in place of the titanium alloy originally selected. In addition, the inner blanket core and magnet system of the conceptual reactor, HFCTR, developed at the Massachusetts Institute of Technology, are assumed to be interchangeable with the comparable components in NUWMAK. These variations permit a range of likely requirements to be assessed

  19. Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically in influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the part of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material inter actions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (author)

  20. Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next-step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several cm from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the engineering design activities of the international thermonuclear experimental reactor project (ITER) and significant progress has been made in better understanding these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/re-deposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (orig.)

  1. Afterheat assessment for conceptual tokamak reactors

    International Nuclear Information System (INIS)

    Goddard, A.J.H.; Jayatissa, S.

    1987-01-01

    Afterheat represents an important consideration in design of conceptual fusion power reactors, particularly during normal or unplanned shutdown. Afterheat calculations have been undertaken for various generic designs, but with special reference to the Culham DEMO reactor. These calculations have included the redistribution of heating by gamma ray transport. Selected temperature response calculations have been undertaken. (author)

  2. Tokamak fusion reactors with less than full tritium breeding

    International Nuclear Information System (INIS)

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed

  3. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

  4. MHD equilibrium properties of tokamak fusion reactor designs

    International Nuclear Information System (INIS)

    Todd, A.M.M.; Gralnick, S.L.; Dalhed, H.E.

    1976-01-01

    The equilibrium properties of several Tokamak Reactor Designs are analyzed and compared for varying pressure and current profiles using the Princeton Equilibrium Code. It is found that the UWMAK configuration has a broader range of equilibria than the Princeton Reference Design configuration, but that the safety factor on axis is less than unity for peaked current distributions. The Argonne Experimental Power Reactor has a satisfactory range of equilibria, but a means of limiting or diverting the plasma has not yet been proposed, and this may substantially change the results obtained

  5. DEALS: a maintainable superconducting magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Hseih, S.Y.; Danby, G.; Powell, J.R.

    1979-01-01

    The feasibility of demountable superconducting magnet systems has been examined in a design study of a DEALS [Demountable Externally Anchored Low Stress] TF magnet for an HFITR [High Field Ignition Test Reactor] Tokamak device. All parts of the system appear feasible, including the demountable superconducting joints. Measurements on small scale prototype joints indicate that movable pressure contact joints exhibit acceptable electrical, mechanical, and cryogenic performance. Such joints permit a relatively simple support structure and are readily demountable. Assembly and disassembly sequences are described whereby any failed portion of the magnet, or any part of the reactor inside the TF coils can be removed and replaced if necessary

  6. TIBER II: an upgraded tokamak igntion/burn experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Perkins, L.J.

    1986-01-01

    We are disIgning a minimum-size Tokamak ignition/Burn Reactor (TIBER II). This design incorporates physics requirements, neutron wall loading and fluence parameters that will make it compatible with a nuclear testing mission. Reactor relevant physics will be tested by using current drive and steady-state operation. Although the design accommodates several current drive options, including neutral beams, the base case uses a combination of lower hybrid and electron-cyclotron radio frequency power. Minimum neutron shielding, compact structures, high magnet-current densities, and remotely maintainable vacuum seals, all contribute to the compact size

  7. Resistive toroidal-field coils for tokamak reactors

    International Nuclear Information System (INIS)

    Kalnavarns, J.; Jassby, D.L.

    1980-11-01

    This paper analyzes the optimization of the geometry of resistive TF coils of rectangular bore for tokamak fusion test reactors and practical neutron generators. In examining the trade-offs between geometric parameters and magnetic field for reactors giving a specified neutron wall loading, either the resistive power loss or the lifetime coil cost can be minimized. Aspects of cooling, magnetic stress, and construction are addressed for several reference designs. Bending moment distributions in closed form have been derived for rectangular coils on the basis of the theory of rigid frames. Candidate methods of fabrication and of implementing demountable joints are summarized

  8. Review of tokamak power reactor and blanket designs in the United States

    International Nuclear Information System (INIS)

    Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

    1986-01-01

    The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

  9. Wildcat: A commercial deuterium-deuterium tokamak reactor

    International Nuclear Information System (INIS)

    Evans, K.; Baker, C.C.; Barry, K.M.

    1983-01-01

    WILDCAT is a conceptual design of a catalyzed deuterium-deuterium tokamak commercial fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing deuterium-tritium (D-T) designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete conceptual design

  10. Tandem mirror and tokamak reactor maintainability comparison

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1981-01-01

    The analysis proceeds through estimates of downtime and resources required for selected maintenance actions and optimization of the replacement fraction, availability and cost of electricity. Scheduled downtime estimates and availability goals provide a basis for determining allowable forced outage downtimes. These analyses have been conducted with the assumption of redundancy wherever feasible but without the impact of maintenance equipment outages. Annual maintenance cost estimates and availabilities for both reactors are found to be approximately equal. However, the tandem mirror reactor capital costs are higher. Reduction of these costs appears feasible with the trend of current design studies toward smaller and more accessible machines

  11. Criteria of the efficiency for radiation protection of tokamak reactor superconducting magnet coils

    International Nuclear Information System (INIS)

    Zimin, S.A.

    1988-01-01

    Factors determining serviceability of the main elements (superconductor, stabilizing conductor, insulation) of superconducting magnet coils for tokamak reactors are discussed. It is suggested that the limiting values of total and specific energy release in the material of superconducting coils, increase in electric resistance of the stabilizing conductor, decrease in the superconductor critical current and damage of the superconducting magnet insulation should be used as criteria of the reactor internal radiation protection efficiency. The conclusion is made that neutron fluence in the magnet coil components considered can be used as a generalized criterion of the first approximation for the evaluation of the protection efficiency

  12. Preconceptual design and assessment of a Tokamak hybrid reactor

    Science.gov (United States)

    Teofilo, V. L.; Leonard, B. R., Jr.; Aase, D. T.; Bickford, W. E.; McCormick, N. J.; McGrath, R. T.; Morrison, J. E.; Perry, R. T.; Schulte, S. C.; Willingham, C. E.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant was performed. The Tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb3Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited Tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs were made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis was made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

  13. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb 3 Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered

  14. Insulator applications in a Tokamak reactor

    International Nuclear Information System (INIS)

    Leger, D.

    1986-06-01

    Insulators, among which insulators ceramics, have great potential applications in fusion reactors. They will be used for all plasma-facing components as protection and, magnetic fusion devices being subject to large electrical currents flowing in any parts of the device, for their electrical insulating properties

  15. 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

  16. 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

  17. A comparison of burn cycle options for Tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1985-01-01

    Consideration of burn cycle options for commercial tokamaks shows that there is substantial motivation to achieve steady state operation. This is partly due to longer replacement periods for first wall and impurity control components, but, in addition, large cost savings are found when magnets, power supplies, and the energy transfer system are not frequently pulsed. The hybrid burn cycle, with a combination of ohmic and noninductive current drive, does not significantly improve the economics of ohmically-driven commercial reactors with large major radius. However, an INTOR-class device has a critically small hole in the doughnut, and we find for this size tokamak that the hybrid cycle is preferred over ohmically-driven operation

  18. Plasma features and alpha particle transport in low-aspect ratio tokamak reactor

    International Nuclear Information System (INIS)

    Xu Qiang; Wang Shaojie

    1997-06-01

    The results of the experiment and theory from low-aspect ratio tokamak devices have proved that the MHD stability will be improved. Based on present plasma physics and extrapolation to reduced aspect ratio, the feature of physics of low-aspect ratio tokamak reactor is discussed primarily. Alpha particle confinement and loss in the self-justified low-aspect ratio tokamak reactor parameters and the effect of alpha particle confinement and loss for different aspect ratio are calculated. The results provide a reference for the feasible research of compact tokamak reactor. (9 refs., 2 figs., 3 tabs.)

  19. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

    1992-01-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  20. 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)

  1. HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

    International Nuclear Information System (INIS)

    Schultz, J.H.; Montgomery, D.B.

    1983-04-01

    HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

  2. The ARIES-III D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1992-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

  3. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  4. Neutronics and photonics calculations for the tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Baker, V.C.; Barnes, J.M.

    1977-03-01

    The results of one-dimensional neutronic and photonic calculations that compare the nuclear performance of blanket and shield designs proposed for use in the Tokamak Experimental Power Reactor are presented. The nuclear analysis was carried out for both nonbreeding and tritium-breeding blanket modules to compare the spatial variations of the radiation flux and energy distributions, nuclear heating, radiation damage, and tritium breeding. Nonbreeding blanket modules that contain potassium plus SS-316 or potassium only as the energy-absorbing medium and breeding blankets that use natural lithium as the fertile material were evaluated as a function of the first-wall cooling scheme

  5. A concept of drastically easy maintenance (DREAM) tokamak reactor

    International Nuclear Information System (INIS)

    Nishio, S.; Murakami, Y.; Adachi, J.; Miura, H.; Aoki, I.; Seki, Y.

    1994-01-01

    Here, a new concept of easy maintenance tokamak reactor is proposed. For the easy maintenance scheme, a plasma configuration with high aspect ratio (with slender doughnut shape) around 6 and a small number of torus sectoring of 12 are introduced. The adjacent torus sectors are mechanically jointed by a bolt/flange structure. A plasma vacuum region is comprised of a cryostat inner surface and 12 horizontal access doors. Overall machine size is determined from a fusion power of 3 GW and almost 100% bootstrap current within moderate Troyon coefficient of 3.5, MHD safety factor of 5 and plasma elongation of 1.6. (author)

  6. OTR-experimental fusion-fission tokamak-reactor concept

    International Nuclear Information System (INIS)

    Kadomtsev, B.B.; Kolbasov, B.N.; Orlov, V.V.

    1987-01-01

    An original concept of OTR-experimental hybrid fusion-fission tokamak-reactor is given. Tendencies to probable changes in the reactor parameters and some engineering alternatives are discussed. The feasibility of ''non-chemical'' fuel cycle with direct neutron enrichment (in the fusion-fission reactor blanket) of the fuel elements made of depleted uranium for their use in fission reactors is considered. Advantages and shortcomings of the blanket cooling with water, liquid metal and with a gas are briefly discussed. The parameters of experimental and commercial fusion-fission reactors and a composition of the nuclear fuel unloaded from such reactors are given. An option for the detail design which should be done by us needs some corrections in plasma geometry and in other parameters with account taken of new experimental and theoretical results, as well as that of the estimates and comparison of different engineering solutions referred to the plasma engineering, electrophysical and nuclear engineering aspects of the design. An initial OTR-concept, the tendencies to the correction of parameters and some engineering alternatives studied in the present stage of design are discussed in the paper. 12 refs, 2 tabs

  7. COAST-4, Design and Cost of Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sink, D.A.; Iwinski, E.M.

    1985-01-01

    1 - Description of program or function: COAST produces a generalized description of a D-T burning tokamak reactor and facility. In each complete calculation, the geometry, dimensions, and ratings of approximately 50 subsystems are determined. In addition, performance data associated with the tokamak and facility operation and a detailed cost estimate, subsystem-by-subsystem, are provided. The cost estimates include the evaluation of direct and indirect capital costs, time-related, and operating costs. Devices which can be sized and costed include: TFTR (Tokamak Fusion Test Reactor), The Next Step (TNS) type of devices, fusion-fission hybrids, engineering demonstration reactors, power producing reactors, and commercial (power and fissile fuel breeding) reactors. Both the ignition mode and the neutral-beam-driven mode of plasma operation are modeled. The plasma engineering calculations involve 0-D models which account for energy balance, particle balance, alpha-particle effects, slowing-down theory, plasma-plasma and beam-plasma fusion reaction rates, impurity effects, profile effects, and n(e)-tau(E) scaling. The tokamak engineering calculations account for poloidal and toroidal magnetic field coil assemblies (both superconducting and copper conductors are possible), neutral beam injectors, blanket/shield assemblies, plasma fueling, divertors, heat dissipation systems, and related power supplies. The modeling for the shaping field coils includes the positioning of each coil relative to the other device components, as well as scaling the coil currents from a reference set of current magnitudes and positions (which are supplied as input and should be consistent with a coil set providing equilibrium and stability conditions for the plasma). The reactor cell containing the tokamak device, the turbine-generator plant and facility, and balance-of-plant systems are taken into account. 2 - Method of solution: The COAST code 'builds' the tokamak around the plasma and

  8. International thermal reactor development

    International Nuclear Information System (INIS)

    Zebroski, E.L.

    1977-01-01

    The worldwide development of nuclear power plants is reviewed. Charts are presented which show the commitment to light-water reactor capacity construction with breakdown by region and country. Additional charts show the major nuclear research centers which have substantial scope in light water reactor development and extensive international activities

  9. Neutral-beam-injected tokamak fusion reactors: a review

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1976-08-01

    The theories of energetic-ion velocity distributions, stability, injection, and orbits were summarized. The many-faceted role of the energetic ions in plasma heating, fueling, and current maintenance, as well as in the direct enhancement of fusion power multiplication and power density, is discussed in detail for three reactor types. The relevant implications of recent experimental results on several beam-injected tokamaks are examined. The behavior of energetic ions is found to be in accordance with classical theory, large total ion energy densities are readily achieved, and plasma equilibrium and stability are maintained. The status of neutral-beam injectors and of conceptual design studies of beam-driven reactors are briefly reviewed. The principal plasma-engineering problems are those associated directly with achieving quasi-stationary operation

  10. Systems study of tokamak fusion--fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations.

  11. Electromagnetic effects involving a tokamak reactor first wall and blanket

    International Nuclear Information System (INIS)

    Turner, L.R.; Evans, K. Jr.; Gelbard, E.; Prater, R.

    1980-01-01

    Four electromagnetic effects experienced by the first wall and blanket of a tokamak reactor are considered. First, the first wall provides reduction of the growth rate of vertical axisymmetric instability and stabilization of low mode number interval kink modes. Second, if a rapid plasma disruption occurs, a current will be induced on the first wall, tending to maintain the field formerly produced by the plasma. Third, correction of plasma movement can begin on a time scale much faster than the L/R time of the first wall and blanket. Fourth, field changes, especially those from plasma disruption or from rapid discharge of a toroidal field coil, can cause substantial eddy current forces on elements of the first wall and blanket. These effects are considered specifically for the first wall and blanket of the STARFIRE commercial reactor design study

  12. Power supply requirements for a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Brooks, J.N.; Kustom, R.L.

    1979-02-01

    The power supply requirements for a 7-M major radius commercial tokamak reactor have been examined, using a system approach combining models of the reactor and poloidal coil set, plasma burn cycle and MHD calculations, and power supply characteristics and cost data. A conventional system using an MGF set and solid-state rectifier/inverter power supplies was studied in addition to systems using a homopolar generator, superconducting energy storage inductor, and dump resistors. The requirements and cost of the power supplies depend on several factors but most critically on the ohmic heating ramp time used for startup. Long ramp times (approx. > 8 s) seems to be feasible, from the standpoint of resistive volt-second losses, and would appear to make conventional systems quite competitive with nonconventional ones, which require further research and development

  13. Manufacturing aspects of toroidal field magnets for tokamak power reactors

    International Nuclear Information System (INIS)

    Smith, G.E.

    1977-01-01

    The manufacturing aspects of two different toroidal field (TF) coils are discussed briefly. The first of these is the Tokamak Fusion Test Reactor (TFTR) TF coil, which is presently in the design phase at the Princeton Plasma Physics Laboratory. The second coil configuration reflects a reference design for a large experimental power reactor (EPR) which originated at the Oak Ridge National Laboratory. This configuration was used by the Grumman Aerospace Corporation of Bethpage, N.Y. as the basis for an investigation of the manufacturing aspects of a large superconducting TF coil, the results of which are summarized. For each of these coils, the major characteristics of the design are first briefly described and the significant manufacturing aspects are then summarized

  14. Systems study of tokamak fusion--fission reactors

    International Nuclear Information System (INIS)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

  15. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  16. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next-step DT (deuterium-tritium) fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D (Research and Development) avenues for their resolution are presented

  17. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  18. Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

    International Nuclear Information System (INIS)

    Hancox, R.; Mitchell, J.T.D.

    1977-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

  19. Compact Commercial Tokamak Reactor (CCTR): a concept for a 500-MWe commercial-tokamak fusion system

    International Nuclear Information System (INIS)

    Gillen, T.J.

    1980-11-01

    A detailed set of self-consistent parameters and costs for the conceptual design of a Compact Commercial Tokamak Reactor (CCTR) is given. Several of the basic design features are the following: an ignited plasma with a major radius of 4.9 m and minor radius of 1.4 m; a net electrical output of 500 MW; a borated-water-cooled, stainless steel shield; and a toroidal field of 12 T at the coil. The design, which utilizes the Westinghouse computer code for the COsting And Sizing of D-T burning Tokamaks (COAST), mainly provides the sizes and geometries associated with the definition of the main component features for which a detailed engineering design can be effectively undertaken. Design study alternatives, including a neutral beam driven design option, a design option with a toroidal field of 13 T at the coil, and a tungsten-shielded option are considered for the CCTR. Also included is the conceptual design of a Compact Fusion Engineering Device

  20. UWMAK-II: a conceptual tokamak reactor design

    International Nuclear Information System (INIS)

    1975-10-01

    This report describes the conceptual design of a Tokamak fusion power reactor, UWMAK-II. The aim of this study is to perform a self consistent and thorough analysis of a probable future fusion power reactor in order to assess the technological problems posed by such a system and to examine feasible solutions. UWMAK-II is a conceptual Tokamak fusion reactor designed to deliver 1716 MWe continuously and to generate 5000 MW(th) during the plasma burn. The structural material is 316 stainless steel and the primary coolant is helium. UWMAK-II is a low aspect ratio, low field design and includes a double null, axisymmetric poloidal field divertor for impurity control. In addition, a carbon curtain, made of two dimensional woven carbon fiber, is mounted on the first vacuum chamber wall to protect the plasma from high Z impurities and to protect the first wall from erosion by charged particle bombardment. The blanket is designed to minimize the inventory of both tritium and lithium while achieving a breeding ratio greater than one. This has led to a blanket design based on the use of a solid breeding material (LiAlO 2 ) with beryllium as a neutron multiplier. The lithium is enriched to 90 percent 6 Li and the blanket coolant is helium at a maximum pressure of 750 psia (5.2 x 10 6 N/m 2 ). A cell of the UWMAK-II blanket design is shown. The breeding ratio is between 1.11 and 1.19 based on one-dimensional discrete ordinates transport calculations, depending on the method of homogenization. Detailed Monte Carlo calculations, which take into account the more complicated geometry, give a breeding ratio of 1.06. The total energy per fusion is 21.56 MeV, which is fairly high

  1. Tokamak power reactor ignition and time dependent fractional power operation

    International Nuclear Information System (INIS)

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-06-01

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

  2. 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

  3. Plan for 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) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D ampersand D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D ampersand D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D ampersand D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D ampersand D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates

  4. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Blevins, J.D.; Stasko, R.R.

    1989-09-01

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  5. Development of large insulator rings for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1978-01-01

    This paper discusses research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applications, fabrication approach and testing activities are highlighted

  6. Safety analyses of the ARIES tokamak reactor designs

    International Nuclear Information System (INIS)

    Herring, J.S.; McCarthy, K.A.; Dolan, T.J.

    1994-01-01

    The ARIES design has sought to maximize environmental and safety advantages of fusion through careful selection of materials and design. The ARIES-I tokamak reactor design consists of an SiC composite structure for the first wall and blanket, cooled by 10MPa helium. The breeder is Li 2 ZrO 3 . The divertor consists of SiC composite tubes coated with 2mm tungsten. Loss-of-cooling accident (LOCA) calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. The ARIES-II design includes liquid lithium and vanadium, both of which have low activation, multiple barriers between the lithium and air and an inert cover gas to prevent lithium-air reactions. The ARIES-II reactor is passively safe with a total 1km early dose of about 88rem (0.88Sv). ARIES-III was an extensive examination of the viability of a D- 3 He fueled tokamak power reactor. Because neutrons are produced only through side reactions (D+D→ 3 He+n, and D+D→T+p followed by D+T→ 4 He+n), the reactor has a reduced activation of the first wall and shield, low afterheat and class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. We modeled a LOCA in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, below 600 C, release fractions are small. We analyzed the disposition of the 20g per day of tritium that is produced by D-D reactions and removed by vacuum pumps. The ARIES-IV coolant is helium and the breeder is lithium oxide. The structure is silicon carbide. Since the neutron multiplier, beryllium metal, is combustible, releasing about 60MJkg -1 , beryllium is the chief source of chemical energy. Less than 10% of the 24 Na inventory is likely to diffuse out of the SiC during a fire in which the beryllium is consumed. Therefore, the offsite dose would be less than 200rem. ((orig.))

  7. Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

    International Nuclear Information System (INIS)

    Baker, C.C.

    1980-01-01

    STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described

  8. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    International Nuclear Information System (INIS)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

  9. A conceptual design of superconducting spherical tokamak reactor

    International Nuclear Information System (INIS)

    Nagayama, Yoshio; Shinya, Kichiro; Tanaka, Yasutoshi

    2012-01-01

    This paper presents a fusion reactor concept named 'JUST (Japanese Universities' Super Tokamak reactor)'. From the plasma confinement system to the power generation system is evaluated in this work. JUST design has features as follows: the superconducting magnet, the steady state operation with high bootstrap current fraction, the easy replacement of neutron damaged first wall, the high heat flux in the divertor, and the low cost (or high β). By winding the OH solenoid over the center stack of toroidal field coil, we have the low aspect ratio and the 80cm thick neutron shield to protect the superconducting center stack. JUST is designed by using the 0-D transport code under the assumption that the energy confinement time is 1.8 times of the IPB98(y,2) scaling. Main parameters are as follows: the major radius of 4.5m, the aspect ratio of 1.8, the elongation ratio of 2.5, the toroidal field of 2.36T, the plasma current of 18MA, the toroidal beta of 22%, the central electron and ion temperature of 15keV and the fusion thermal power of 2.4GW. By using the mercury heat exchanger and the steam turbine, the heat efficiency is 33% and the electric power is 0.74GW. (author)

  10. Oak Ridge Tokamak experimental power reactor study scoping report

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, M.

    1977-03-01

    This report presents the scoping studies performed as the initial part of the program to produce a conceptual design for a Tokamak Experimental Power Reactor (EPR). The EPR as considered in this study is to employ all systems necessary for significant electric power production at continuous high duty cycle operation; it is presently scheduled to be the final technological step before a Demonstration Reactor Plant (Demo). The scoping study tasks begin with an exploration and identification of principal problem areas and then concentrate on consideration and evaluation of alternate design choices for each of the following major systems: Plasma Engineering and Physics, Nuclear, Electromagnetics, Neutral Beam Injection, and Tritium Handling. In addition, consideration has been given to the integration of these systems and requirements arising out of their incorporation into an EPR. One intent of this study is to document the paths explored in search of the appropriate EPR characteristics. To satisfy this intent, the explorations are presented in chart form outlining possible options in key areas with extensive supporting footnotes. An important result of the scoping study has been the development and definition of an EPR reference design to serve as (1) a common focus for the continuing design study and (2) a guide for associated development programs. In addition, the study has identified research and development requirements essential to facilitate the successful conceptual design, construction, and operation of an EPR.

  11. Deuterium-tritium experiments on the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D. [Princeton Plasma Physics Lab., NJ (United States); Anderson, J.L.; Barnes, C.W. [Los Alamos National Lab., NM (United States)] [and others

    1994-11-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.

  12. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D. [and others

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.

  13. Energy conversion options for ARIES-III - A conceptual D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Santarius, J.F.; Blanchard, J.P.; Emmert, G.A.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Ghoneim, N.M.; Hasan, M.Z.; Mau, T.K.; Greenspan, E.; Herring, J.S.; Kernbichler, W.; Klein, A.C.; Miley, G.H.; Miller, R.L.; Peng, Y.K.M.

    1989-01-01

    The potential for highly efficient conversion of fusion power to electricity provides one motivation for investigating D- 3 He fusion reactors. This stems from: (1) the large fraction of D- 3 He power produced in the forms of charged particles and synchrotron radiation which are amenable to direct conversion, and (2) the low neutron fluence and lack of tritium breeding constraints, which increase design flexibility. The design team for a conceptual D- 3 He tokamak reactor, ARIES-III, has investigated numerous energy conversion options at a scoping level in attempting to realize high efficiency. The energy conversion systems have been studied in the context of their use on one or more of three versions of a D- 3 He tokamak: a first stability regime device, a second stability regime device, and a spherical torus. The set of energy conversion options investigated includes bootstrap current conversion, compression-expansion cycles, direct electrodynamic conversion, electrostatic direct conversion, internal electric generator, liquid metal heat engine blanket, liquid metal MHD, plasma MHD, radiation boiler, scrape-off layer thermoelectric, synchrotron radiation conversion by rectennas, synchrotron radiation conversion by thermal cycles, thermionic/AMTEC/thermal systems, and traveling wave conversion. The original set of options is briefly discussed, and those selected for further study are described in more detail. The four selected are liquid metal MHD, plasma MHD, rectenna conversion, and direct electrodynamic conversion. Thermionic energy conversion is being considered, and some options may require a thermal cycle in parallel or series. 17 refs., 3 figs., 1 tab

  14. Internal magnetic field measurement in tokamak plasmas using a ...

    Indian Academy of Sciences (India)

    Abstract. In a tokamak plasma, the poloidal magnetic field profile closely depends on the current density profile. We can deduce the internal magnetic field from the analysis of circular polarization of the spectral lines emitted by the plasma. The theory of the measurement and a detailed design of the Zeeman polarimeter ...

  15. Internal Magnetic Configuration Measured by ECE Imaging on EAST Tokamak

    International Nuclear Information System (INIS)

    Xu Ming; Wen Yizhi; Xie Jinlin; Yu Changxuan; Gao Bingxi; Xu Xiaoyuan; Liu Wandong; Hu Liqun; Sun Youwen; Qian Jinping; Wan Baonian

    2013-01-01

    ECE imaging (electron cyclotron emission imaging) is an important diagnostic which can give 2D imaging of temperature fluctuation in the core of tokamak. A method based on ECE imaging is introduced which can give the information of the position of magnetic axis and the structure of internal magnetic surface for EAST tokamak. The EFIT equilibrium reconstruction is not reliable due to the absence of important core diagnostic at the initial phase for EAST, so the information given by ECE imaging could help to improve the accuracy of EFIT equilibrium reconstruction. (magnetically confined plasma)

  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. 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

  18. Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Hill, K.W.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Marmar, E.S.; Snipes, J.A.; Terry, J.L.; Batha, S.; Bell, R.E.; Bitter, M.; Bush, C.E.; Chang, Z.; Darrow, D.S.; Ernst, D.; Fredrickson, E.; Grek, B.; Herrmann, H.W.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Levinton, F.M.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.; Ramsey, A.T.; Roquemore, A.L.; Skinner, C.H.; Stevenson, T.; Stratton, B.C.; Synakowski, E.; Taylor, G.; von Halle, A.; von Goeler, S.; Wong, K.L.; Zweben, S.J.

    1996-01-01

    Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium endash tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10 21 m -3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun. copyright 1996 American Institute of Physics

  19. Ion cyclotron transmission spectroscopy in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Greene, G.J.

    1993-09-01

    The propagation of waves in the ion cyclotron range of frequencies has been investigated experimentally in the Tokamak Fusion Test Reactor. A small, broadband, radiofrequency (rf) magnetic probe located outside the plasma limiter, at a major radius near that of the plasma center, was excited with a low power, frequency swept source (1--200 MHz). Waves propagating to a distant location were detected with a second, identical probe. The rf transmission spectrum revealed a region of attenuation over a band of frequencies for which the minority fundamental resonance was located between the outer plasma edge and the major radius of the probe location. Distinct, non-overlapping attenuation bands were observed from hydrogen and helium-3 minority species; a distinct tritium band should be observed in future DT experiments. Rapid spectrum acquisition during a helium-3 gas puff experiment showed that the wave attenuation involved the plasma core and was not a surface effect. A model in which the received power varied exponentially with the minority density, averaged over the resonance region, fit the time evolution of the probe signal relatively well. Estimation of a 1-d tunneling parameter from the experimental observations is discussed. Minority concentrations of less than 0.5 % can be resolved with this measurement.

  20. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles.

  1. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles

  2. 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

  3. Abstracts of the International seminar 'Experimental possibilities of KTM tokamak and research programme'

    International Nuclear Information System (INIS)

    2005-01-01

    The International seminar 'Experimental possibilities of KTM tokamak and research programme' was held in 10-12 October 2005 in Astana city (Kazakhstan). The seminar was dedicated to problems of KTM tokamak commissioning. The Collection of abstracts comprises 45 papers

  4. Problems in modeling TF ripple loss of fast alphas from a tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hively, L.M.

    1987-01-01

    The present status of modeling TF ripple loss of fast alphas from tokamaks is summarized. The modeling issues are discussed, and several new aspects of this problem are described, including gyromotion, radial electric field, and sawtoothing. Existing models predict that TF ripple loss of fast alphas will have a low-to-moderate impact on the design of a tokamak engineering test reactor (ETR). 52 refs., 3 figs., 2 tabs.

  5. In-vessel tritium measurements using beta decay in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zweben, S.J.; Gentile, C.; Mueller, D.; Nagy, A.; Pearson, G.; Skinner, C.H.; Walsh, J. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States)

    1999-01-01

    Tritium on the inside walls of the Tokamak Fusion Test Reactor was detected by configuring the vacuum vessel as an ionization chamber and measuring the secondary electron current from the tritium beta decay. The vessel was typically filled with {approx}5 Torr of dry nitrogen and the secondary electron current was collected by an internal electrode biased to about +15 V with respect to the vessel wall. The measured variations of the collected current with gas pressure, bias voltage, and applied magnetic fields are described, as well as an {ital in situ} calibration made by injecting a known amount of tritium gas. Improved versions of this diagnostic may be useful to track the in-vessel content of surface tritium in future fusion devices. {copyright} {ital 1999 American Institute of Physics.}

  6. Damping of reactor internals

    International Nuclear Information System (INIS)

    Singleton, N.R.; Bohm, G.J.

    1977-01-01

    This paper presents and discusses the results of a study of internals damping using data obtained from wave analysis of PWR plant flow tests, and shaker tests. The damping values were obtained from vibration data taken during the pre-operational testing of several reactor plants and some in air shaker tests. Parameters available in the data include the presence of the core, the presence and position of the control rod drive line, reactor coolant temperature, and combination of reactor coolant pumps in operation. Modal damping values for the structures reported were obtained from the decay of autocorrellograms or from the modal response half-power bandwidths of frequency spectra. For the lower frequency core barrel-reactor vessel beam modes 2% to 5% damping values were found for minimum damping values. Significantly higher values are found in the data when, for example, intermittent contact occurs at the core barrel level supports. Core barrel and thermal shield shell modes having natural frequencies in the frequency range of interest for seismic response calculations exhibit damping values generally on the order of 1% to 2%. Higher frequency, very low amplitude, shell modes of these structures can have damping values of less than 1%. Damping values for guide tubes were found to have minimum values of 2% to 5% depending on their core location. The cross flow velocity and thus the floor turbulence excited amplitude is higher for guide tubes in core locations near the outlet nozzles. Information on the damping of upper support columns which are similarly excited is also given. Damping values reported are suitable for normal operation design conditions, i.e., for oscillatory behavior and relatively small amplitudes. The extrapolation of the data to obtain realistic values for large seismic events and for loss of coolant accidents is also discussed

  7. High-energy tritium beams as current drivers in tokamak reactors

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; Grisham, L.R.

    1983-04-01

    The effect on neutral-beam design and reactor performance of using high-energy (approx. 3-10 MeV) tritium neutral beams to drive steady-state tokamak reactors is considered. The lower current of such beams leads to several advantages over lower-energy neutral beams. The major disadvantage is the reduction of the reactor output caused by the lower current-drive efficiency of the high-energy beams

  8. Oak Ridge Tokamak experimental power reactor study reference design

    International Nuclear Information System (INIS)

    Roberts, M.; Bettis, E.S.

    1975-11-01

    A Tokamak EPR Reference Design is presented as a basis for further design study leading to a Conceptual Design. The set of basic plasma parameters selected--minor radius of 2.25 m, major radius of 6.75 m, magnetic field on axis of 4.8 T and plasma current of 7.2 MA--should produce a reactor-grade plasma with a significant neutron flux, even with the great uncertainty in plasma physics scaling from present experience to large sizes. Neutronics and heat transfer calculations coupled with mechanical design and materials considerations were used to develop a blanket and shield capable of operating at high temperature, protecting the surrounding coils, being maintained remotely and, in a few experimental modules, breeding tritium. Nb 3 Sn and NbTi superconductors are used in the toroidal field coil design. The coil system was developed for a maximum field of 11 T at the winding (to give a field on axis of 4.8 T), and combines multifilamentary superconducting cable with forced flow of supercritical helium enclosed in a steel conduit. The structural system uses a stainless steel center bucking ring and intercoil box beam bracing to provide rigid support for coils against the centering force, overturning moments from poloidal fields and faults, other external forces, and thermal stresses. The poloidal magnetics system is specially designed both to reduce the total volt-second energy requirements and to reduce the magnitude of the rate of field change at the toroidal field coils. The rate of field change imposed upon the toroidal field coils is reduced by at least a factor of 3.3 compared to that due to the plasma alone. Tritium processing, tritium containment and vacuum systems employ double containment and atmospheric cleanup to minimize releases. The document also contains discussions of systems integration and assembly, key research and development needs, and schedule considerations

  9. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Steady-state tokamak reactors: SSTR and A-SSTR

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru

    1998-01-01

    The SSTR is a steady-state tokamak reactor based on a bootstrap current with near term technologies. The A-SSTR is an aggressive fusion reactor concept designed to address issues of economical competitiveness. (author)

  10. Fast wave current drive in reactor scale tokamak plasmas

    International Nuclear Information System (INIS)

    Becoulet, A.; Moreau, D.; Saoutic, B.

    1991-01-01

    The possibility for driving current in large tokamak plasmas using the fast magnetosonic wave is analysed in terms of linear propagation-absorption, and also in terms of quasilinear absorption through an hamiltonian analysis of the wave-particle interaction. The tokamak geometry is shown to strongly influence the capability for the fast wave to sustain a significant part of the toroidal current. Synergetic effects with other scenarios are also discussed

  11. Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

    1980-09-01

    A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

  12. Lessons learned from the Tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)

    International Nuclear Information System (INIS)

    Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Werley, K.A.

    1994-01-01

    Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety and health (ES ampersand H) characteristics of projected tokamak power plants. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances relative to present understanding in physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advanced tokamak plasmas configured in the second-stability regime that achieve both high β and bootstrap fractions near unity through strong profile control offer high promise in this regard

  13. Lessons learned from the tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)

    International Nuclear Information System (INIS)

    Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Werley, K.A.

    1994-01-01

    Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety, and health (ES ampersand H) characteristics of projected tokamak power plants. Summarized herein are the composite conclusions and lessons developed in the course of four conceptual tokamak power-plant designs. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances in both physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advances in materials are also needed for the exploitation of environmental advantages otherwise inherent in fusion power

  14. Overview of the TIBER II [Tokamak Ignition/Burn Experimental Reactor] design

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.

    1987-01-01

    The TIBER II Tokamak Ignition/Burn Experimental Reactor design is the result of efforts by numerous people and institutions, including many fusion laboratories, universities, and industries. While subsystems will be covered extensively in other reports, this overview will attempt to place the work in perspective. Major features of the design are compact size, low cost, and steady-state operation. These are achieved through plasma shaping and innovative features such as radiation tolerant magnets and optimized shielding. While TIBER II can operate in a pulsed mode, steady-state is preferred for nuclear testing. Current drive is achieved by a combination of lower hybrid and neutral beams. In addition, 10 MW of ECR is added for disruption control and current drive profiling. The TIBER II design has been the US option in preparation for the International Thermonuclear Experimental Reactor (ITER). Other equivalent national designs are the NET in Europe, the FER in Japan and the OTR in the USSR. These designs will help set the basis for the new international design effort. 9 refs

  15. A comparison of pulsed and steady-state tokamak reactor burn cycles. Pt. 2

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.; Kim, S.

    1985-01-01

    Pulsed operation of a tokamak reactor imposes cost penalties due to such problems as mechanical fatigue and the need to periodically transfer large amounts of energy to various reactor components. This study focuses on lifetime limitations and capital costs of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas include: fatigue in pulsed poloidal field coils; out-of-plane bending fatigue in toroidal field coils; electric power supply costs; and noninductive current driver costs. A capital cost comparison is made for tokamak reactors operating under the four distinct operating cycles which have been proposed. Since high availability and a low cost of energy will be mandatory for a commercial fusion reactor, we can characterize improvements in physics and technology which will help achieve these goals for different burn cycles. A key conclusion is that steady-state operation is likely to result in the least expensive tokamak reactor (perhaps 20% cheaper than the best pulsed reactor), provided noninductive current drive efficiency can be increased roughly four-fold over present-day experimental results. (orig.)

  16. Wave-driver options for low-aspect-ratio steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1981-02-01

    Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A = 2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value

  17. Lifetime evaluation for thermal fatigue: application at the first wall of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Merola, M.; Biggio, M.

    1989-01-01

    Thermal fatigue seems to be the most lifetime limiting phenomenon for the first wall of the next generation Tokamak fusion reactors. This work deals with the problem of the thermal fatigue in relation to the lifetime prediction of the fusion reactor first wall. The aim is to compare different lifetime methodologies among them and with experimental results. To fulfil this purpose, it has been necessary to develop a new numerical methodology, called reduced-3D, especially suitable for thermal fatigue problems

  18. Calculation of the gamma radiation levels in and around the NET-DN tokamak reactor

    International Nuclear Information System (INIS)

    Hoejerup, C.F.

    1989-07-01

    Calculations of the gamma doses at a TOKAMAK fusion reactor, the NET-DN (1988) are presented. Neutron fluxes in the structures of the reactor are calculated by Monte Carlo methods (MCNP-2) and activations from the neutron induced reactions are determined by the ACTIVA computer programme. By Monte Carlo methods and the application of an approximative reciprocity principle, the gamma fluxes and doses are finally calculated. (author)

  19. Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

    International Nuclear Information System (INIS)

    Guthrie, J.A.S.

    1980-04-01

    In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

  20. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Science.gov (United States)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  1. Tokamak

    International Nuclear Information System (INIS)

    Meglicki, Z.

    1995-01-01

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs

  2. Remote servicing considerations for near term tokamak power reactors (TNS). Final summary

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1977-01-01

    Next generation Tokamaks require special consideration for remote servicing. Three major problems are highlighted: (1) movement of heavy components, (2) remote connection/disconnection of joints, and (3) remote cutting, welding, and leak detection. The first problem is assumed to be handled with existing expertise and is not considered. The remaining problems are thought to be minimized by considering two engineering departures from conventional tokamak design; locating the field shaping coils outside of the toroidal coils and enclosing the total device within an evacuated reactor cell. Five topics under this vacuum building concept are discussed: incremental cost, vacuum pumping, tritium containment, activation topology, and first year operations

  3. A cryogenic system for TIBER II [Tokamak Ignition/Burn Experimental Reactor

    International Nuclear Information System (INIS)

    Slack, D.S.; Kerns, J.A.

    1987-01-01

    Phase II of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) study describes one option for a small, economical, next-generation tokamak [1,2]. Because of its small size, minimum shielding is used between the plasma and the toroidal-field (TF) coils. Consequently, a large cryogenic system (approximately 70 kW at 4.5 K) capable of delivering forced-flow helium is required. This paper describes a cryogenic system that meets this requirement and includes TIBER-II requirements. 3 refs

  4. What is past is prologue: future directions in tokamak power reactor design research

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    Conceptual tokamak power reactor designs over the last five years have provided us with many fundamental insights regarding tokamaks as fusion reactors. This first generation of studies has helped lay the groundwork upon which to build improvements in reactor design and begin a process of optimization. After reviewing the first generation of studies and the primary conclusions they produced, we discuss four current designs that are representative of present trends in this area of research. In particular, we discuss the trends towards reduced reactor size and higher neutron wall loadings. Moving in this direction requires new approaches to many subsystem designs. We describe new approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets. We close with a discussion of the future role of conceptual reactor design research and the need for close interaction with ongoing experiments in fusion technology

  5. A conceptual design of a negative-ion-grounded advanced tokamak reactor

    International Nuclear Information System (INIS)

    Yamamoto, Shin; Ohara, Yoshihiro; Tani, Keiji

    1988-05-01

    The NAVIGATOR concept is based on the negative-ion-grounded 500 keV 20 MW neutral beam injection system (NBI system), which has been proposed and studied at JAERI. The NAVIGATOR concept contains two categories; one is the NAVIGATOR machine as a tokamak reactor, and the other is the NAVIGATOR philosophy as a guiding principle in fusion research. The NAVIGATOR machine implies an NBI heated and full inductive ramped-up reactor. The NAVIGATOR concept should be applied in a phased approach to and beyond the operating goal for the FER (Fusion Experimental Reactor, the next generation tokamak machine in Japan). The mission of the FER is to realize self-ignition and a long controlled burn of about 800 seconds and to develop and test fusion technologies, including the tritium fuel cycle, superconducting magnet, remote maintenance and breeding blanket test modules. The NAVIGATOR concept is composed of three major elements, that is, reliable operation scenarios, reliable maintenability and sufficient flexibility of the reactor. The NAVIGATOR concept well supports the ideas of phased operation and phased construction of the FER, which will result in the reduction of technological risk. The NAVIGATOR concept is expected to bring forth the fruits growing up in the present large tokamak machines in the form of next generation machines. In addition, the NAVIGATOR concept will supply many required databases for the DEMO reactor. The details of the NAVIGATOR concept is described in this paper, and the concept may indicate a feasible strategy for developing fusion research. (author)

  6. The ARIES-II and ARIES-IV second-stability tokamak reactors

    International Nuclear Information System (INIS)

    Najmabadi, F.; Conn, R.W.; Hasan, M.Z.; Mau, T.-K.; Sharafat, S.; Baxi, C.B.; Leuer, J.A.; McQuillan, B.W.; Puhn, F.A.; Schultz, K.R.; Wong, C.P.C.; Brooks, J.; Ehst, D.A.; Hassanein, A.; Hua, T.; Hull, A.; Mattis, R.; Picologlou, B.; Sze, D.-K.; Dolan, T.J.; Herring, J.S.; Bathke, C.G.; Krakowski, R.A.; Werley, K.A.; Bromberg, L.; Schultz, J.; Davis, F.; Holmes, J.A.; Lousteau, D.C.; Strickler, D.J.; Jardin, S.C.; Kessel, C.; Snead, L.; Steiner, D.; Valenti, M.; El-Guebaly, L.A.; Emmert, G.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.; Sviatoslavsky, I.N.; Cheng, E.T.

    1992-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. The ARIES-III study focuses on the potential of tokamaks to operate with D- 3 He fuel system as an alternative to deuterium and tritium. The ARIES-II and ARIES-IV designs have the same fusion plasma but different fusion-power-core designs. The ARIES-II reactor uses liquid lithium as the coolant and tritium breeder and vanadium alloy as the structural material in order to study the potential of low-activation metallic blankets. The ARIES-IV reactor uses helium as the coolant, a solid tritium-breeding material, and silicon carbide composite as the structural material in order to achieve the safety and environmental characteristic of fusion. In this paper the authors describe the trade-off leading to the optimum regime of operation for the ARIES-II and ARIES-IV second-stability reactors and review the engineering design of the fusion power cores

  7. Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

    International Nuclear Information System (INIS)

    Nishio, S.; Ueda, S.; Kurihara, R.; Kuroda, T.; Miura, H.; Sako, K.; Takase, H.; Seki, Y.; Adachi, J.; Yamazaki, S.; Hashimoto, T.; Mori, S.; Shinya, K.; Murakami, Y.; Senda, I.; Okano, K.; Asaoka, Y.; Yoshida, T.

    2000-01-01

    If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features of the prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900 deg. C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal

  8. Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Bell, M.G.; Beer, M.

    1997-02-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l i ). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q a ∼ 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l i plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q 0 > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions

  9. Generation IV reactors: international projects

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

    2003-01-01

    Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

  10. Elements of a method to scale ignition reactor Tokamak

    International Nuclear Information System (INIS)

    Cotsaftis, M.

    1984-08-01

    Due to unavoidable uncertainties from present scaling laws when projected to thermonuclear regime, a method is proposed to minimize these uncertainties in order to figure out the main parameters of ignited tokamak. The method mainly consists in searching, if any, a domain in adapted parameters space which allows Ignition, but is the least sensitive to possible change in scaling laws. In other words, Ignition domain is researched which is the intersection of all possible Ignition domains corresponding to all possible scaling laws produced by all possible transports

  11. Conceptual designs of tokamak reactor and R D

    International Nuclear Information System (INIS)

    Fukai, Yuzo; Yamato, Harumi; Sawada, Yoshio

    1983-01-01

    The conceptual design of both FER (Fusion Experimental Reactor) and R-project is now under way as the new step of JT-60. From the engineering viewpoint, these reactors, requiring D-T operation, have the challenge, such as the handling of tritium and components irradiated by neutron bombardment. Toshiba's design team is participating to these projects in order to realize the reactor and plant concept coping with the above objectives. This paper represents the conceptual design contributions of the FER and R-project as well as R D technology which are now under development, such as tritium handling app aratus, reactor materials, etc. (author)

  12. Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

    International Nuclear Information System (INIS)

    Sink, D.A.; Gibson, G.

    1979-03-01

    The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based on two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma

  13. The ARIES-I high-field-tokamak reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    Miller, R.L.

    1989-01-01

    The multi-institutional ARIES study has examined the physics, technology, safety, and economic issues associated with the conceptual design of a tokamak magnetic-fusion reactor. The ARIES-I variant envisions a DT-fueled device based on advanced superconducting coil, blanket, and power-conversion technologies and a modest extrapolation of existing tokamak physics. A comprehensive systems and trade study has been conducted as an integral and ongoing part of the reactor assessment in order to identify an acceptable design point to be subjected to detailed analysis and integration as well as to characterize the ARIES-I operating space. Results of parametric studies leading to the identification of such a design point are presented. 15 refs., 6 figs., 2 tabs

  14. Application of high temperature ceramic superconductors (CSC) to commercial tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Kim, S.; Gohar, Y.; Turner, L.; Smith, D.L.; Mattas, R.

    1987-10-01

    Ceramic superconductors operating near liquid nitrogen temperature may experience higher heating rates without losing stability, compared to conventional superconductors. This will permit cable design with less stabilizer, reducing fabrication costs for large fusion magnets. Magnet performance is studied for different operating current densities in the superconductor, and cost benefits to commercial tokamak reactors are estimated. It appears that 10 kA . cm -2 (at 77 K and ∼10 T) is a target current density which must be achieved in order for the ceramic superconductors to compete with conventional materials. At current densities around 50 kA . cm -2 most potential benefits have already been gained, as magnet structural steel begins to dominate the cost at this point. For a steady state reactor reductions of ∼7% are forecast for the overall capital cost of the power plant in the best case. An additional ∼3% cost saving is possible for pulsed tokamaks. 9 refs., 4 figs., 8 tabs

  15. Safety analysis of superconducting toroidal field magnet for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-02-01

    Safety analysis of the superconducting toroidal field magnet for a Tokamak experimental fusion reactor has been carried out. Works were accident classification, FMEA and FTA analyses, coil stability and quench behavior calculations, failure detection and coil protection system designs, structure analysis, fracture and fatigue studies, and earthquake response analysis. Accident analysis of cryostat and refrigeration system was also performed. The objective of this work is to reveal technological problems of the toroidal field magnet by safety analysis. (author)

  16. TRAIL: a Tokamak Rail Gun Limiter for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Powell, J R; Yu, W S; Fillo, J A; Usher, J L

    1980-01-01

    An attractive new limiter concept is investigated. The Tokamak Rail Gun Limiter (TRAIL) system impacts a stream of moderate velocity pellets (100 to 200 m/sec through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled, after cooling, to the injector in an E-M mass accelerator. Heat fluxes of approx. 30,000 W/cm/sup 2/ can be readily accommodated by the pellets, with very low recirculating power requirements (approx. 0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state-of-the-art (several km/sec). Accelerators injecting pellets at approx. 1 km/sec can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs.

  17. Review of deuterium--tritium results from the Tokamak Fusion Test Reactor*

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, K. M.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J. L.; Anderson, J W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, C. W.; Barnes, G.; Batha, S.; Bateman, G.; Beer, M; Bell, M. G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N. L.; Brunkhorst, C.; Budny, R.; Bush, C. E.; Camp, R.; Caorlin, M.; Carnevale, H.; Cauffman, S.; Chang, Z.; Chang, C. S.; Cheng, C. Z.; Chrzanowski, J.; Collins, J.; Coward, G.; Cropper, M.; Darrow, D. S; Daugert, R.; DeLooper, J.; Dendy, R.; Dorland, W.; Dudek, L.; Duong, H.; Durst, R.; Efthimion, P. C.; Ernst, D.; Evenson, H.; Fisch, N.; Fisher, R.; Fonck, R. J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G. Y.; Fujita, T.; Furth, H. P.; Garzotto, V.; Gentile, C.; Gilbert, J.; Gioia, J.; Gorelenkov, N.; Grek, B.; Grisham, L. R.; Hammett, G.; Hanson, G. R.; Hawryluk, R. J.; Heidbrink, W.; Herrmann, H. W.; Hill, K. W.; Hosea, J.; Hsuan, H.; Hughes, M.; Hulse, R.; Janos, A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kalish, M.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Labik, G.; Lam, N. T.; LaMarche, P. H.; Lawson, E.; LeBlanc, B.; Levine, J.; Levinton, F. M.; Loesser, D.; Long, D.; Loughlin, M. J.; Machuzak, J.; Majeski, R.; Mansfield, D. K.; Marmar, E. S.; Marsala, R.; Martin, A.; Martin, G.; Mazzucato, E.; Mauel, M.; McCarthy, M. P.; McChesney, J.; McCormack, B.; McCune, D. C.; McKee, G.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Mirnov, S. V.; Mueller, D.; Murakami, M.; Murphy, J. A.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Newman, R.; Norris, M.; O`Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D. K.; Park, H.; Park, W.; Parks, P.; Paul, S. F.; Pearson, G.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C. K.; Phillips, M.; Pitcher, S.; Pysher, R.; Qualls, A. L.; Raftopoulos, S.; Ramakrishnan, S.; Ramsey, A.; Rasmussen, D. A.; Redi, M. H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A. L.; Ruskov, E.; Sabbagh, S. A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G.; Scillia, R.; Scott, S. D.; Semenov, I.; Senko, T.; Sesnic, S.; Sissingh, R.; Skinner, C. H.; Snipes, J.; Stencel, J.; Stevens, J.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Swanson, J.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Terry, J.; Thompson, M. E.; Tighe, W.; Timberlake, J. R.; Tobita, K.; Towner, H. H.; Tuszewski, M.; Halle, A. Von; Vannoy, C.; Viola, M.; Goeler, S. Von; Voorhees, D.; Walters, R. T.; Wester, R.; White, R.; Wieland, R.; Wilgen, J. B.; Williams, M.; Wilson, J. R.; Winston, J.; Wright, K.; Wong, K. L.; Woskov, P.; Wurden, G. A.; Yamada, M.; Yoshikawa, S.; Young, K. M.; Zarnstorff, M. C.; Zavereev, V.; Zweben, S. J.

    1995-01-01

    The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≈ 2.8 MW m₋3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni (0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

  18. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles

  19. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    Energy Technology Data Exchange (ETDEWEB)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles.

  20. Status of Tokamak Fusion Test Reactor neutron activation

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Rule, K. [Princeton Plasma Physics Lab., NJ (United States)

    1996-12-31

    Measurements have been made following TFTR D-T campaigns to characterize the behavior of D-T fusion reactor neutron activation using Ionization Chamber, Geiger Mueller, and Ge detector gamma-ray spectroscopy measurements. The results exhibit decay rates characteristic of the materials and geometries of the Test Cell hardware, and allow extrapolation to higher fusion power yields. The results can be used for benchmarking D-T fusion reactor activation simulations for accurate determinations of low activation long-lived cooling. 5 refs., 8 figs.

  1. Reactor aspects of counterstreaming-ion tokamak plasmas

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1975-06-01

    Toroidal DT plasmas in which the D and T ions make up two distinct, quasi-thermal velocity distributions, oppositely displaced in velocity along the magnetic axis, are discussed. Such counterstreaming distributions can be set up by introducing all ions by tangential injection of neutral beams, and by removing ions from the plasma shortly after they have decelerated to an energy approximate to or less than 2T/sub e/ by Coulomb drag on the plasma electrons. A simple physical model for counterstreaming-ion operation is postulated, which allows one to deduce the ion velocity distributions and required energy and particle confinement times that are in good agreement with the results of previous Fokker-Planck calculations. The variations of fusion reactivity, power gain, and power density with injection energy and electron temperature are presented. The practical problems of implementing counter-streaming operation in a tokamak, such as charge-exchange losses, the prompt removal of cold ions, and the effect of impurities are discussed. (U.S.)

  2. Recent developments in engineering and technology concepts for prospective tokamak fusion reactors

    International Nuclear Information System (INIS)

    Ford, G.W.K.

    1987-01-01

    The tokamak has become the most developed magnetic fusion system and it appears likely that break-even and possibly ignition will first be demonstrated in existing machines of this type. Yet larger tokamaks could also demonstrate the essential technologies for the production of useful power. World-wide, well over a hundred tritium-breeder/heat-removal blanket concepts have been devised and preliminary engineering design studies undertaken, but the effort deployed on breeding and power recovery systems has been very small compared with that assigned to plasma research and development. The European Communities' NET (Next European Torus) project may offer an opportunity to redress this imbalance. The NET pre-design stage now in progress for some three years has selected many of the best features of plasma and nuclear design from the world's total efforts in these fields, and the NET concept is described in this paper as exemplifying where magnetic fusion power reactor technology stands today. It is concluded that although there are numerous more advanced types of magnetic confinement fusion reactor at early stages of their physics development, the tokamak offers the best opportunity for the early demonstration of fusion power

  3. Some stress-related issues in tokamak fusion reactor first walls

    International Nuclear Information System (INIS)

    Majumdar, S.; Pai, B.; Ryder, R.H.

    1987-01-01

    Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m 2 , and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems

  4. Maintenance oriented tokamak reactor with low activation material and high aspect ratio configuration

    International Nuclear Information System (INIS)

    Nishio, S.; Ueda, S.; Aoki, I.

    1997-01-01

    Some of the difficulties of maintenance of a tokamak system are due to the radioactivity of its materials and the complexity of its configuration. To mitigate the difficulties, very low activation SiC/SiC composite material and a torus configuration with a high aspect ratio of eight are introduced. The torus system is divided radially into equal sectors, each of which forms an assembling unit. All piping and feeder systems are shifted to the spacious central torus region. The new concept is called DREAM tokamak reactor, which is an acronym for DRastically EAsy Maintenance. The maximum toroidal field strength of 20 T and the plasma major and minor radii of 16 m and 2 m lead to a high fusion power of 5.5 GW, under moderate physics constraints such as an MHD safety factor of three, a Troyon factor of three for the beta limit and an energy confinement time enhancement factor of two. (author)

  5. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Lee, K. W.; Hwang, C. K.; Hong, B. G.; Hong, G. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-05-01

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new.

  6. Tokamak reactor for treating fertile material or waste nuclear by-products

    Science.gov (United States)

    Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

    2012-10-02

    Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

  7. Status report on the conceptual design of a commercial tokamak hybrid reactor (CTHR)

    International Nuclear Information System (INIS)

    1979-09-01

    A preliminary conceptual design is presented for an early twenty-first century fusion hybrid reactor called the Commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) plants. The study has been made in sufficient depth to indicate no insurmountable technical problems exist and has provided a basis for valid cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources

  8. Status report on the conceptual design of a commercial tokamak hybrid reactor (CTHR)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-09-01

    A preliminary conceptual design is presented for an early twenty-first century fusion hybrid reactor called the Commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) plants. The study has been made in sufficient depth to indicate no insurmountable technical problems exist and has provided a basis for valid cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources.

  9. A commercial tokamak reactor using super high field superconducting magnets

    International Nuclear Information System (INIS)

    Schwartz, J.; Bromberg, L.; Cohn, D.R.; Williams, J.E.C.

    1988-01-01

    This paper explores the range of possibilities for producing super high fields with advanced superconducting magnets. Obtaining magnetic fields greater than about 18 T at the coil in a large superconducting magnet system will require advances in many areas of magnet technology. These needs are discussed and potential solutions (advanced superconductors, structural materials and design methods) evaluated. A point design for a commercial reactor with magnetic field at the coil of 24 T and fusion power of 1800 MW is presented. Critical issues and parameters for magnet design are identified. 20 refs., 9 figs., 4 tabs

  10. An advanced conceptual Tokamak fusion power reactor utilizing closed cycle helium gas turbines

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    UWMAK-III is a conceptual Tokamak reactor designed to study the potential and the problems associated with an advanced version of Tokamaks as power reactors. Design choices have been made which represent reasonable extrapolations of present technology. The major features are the noncircular plasma cross section, the use of TZM, a molybdenum based alloy, as the primary structural material, and the incorporation of a closed-cycle helium gas turbine power conversion system. A conceptual design of the turbomachinery is given together with a preliminary heat exchanger analysis that results in relatively compact designs for the generator, precooler, and intercooler. This paper contains a general description of the UWMAK-III system and a discussion of those aspects of the reactor, such as the burn cycle, the blanket design and the heat transfer analysis, which are required to form the basis for discussing the power conversion system. The authors concentrate on the power conversion system and include a parametric performance analysis, an interface and trade-off study and a description of the reference conceptual design of the closed-cycle helium gas turbine power conversion system. (Auth.)

  11. Potential minimum cost of electricity of superconducting coil tokamak power reactors

    International Nuclear Information System (INIS)

    Reid, R.L.; Peng, Y-K. M.

    1989-01-01

    The potential minimum cost of electricity (COE) for superconducting tokamak power reactors is estimated by increasing the physics (confinement, beta limit, bootstrap current fraction) and technology [neutral beam energy, toroidal field (TF) coil allowable stresses, divertor heat flux, superconducting coil critical field, critical temperature, and quench temperature rise] constraints far beyond those assumed for ITER until the point of diminishing returns is reached. A version of the TETRA systems code, calibrated with the ITER design and modified for power reactors, is used for this analysis, limiting this study to reactors with the same basic device configuration and costing algorithms as ITER. A minimum COE is reduced from >200 to about 80 mill/kWh when the allowable design constraints are raised to 2 times those of ITER. At 4 times the ITER allowables, a minimum COE of about 60 mill/kWh is obtained. The corresponding tokamak has a major radius of approximately 4 m, a plasma current close to 10 MA, an aspect ratio of 4, a confinement H- factor ≤3, a beta limit of approximately 2 times the first stability regime, a divertor heat flux of about 20 MW/m 2 , a Β max ≤ 18 T, and a TF coil average current density about 3 times that of ITER. The design constraints that bound the minimum COE are the allowable stresses in the TF coil, the neutral beam energy, and the 99% bootstrap current (essentially free current drive). 14 refs., 4 figs., 2 tabs

  12. Tokamak power systems studies, FY 1986: A second stability power reactor

    International Nuclear Information System (INIS)

    Ehst, D.; Baker, C.; Billone, M.

    1987-03-01

    This report presents the results of the work at Argonne National Laboratory (ANL) during FY-1986 on the Tokamak Power Systems Study (TPSS). The purpose of the TPSS is to explore and develop ideas that would lead to improvements in the tokamak as a power reactor concept. The work at ANL concentrated on plasma engineering, impurity control, and the blanket/first wall/shield system. The work in FY-1986 extended these studies and focused them on a reference design point. The key features of the design point include: second stability regime with higher β and larger aspect ratio, steady-state operation with fast wave current drive, impurity control via a self-pumped slot limiter, a self-cooled liquid lithium, vanadium alloy blanket with simplified poloidal flow, and reduced reactor building volume with vertical lift maintenance. Sufficient work was carried out to report a preliminary cost estimate. In addition, reactor implications of steady-state operation in the first stability regime were also studied. 174 refs., 124 figs., 65 tabs

  13. Conceptual integrated approach for the magnet system of a tokamak reactor

    International Nuclear Information System (INIS)

    Duchateau, J.-L.; Hertout, P.; Saoutic, B.; Artaud, J.-F.; Zani, L.; Reux, C.

    2014-01-01

    Highlights: • We give a conceptual approach of a fusion reactor magnet system based on analytical formula. • We give design criteria for the CS and TF cable in conduit conductors and for the magnet system structural description. • We apply this conceptual approach to ITER and we crosscheck with actual characteristics. • We apply this conceptual approach to a possible version of DEMO. - Abstract: In the framework of the reflexion about DEMO, a conceptual integrated approach for the magnet system of a tokamak reactor is presented. This objective is reached using analytical formulas which are presented in this paper, coupled to a Fortran code ESCORT (Electromagnetic Superconducting System for the Computation of Research Tokamaks), to be integrated into SYCOMORE, a code for reactor modelling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modelling task force. The analytical formulas deal with all aspects of the magnet system, starting from the derivation of the TF system general geometry, from the plasma main characteristics. The design criteria for the cable current density and the structural design of the toroidal field and central solenoid systems are presented, enabling to deliver the radial thicknesses of the magnets and enabling also to estimate the plasma duration of the plateau. As a matter of fact, a pulsed version DEMO is presently actively considered in the European programmes. Considerations regarding the cryogenics and the protection are given, affecting the general design. An application of the conceptual approach is presented, allowing a comparison between ESCORT output data and actual ITER parameters and giving the main characteristics of a possible version for DEMO

  14. 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

  15. The impact of servicing requirements on Tokamak fusion reactor design

    International Nuclear Information System (INIS)

    Mitchell, J.T.D.

    1978-01-01

    There is one design requirement which appears to be common to all D-T fusion reactors, irrespective of the plasma confinement principle embodied in the design. It must be practicable to repair the structure or 'first wall' nearest the plasma. A repair can include a range of operations between finding and sealing a pinhole leak in a weld to complete replacement of the whole structure of the breeding blanket. The two reasons why these repairs are necessary are: 1) The variability and defects in generally perfect products, such as those caused by errors of human performance or by flaws in raw materials. 2) The extreme severity of the environment of the first wall, which causes the deterioration of the properties of materials situated there. This deterioration is the result of neutron irradiation and transmutation in the materials under conditions of stress and high temperature and of surface corrosion and erosion

  16. First wall and limiter lifetime in pulsed tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.; Majumdar, S.; Cha, Y.; Hassanein, A.

    1983-12-01

    This study concentrates on the structural integrity of certain reactor subsystems under cyclic operation to answer the question: how long a burn pulse is needed to achieve the benefits of steady-state operation. Component lifetime in the steady-state is limited by three effects: radiation damage, disruptions, and sputtering erosion. Cyclic operation modifies one of these (the number of disruptions may increase with the number of burn cycles) and introduces a fourth life limit, thermal fatigue. Our design strategy is to determine the structure and coating thicknesses which maximize component lifetime against all life limitations. After calculating disruption damage (vaporization, melting) for candidate materials we present the lifetime analysis for different structures

  17. Superconducting coil design for a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Turner, L.R.; Wang, S.T.; Smelser, P.

    1977-01-01

    Superconducting toroidal field (TF) and polodial-field (PF) coils have been designed for the proposed Argonne National Laboratory experimental power reactor (EPR). Features of the design include: (1) Peak field of 8 T at 4.2 K or 10 T at 3.0 K. (2) Constant-tension shape for the TF coils, corrected for the finite number (16) of coils. (3) Analysis of errors in coil alignment. (4) Comparison of safety aspects of series-connected and parallel-connected coils. (5) A 60 kA sheet conductor of NbTi with copper stabilizer and stainless steel for support. (6) Superconducting PF coils outside the TF coils. (7) The TF coils shielded from pulsed fields by high-purity aluminum

  18. 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

  19. The effective cost of tritium for tokamak fusion power reactors with reduced tritium production systems

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Evans, K.

    1983-01-01

    If sufficient tritium cannot be produced and processed in tokamak blankets then at least two alternatives are possible. Tritium can be purchased; or reactors with reduced tritium (RT) content in the plasma can be designed. The latter choice may require development of magnet technology etc., but the authors show that the impact on the cost-of-electricity may be mild. Cost tradeoffs are compared to the market value of tritium. Adequate tritium production in fusion blankets is preferred, but the authors show there is some flexibility in the deployment of fusion if this is not possible

  20. First wall and blanket design for the STARFIRE commercial tokamak power reactor

    International Nuclear Information System (INIS)

    Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

    1979-01-01

    The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

  1. Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

    International Nuclear Information System (INIS)

    Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

    1976-01-01

    This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended

  2. The ARIES-III D-3He tokamak reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1991-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-3 design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. Results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-1 is included. 11 refs., 5 figs

  3. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    CERN Document Server

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  4. Neutron Activation Cool-down of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ascione, G.; Kugel, H.W.; Kumar, A.; Tilson, Jr, C.

    1998-06-10

    Tokamak Fusion Test Reactor (TFTR) final operations and post-shutdown neutron activation measurements were made. Ionization chambers were used to follow TFTR activation during operations and after shutdown. Gamma-ray energy spectroscopy measurements were performed to characterize TFTR activation at accessible vessel-bays and on sample hardware removed from structures at various distances from the vessel. The results demonstrate long-lived activations from common, commercially available materials used in the fabrication and field engineering of TFTR. The measurements allow characterization of residual TFTR neutron activation, the projection of residual activation decay, and benchmarking of low activation issues simulations.

  5. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    M.E. Lumia; C.A. Gentile

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed

  6. Design constraints for rf-driven steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1979-02-01

    Plasma current density profiles are computed due to electron Landau damping of lower hybrid waves launched into model tokamak density and temperature profiles. The total current and current profile shape are chosen consistent with magnetohydrodynamic equilibrium for a variety of temperature and density distributions and plasma beta values. Surface current equilibria appear attractive and are accessible to waves with n/sub z/ as low as 1.2. By suitably choosing the spectrum location and width it is possible to drive the 9.8 MA current of a 7.0-m reactor with as little as 2.8% of the fusion power recirculated as rf input from the waveguides

  7. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    M.E. Lumia; C.A. Gentile

    2002-01-18

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  8. Internal transport barrier in tokamak and helical plasmas

    Science.gov (United States)

    Ida, K.; Fujita, T.

    2018-03-01

    The differences and similarities between the internal transport barriers (ITBs) of tokamak and helical plasmas are reviewed. By comparing the characteristics of the ITBs in tokamak and helical plasmas, the mechanisms of the physics for the formation and dynamics of the ITB are clarified. The ITB is defined as the appearance of discontinuity of temperature, flow velocity, or density gradient in the radius. From the radial profiles of temperature, flow velocity, and density the ITB is characterized by the three parameters of normalized temperature gradient, R/{L}T, the location, {ρ }{ITB}, and the width, W/a, and can be expressed by ‘weak’ ITB (small R/{L}T) or ‘strong’ (large R/{L}T), ‘small’ ITB (small {ρ }{ITB}) or ‘large’ ITB (large {ρ }{ITB}), and ‘narrow’ (small W/a) or ‘wide’ (large W/a). Three key physics elements for the ITB formation, radial electric field shear, magnetic shear, and rational surface (and/or magnetic island) are described. The characteristics of electron and ion heat transport and electron and impurity transport are reviewed. There are significant differences in ion heat transport and electron heat transport. The dynamics of ITB formation and termination is also discussed. The emergence of the location of the ITB is sometimes far inside the ITB foot in the steady-state phase and the ITB region shows radial propagation during the formation of the ITB. The non-diffusive terms in momentum transport and impurity transport become more dominant in the plasma with the ITB. The reversal of the sign of non-diffusive terms in momentum transport and impurity transport associated with the formation of the ITB reported in helical plasma is described. Non-local transport plays an important role in determining the radial profile of temperature and density. The spontaneous change in temperature curvature (second radial derivative of temperature) in the ITB region is described. In addition, the key parameters of the control of the

  9. Four ignition TNS tokamak reactor systems: design summary

    International Nuclear Information System (INIS)

    Flanagan, C.A.

    1977-10-01

    Principal TNS objectives assumed included: (1) demonstration of ignition and burning dynamics; and (2) reactor technology forcing. The selection of an overall design approach for TNS required an early quantitative assessment of the most important design issues; namely, choice of ignition plasma design conditions (principally size and confining field of axis), and choice of toroidal field coil technology (resistive or superconducting windings). The design space investigated in this study ranged from ignited plasmas (elongated) with minor radii varying between 0.8 m (TFTR-like) and approximately 2.0 m (EPR-like). Four TF coil types were examined; these included copper, NbTi, Nb 3 Sn, and a hybrid design employing nested coils of copper and NbTi. A final step involved a further comparison of the four reference concepts using decision modeling techniques as a mechanism for selecting a preferred design approach for the TNS mission. Section 3.0 describes the TNS study process. Section 4.0 presents a summary of the parameters for the four reference point designs. Finally, Section 5.0 presents a brief description of the design features of many of the systems comprising the TNS design

  10. International cooperation on breeder reactors

    International Nuclear Information System (INIS)

    Gray, J.E.; Kratzer, M.B.; Leslie, K.E.; Paige, H.W.; Shantzis, S.B.

    1978-01-01

    In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

  11. Internal m=1, n=1 helical mode in a tokamak with nonmonotonic current profile

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Mikhajlovskij, A.B.

    1988-01-01

    Internal helical mode in a tokamak with two resonance surfaces, on which storing coefficient reduces to unity is studied theoretically. A general criterion for the investigated perturbations stability is obtained. Dispersion equation, describing both ideal and resistive helical modes, is derived. Analytic calculations for the case of perturbations localized near the tokamak axis are made. It is shown that in the framework of standard ideal hydrodynamics such perturbations are unstable at characteristic nonmonotonous profiles of the current

  12. Design study of electrical power supply system for tokamak fusion power reactor

    International Nuclear Information System (INIS)

    1977-01-01

    Design study of the electrical power supply system for a 2000MWt Tokamak-type fusion reactor has been carried out. The purposes are to reveal and study problems in the system, leading to a plan of the research and development. Performed were study of the electrical power supply system and design of superconducting inductive energy storages and power switches. In study of the system, specification and capability of various power supplies for the fusion power reactor and design of the total system with its components were investigated. For the superconducting inductive energy storages, material choice, design calculation, and structural design were conducted, giving the size, weight and performance. For thyristor switches, circuit design in the parallel / series connection of element valves and cooling design were studied, providing the size and weight. (auth.)

  13. What is past is prologue: future directions in Tokamak Power Reactor Design Research

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    After reviewing the first generation of studies and the primary conclusions they produced, four current designs are discussed that are representative of present trends in this area of research. In particular, the trends towards reduced reactor size and higher neutron wall loadings are discussed. Moving in this direction requires new approaches to many subsystem designs. New approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets are described. A discussion is given of the future role of conceptual reactor design research and the need for close interactions with ongoing experiments in fusion technology

  14. Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

    International Nuclear Information System (INIS)

    Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.; Attenberger, S.E.

    1988-01-01

    The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor plasma (Tokamak Ignition/Burn Experimental Reactor) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-dimensional transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alpha concentration significantly influence the ignition and steady-state burn capability

  15. D-3He tokamak reactor with inductive a.c. and unidirectional current operation

    International Nuclear Information System (INIS)

    Mitarai, O.

    1995-01-01

    We propose inductively operated D- 3 He tokamak reactors (R=9.5m, a=2.8 and 3.2m, B t =10 and 7.5T, and I p =53.6 and 52.5MA respectively) with an a.c. or unidirectional current (UDC) mode, operated in the first stability regime, with a small current drive power of about 500kW in the plasma side, and about 2MW in the transformer side. The high electron temperature T e (0)∼62.5keV together with a high bootstrap current fraction greater than 70% decreases the plasma resistance (to less than 1nΩ) and the loop voltage (less than 10mV range) and then prolongs the pulse length to more than 15h. The ignition characteristics of a D- 3 He tokamak reactor are analyzed by the operation path method on the P ht τ E 2 -T plane and POPCON for different confinement scalings. ((orig.))

  16. Application of high temperature ceramic superconductors (CSC) to commercial tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Kim, S.; Gohar, Y.; Turner, L.; Smith, D.L.; Mattas, R.

    1988-08-01

    Ceramic superconductors operating near liquid nitrogen temperature may experience higher heating rates without losing stability, compared conventional superconductors. This will permit cable design with less stabilizer, reducing fabrication costs for large fusion magnets. Magnet performance is studied for different operating current densities in the superconductor, and cost benefits to commercial tokamak reactors are estimated. It appears that 10 kA /center dot/ cm/sup /minus/2/ (at 77 K and /approximately/10 T) is a target current density which must be achieved in order for the ceramic superconductors to compete with conventional materials. At current densities around 50 kA /center dot/ cm/sup /minus/2/ most potential benefits have already been gained, as magnet structural steel begins to dominate the cost at this point. For a steady state reactor reductions of /approximately/7% are forecast for the overall capital cost of the power plant in the best case. An additional /approximately/3% cost saving is possible for pulsed tokamaks. 9 refs., 4 figs., 8 tabs

  17. Deuterium-Tritium Simulations of the Enhanced Reversed Shear Mode in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D.R.; Manickam, J.; Scott, S.D.; Zarnstorff

    1997-04-01

    The potential performance, in deuterium-tritium plasmas, of a new enhanced con nement regime with reversed magnetic shear (ERS mode) is assessed. The equilibrium conditions for an ERS mode plasma are estimated by solving the plasma transport equations using the thermal and particle dif- fusivities measured in a short duration ERS mode discharge in the Tokamak Fusion Test Reactor [F. M. Levinton, et al., Phys. Rev. Letters, 75, 4417, (1995)]. The plasma performance depends strongly on Zeff and neutral beam penetration to the core. The steady state projections typically have a central electron density of {approx}2:5x10 20 m{sup -3} and nearly equal central electron and ion temperatures of {approx}10 keV. In time dependent simulations the peak fusion power, {approx} 25 MW, is twice the steady state level. Peak performance occurs during the density rise when the central ion temperature is close to the optimal value of {approx} 15 keV. The simulated pressure profiles can be stable to ideal MHD instabilities with toroidal mode number n = 1, 2, 3, 4 and {infinity} for {beta}{sub norm} up to 2.5; the simulations have {beta}{sub norm} {le} 2.1. The enhanced reversed shear mode may thus provide an opportunity to conduct alpha physics experiments in conditions imilar to those proposed for advanced tokamak reactors.

  18. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  19. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. DREAM (DRastically EAsy Maintenance) tokamak

    International Nuclear Information System (INIS)

    Nishio, Satoshi

    1998-01-01

    If the major part of the electric power demand will be supplied by tokamak fusion power plants, a suitable tokamak reactor must be an ultimate goal, i.e., the reactor must be excellent both in terms of construction cost and safety aspects including operation availability (maintainability and reliability). In attaining this goal, an approach focusing on both safety and availability (including reliability and maintainability) issues is the most promising strategy. The tokamak reactor concept with a very high aspect ratio configuration and SiC/SiC composite structural materials is compatible with this approach, which is called the DREAM (DRastically EAsy Maintenance) approach. The SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to good accessibility for the maintenance of machines. As an intermediate steps between an experimental reactor such as ITER and the ultimate goal, the development of prototype reactor which demonstrates electric power generation and an initial-phase commercial reactor which demonstrates for COE (cost of electricity) competitiveness has been investigated. Especially for the prototype reactor, material and technological immaturity must be considered. (J.P.N.)

  20. The international thermonuclear reactor (ITER)

    International Nuclear Information System (INIS)

    Fowler, T.K.; Henning, C.D.

    1987-01-01

    Four governmental groups, representing Europe, Japan, USSR and U.S. met in March 1987 to consider a new international design of a magnetic fusion device for the 1990's. An interim group was appointed. The author gives a brief synopsis of what might be thought of as a draft charter. The starting point is the objective of the ITER device, which is summarized as demonstrating both scientific and technical feasibility of fusion. The paper presents an update on the current thinking and technical aspects for the International Thermonuclear Experimental Reactor (ITER). This covers not only what is happening in the U.S. but also some reports of preliminary thinking of the last technical work that occurred in Vienna

  1. ORNL TNS Program: plasma engineering considerations and innovations for a medium field tokamak fusion reactor

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Attenberger, S.E.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

    1977-12-01

    Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma anti β values up to 10%, and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors; some of the suggested innovations will be tested in upcoming experiments

  2. Safety in the ARIES-III D-3He tokamak reactor design

    International Nuclear Information System (INIS)

    Herring, J.S.; Dolan, T.J.

    1992-01-01

    This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

  3. Internal magnetic field measurement in tokamak plasmas using a ...

    Indian Academy of Sciences (India)

    Abstract. In a tokamak plasma, the poloidal magnetic field profile closely depends on the current density profile. ... Transitions between the energy levels that are split by the magnetic field give rise to several components of a spectral ... train of charged pulses having an amplitude proportional to the light intensity sensed by.

  4. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  5. Technological and environmental prospects of low aspect ratio tokamak reactor VECTOR

    International Nuclear Information System (INIS)

    Nishio, S.; Tobita, K.; Isono, T.; Nakamura, Y.; Sato, M.; Sakurai, S.; Yamauchi, M.; Nishitani, T.; Tani, K.; Sengoku, S.; Kudo, Y.; Song, Y.; Tokimatsu, K.; Shinya, K.; Senda, I.; Konishi, S.

    2005-01-01

    Optimization of VECTOR design parameters has led to a fusion output of 2.5 GW with a small reactor weight of 8,800 tons. Current ramp-up simulation with TSC (Tokamak Simulation Code) demonstrated a stable current ramp-up with externally non-inductive current drive and bootstrap current when a central current is induced externally enough to avoid an extreme reversed shear such as current hole. It was numerically confirmed that such a low-A reactor would have an advantage of α-particle confinement. The divertor is designed to use a flux expansion of scrape-off-layer in the divertor region to maintain the heat flux on the divertor plate to be lower than 10 MW/m 2 without remote radiative cooling. Because of its compactness, CO 2 emission in a life cycle of a VECTOR power plant is estimated to be as low as 3.2 g-CO 2 /kWh, being lower than that of an ITER-sized DEMO reactor (4.9 g-CO 2 /kWh). As to the waste management of VECTOR, on the basis of reactor design and radiological considerations, we suggest reusing a liquid metal breeding material (PbLi) and neutron shield material (TiH 2 ) in successive reactors. Due to this waste management, its disposal waste would be reduced to as low as 3,000-4,000 tons, which is comparable with the radioactive waste of a light water reactor (4,000 tons in metal). (author)

  6. Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

    International Nuclear Information System (INIS)

    Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

    1995-01-01

    The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

  7. Materials and components for X-ray diagnostic use on the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Moshey, E.A.

    1982-01-01

    X-ray diagnostic equipment will operate on Princeton's Tokamak Fusion Test Reactor (TFTR) during Hydrogen (HH and DH), Deuterium (DD) and Tritium (DT) discharges. The environmental requirements on diagnostic equipment with direct conductance to the Tokamak's vacuum are demanding. The materials and components will be subjected to: (a) ultra-high vacuum of 1 x 10 -8 torr, (b) temperature cycling from 15 0 C to 250 0 C, (c) radiation to 1 x 10 8 rads, (d) magnetic fields to 6 Tesla. In addition, selection of materials must also be based upon minimizing the formation of significant quantities of long lived radioactive elements created by the bombardment of 14 MeV neutrons. Shielding materials must also meet flammability requirements. This paper deals with the selection of materials and components as used on the TFTR X-ray Imaging Systems and the TFTR Pulse-Height Analysis Systems. The trade-offs that led to selection of materials are discussed. Test results and sources of test data are presented

  8. A Study of Performance in Low-Power Tokamak Reactor with Integrated Predictive Modeling Code

    International Nuclear Information System (INIS)

    Pianroj, Y.; Onjun, T.; Suwanna, S.; Picha, R.; Poolyarat, N.

    2009-07-01

    Full text: A fusion hybrid or a small fusion power output with low power tokamak reactor is presented as another useful application of nuclear fusion. Such tokamak can be used for fuel breeding, high-level waste transmutation, hydrogen production at high temperature, and testing of nuclear fusion technology components. In this work, an investigation of the plasma performance in a small fusion power output design is carried out using the BALDUR predictive integrated modeling code. The simulations of the plasma performance in this design are carried out using the empirical-based Mixed Bohm/gyro Bohm (B/gB) model, whereas the pedestal temperature model is based on magnetic and flow shear (δ α ρ ζ 2 ) stabilization pedestal width scaling. The preliminary results using this core transport model show that the central ion and electron temperatures are rather pessimistic. To improve the performance, the optimization approach are carried out by varying some parameters, such as plasma current and power auxiliary heating, which results in some improvement of plasma performance

  9. Study of an optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak

    International Nuclear Information System (INIS)

    Hong, Bong Guen; Kim, Hoseok

    2016-01-01

    Highlights: • Optimum configuration of a transmutation reactor based on a low aspect ratio tokamak was found. • Inboard and outboard radial build are determined by plasma physics, engineering and neutronics constraints. • Radial build and equilibrium fuel cycle play a major role in determining the transmutation characteristics. - Abstract: We determine the optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak. For self-consistent determination of the radial build of the reactor components, we couple a tokamak systems analysis with a radiation transport calculation. The inboard radial build of the reactor components is obtained from plasma physics and engineering constraints, while outboard radial builds are mainly determined by constraints on neutron multiplication, the tritium-breeding ratio, and the power density. We show that the breeding blanket model has an effect on the radial build of a transmutation blanket. A burn cycle has to be determined to keep the fast neutron fluence plasma-facing material below its radiation damage limit. We show that the radial build of the transmutation reactor components and the equilibrium fuel cycle play a major role in determining the transmutation characteristics.

  10. Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor

    International Nuclear Information System (INIS)

    Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.

    1976-01-01

    As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield

  11. Hydrogen isotope exchange and conditioning in graphite limiters used in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    LaMarche, P.H.; Dylla, H.F.; McCarthy, P.J.; Ulrickson, M.

    1986-01-01

    Isotopic exchange experiments performed in the Tokamak Fusion Test Reactor (TFTR) are used to examine the outgassing and diffusive properties of graphite used as the plasma limiter. Changeover from hydrogen to deuterium for different periods ranges from approx.600 to 60 plasma discharges, which appears to be correlated in the limiter temperature. We present a simple analytical model that predicts a fast transient (approx.10 plasma discharges) changeover, where the deuterium fueling dilutes the adsorbed and near surface hydrogen, and a slowly changing term where bulk hydrogen diffuses to the surface. Using this model we can extract an activation energy for diffusion of 0.15 +- 0.02 eV. We hypothesize that interpore diffusion for this porous (approx.15%) material is consistent with our observations

  12. Initial confinement studies of ohmically heated plasmas in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Bell, M.; Blanchard, W.R.

    1984-06-01

    Initial operation of the Tokamak Fusion Test Reactor (TFTR) has concentrated upon confinement studies of ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (tau/sub E/) are 0.1 to 0.2 s for a line-average density range (anti n/sub e/) of 1 to 2.5 x 10 19 m -3 with electron temperatures of T/sub e/(o) approx. 1.2 to 2.2 keV, ion temperatures of T/sub i/(o) approx. 0.9 to 1.5 keV, and Z/sub eff/ approx. 3. A comparison of PLT, PDX, and TFTR plasma confinement supports a dimension-cubed scaling law

  13. Measurements of tritium retention and removal on the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Kamperschroer, J.; LaMarche, P.; Mueller, D.; Nagy, A.; Scott, S.; Ascione, G.; Amarescu, E.; Camp, R.; Casey, M.; Collins, J.; Cropper, M.; Gentile, C.; Gibson, M.; Hosea, J.; Kalish, M.; Langford, J.; Langish, S.; Mika, R.; Owens, D.K.; Pearson, G.; Raftopoulos, S.; Raucci, R.; Stevenson, T.; Von Halle, A.; Voorhees, D.; Walters, T.; Winston, J.

    1996-01-01

    Recent experiments on the Tokamak Fusion Test Reactor have afforded an opportunity to measure the retention of tritium in a graphite limiter that is subject to erosion, codeposition, and high neutron flux. The tritium was injected by both gas puff and neutral beams. The isotopic mix of hydrogenic recycling was measured spectroscopically and the tritium fraction T/(H+D+T) transiently increased to as high as 75%. Some tritium was pumped out during the experimental run and some removed in a subsequent campaign using various clean-up techniques. While the short term retention of tritium was high, various conditioning techniques were successful in removing ≅8000 Ci and restoring the tritium inventory to a level well below the administrative limit. copyright 1996 American Vacuum Society

  14. Development of titanium alloy metallic balloon for locking system in tokamak fusion reactor

    International Nuclear Information System (INIS)

    Sato, S.; Nishio, S.

    1995-01-01

    The development of a locking structure of in-vessel components such as blanket vessels is one of the key issues for a tokamak fusion reactor. A concept of a locking structure with a cotter (wedge) and cotter driver has been proposed. We considered the transformation of a metallic balloon (MB) with thin wall and flat shape cross-section very promising as a cotter driver. We tried to fabricate the MB made of different kinds of titanium alloy by superplastic forming (SPF). Several mockups of the MB were successfully fabricated and used for performance tests. A mockup of the locking structure with blanket vessels, cotter structures and MBs has also been fabricated. Using this mockup, the locking mechanism was verified through an experimental study. (orig.)

  15. Measurements of tritium retention and removal on the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H.; Blanchard, W.; Kamperschroer, J.; LaMarche, P.; Mueller, D.; Nagy, A.; Scott, S.; Ascione, G.; Amarescu, E.; Camp, R.; Casey, M.; Collins, J.; Cropper, M.; Gentile, C.; Gibson, M.; Hosea, J.; Kalish, M.; Langford, J.; Langish, S.; Mika, R.; Owens, D.K.; Pearson, G.; Raftopoulos, S.; Raucci, R.; Stevenson, T.; Von Halle, A.; Voorhees, D.; Walters, T.; Winston, J. [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States)

    1996-11-01

    Recent experiments on the Tokamak Fusion Test Reactor have afforded an opportunity to measure the retention of tritium in a graphite limiter that is subject to erosion, codeposition, and high neutron flux. The tritium was injected by both gas puff and neutral beams. The isotopic mix of hydrogenic recycling was measured spectroscopically and the tritium fraction T/(H+D+T) transiently increased to as high as 75{percent}. Some tritium was pumped out during the experimental run and some removed in a subsequent campaign using various clean-up techniques. While the short term retention of tritium was high, various conditioning techniques were successful in removing {approx_equal}8000 Ci and restoring the tritium inventory to a level well below the administrative limit. {copyright} {ital 1996 American Vacuum Society}

  16. Numerical calculation of high frequency fast wave current drive in a reactor grade tokamak

    International Nuclear Information System (INIS)

    Ushigusa, Kenkichi; Hamamatsu, Kiyotaka

    1988-02-01

    A fast wave current drive with a high frequency is estimated for a reactor grade tokamak by the ray tracing and the quasi-linear Fokker-Planck calculations with an assumption of single path absorption. The fast wave can drive RF current with the drive efficiency of η CD = n-bar e (10 19 m -3 )I RC (A)R(m)/P RF (W) ∼ 3.0 when the wave frequency is selected to be f/f ci > 7. A sharp wave spectrum and a ph|| >/υ Te ∼ 3.0 are required to obtain a good efficiency. A center peaked RF current profile can be formed with an appropriate wave spectrum even in the high temperature plasma. (author)

  17. Parameter study of high-β tokamak reactors with circular and strongly elongated cross section

    International Nuclear Information System (INIS)

    Herold, H.

    1977-05-01

    A simplified reactor model is used to study the influence of critical β-values on economy parameters and dimensions of possible long time pulsed tokamak reactors. Various betas deduced from stability and equilibrium MHD theory are introduced and put into the scaling in context with technological constraints, as maximum B-field, core constraint, maximum wall loading a.o. The plasma physical concepts treated comprise circular and strongly elongated cross section and approximated FCT equilibria. The computational results are presented as plots of possible economy parameter ranges (magnet energy, wall loading, volumina, investment costs per unit power) dependent on β for suitably chosen hierarchies of the constraints. - A burn time reduction by the build ups of α-pressure may be possible for the pressure profile sensitive high-β equilibria (FCT). Burn times in the 1O sec range, resulting from simple estimates, would about cancel the economic advantages of reactors with high-β equilibria compared to a β = 5% standardreactor (UWMAK I). (orig.) [de

  18. Development of integrated insulation joint for cooling pipe in tokamak reactor

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Abe, Tetsuya; Kawamura, Masashi; Yamazaki, Seiichiro.

    1994-08-01

    In a tokamak fusion reactor, an electrically insulated part is needed for an in-vessel piping system in order to break an electric circuit loop. When a closed loop is formed in the piping system, large induced electromagnetic forces during a plasma disruption (rapid plasma current quench) could give damages on the piping system. Ceramic brazing joint is a conventional method for the electric circuit break, but an application to the fusion reactor is not feasible due to its brittleness. Here, a stainless steel/ceramics/stainless steel functionally gradient material (FGM) has been proposed and developed as an integrated insulation joint of the piping system. Both sides of the joint can be welded to the main pipes, and expected to be reliable even in the fusion reactor environment. When the FGM joint is manufactured by way of a sintering process, a residual thermal stress is the key issue. Through detailed computations of the residual thermal stress and several trial productions, tubular elements of FGM joints have been successfully manufactured. (author)

  19. Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

    International Nuclear Information System (INIS)

    Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

    1985-01-01

    Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

  20. The International Reactor Dosimetry File (IRDF-85)

    International Nuclear Information System (INIS)

    Cullen, D.E.; McLaughlin, P.K.

    1985-04-01

    This document describes the contents of the second version of the International Reactor Dosimetry File (IRDF-85), distributed by the Nuclear Data Section of the International Atomic Energy Agency. This library superseded IRDF-82. (author)

  1. Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

    Science.gov (United States)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

  2. Culham Conceptual Tokamak Mark II. Design study of the layout of a twin-reactor fusion power station

    International Nuclear Information System (INIS)

    Guthrie, J.A.S.; Harding, N.H.

    1981-07-01

    This report describes the building layout and outline design for the nuclear complex of a fusion reactor power station incorporating two Culham Conceptual Tokamak Reactors Mk.II. The design incorporates equipment for steam generation, process services for the fusion reactors and all facilities for routine and non-routine servicing of the nuclear complex. The design includes provision of temporary facilities for on site construction of the major reactor components and shows that these facilities may be used for disassembly of the reactors either for major repair and/or decommissioning. Preliminary estimates are included, which indicate the cost benefits to be obtained from incorporating two reactors in one nuclear complex and from increased wall loading. (author)

  3. Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design

    International Nuclear Information System (INIS)

    Qiu Qing-Lai; Xiao Bing-Jia; Guo Yong; Liu Lei; Wang Yue-Hang

    2017-01-01

    Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other. (paper)

  4. Treating the neutronics of a (d,t)-fusion reactor operating on the tokamak principle (NET)

    International Nuclear Information System (INIS)

    Fischer, U.

    1990-10-01

    The primary aim of this work is to check the simplified one-dimensional neutronic approach being used frequently in design calculations. Therefore the neutronics of the NET (Next European Torus)-reactor is treated in a three-dimensional torus sector model by means of Monte Carlo calculations with the MCNP-code. Various blanket variants with different neutronic characteristics are taken into account. All of them had been developed for use in the NET-reactor: a ceramic solid breeder blanket using beryllium as neutron multiplier, a self-cooled liquid metal blanket using the eutectic alloy Pb-17Li or, alternatively, pure lithium as breeding material/coolant, and an aqueous lithium salt solution blanket. It is shown, that the one-dimensional approach can be applied in design calculations for evaluating power density distributions, if the plasma source is normalized in a consistent manner, if its spatial distribution is choosen appropriately, and, furthermore, if its angular dependence is taken into account. A three-dimensional treatment of the actual tokamak geometry, however, is necessary for determining the breeding performance of the blanket variants and for performing shielding calculations of the whole system blanket/shield. (orig./HP) [de

  5. Design study of superconducting toroidal magnet for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-10-01

    Design study of the superconducting toroidal magnet for the Tokamak experimental fusion reactor has been carried out. Performed here were conductor design, magnetic field calculation, design of coil support, stress analysis, design of refrigeration system and safety analysis. The maximum toroidal field at the coil is 12.0 T, providing 6.0 T at the plasma center. The coil bore is 7.3 x 11.2 m, and the coil shape is deformed constant-tension D-shape. 16 coil design was chosen, considering the necessity for good access for the evacuation port or the support structure of blanket and shield. The maximum field ripple obtained in the plasma region is 0.47%. The operational current is 25,100 amperes, and the cryogenic stabilization is fulfilled. Nb 3 Sn superconductor was chosen in spite of its brittleness, to attain a toroidal field of 12 T which is higher than a practical limit for NbTi superconductor. The development of large Nb 3 Sn coil technology would be necessary to realize an economic fusion power reactor. (auth.)

  6. Calculations of alpha particle loss for reversed magnetic shear in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Redi, M.H.; White, R.B.; Batha, S.H.; Levinton, F.M.; McCune, D.C.

    1997-03-01

    Hamiltonian coordinate, guiding center code calculations of the toroidal field ripple loss of alpha particles from a reversed shear plasma predict both total alpha losses and ripple diffusion losses to be greater than those from a comparable non-reversed magnetic shear plasma in the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. High central q is found to increase alpha ripple losses as well as first orbit losses of alphas in the reversed shear simulations. A simple ripple loss model, benchmarked against the guiding center code, is found to work satisfactorily in transport analysis modelling of reversed and monotonic shear scenarios. Alpha ripple transport on TFTR affects ions within r/a=0.5, not at the plasma edge. The entire plasma is above threshold for stochastic ripple loss of alpha particles at birth energy in the reversed shear case simulated, so that all trapped 3.5 MeV alphas are lost stochastically or through prompt losses. The 40% alpha particle loss predictions for TFTR suggest that reduction of toroidal field ripple will be a critical issue in the design of a reversed shear fusion reactor.

  7. MIRI: A multichannel far-infrared laser interferometer for electron density measurements on TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Park, H.K.; Johnson, L.C.; Anderson, H.M.; Chouinard, R.; Foote, V.S.; Ma, C.H.; Clifton, B.J.

    1987-07-01

    A ten-channel far-infrared laser interferometer which is routinely used to measure the spatial and temporal behavior of the electron density profile on the TFTR tokamak is described and representative results are presented. This system has been designed for remote operation in the very hostile environment of a fusion reactor. The possible expansion of the system to include polarimetric measurements is briefly outlined. 13 refs., 8 figs

  8. Reliability test for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Uchiyama, Junichi

    1998-01-01

    41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

  9. Cryogenic aspects of a demountable toroidal field magnet system for tokamak type fusion reactors

    International Nuclear Information System (INIS)

    Hsieh, S.Y.; Powell, J.; Lehner, J.

    1977-01-01

    A new concept for superconducting Toroidal Field (TF) magnet construction is presented. It is termed the ''Demountable Externally Anchored Low Stress'' (DEALS) magnet system. In contrast to continuous wound conventional superconducting coils, each magnet coil is made from several straight coil segments to form a polygon which can be joined and disjoined to improve reactor maintenance accessibility or to replace failed coil segments if necessary. A design example is presented of a DEALS magnet system for a UWMAK II size reactor. The overall magnet system is described, followed by a detailed analysis of the major heat loads in order to assess the refrigeration requirements for the concept. Despite the increased heat loads caused by high current power leads (200,000 amps) and the coil warm reinforcement support system, the analysis shows that at most, only about one percent (approximately 20 Mw) of the plant electrical output (approximately 2,000 Mw) is needed to operate the magnet cryogenic system. The advantages and the drawbacks of the DEALS magnet system are also discussed. The advantages include: capability to replace failed coils, increased accessibility to the blanket shield assembly, reduced reliability requirements for the magnet, much lower stress in conductor, easier application of improved high field brittle superconductors like Nb 3 Sn, improved magnet safety features, etc. The drawbacks are the increased refrigeration requirements and the necessity of a movable coil support system. A comparison with a conventional magnet system is made. It is concluded that the benefits of the DEALS approach far outweigh its penalties, and that the DEALS concept is the most practical, economical way to construct TF magnet systems for Tokamak reactors

  10. Reactor Structure Materials: Corrosion of Reactor Core Internals

    International Nuclear Information System (INIS)

    Van Dyck, S.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

  11. Structures and Materials of Reactor Internals for PWR in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Kim, W. S.; Kwon, S. C.; Kwon, J. H.; Kim, Y. S.; Kim, H. P.; Yoo, C. S.; Lee, S. R.; Jung, M. K.; Hwang, S. S

    2007-10-15

    Nuclear reactor types in Korea are PWR type reactor (Westinghouse, Combustion Engineering, Farmatome type) and CANDU type reactor. Structures and Materials for reactor internal of PWR type were investigated. Reactor internal was composed of lower core support structure, upper core support assembly, incore instrumentation support structure. Lower core support structure of these structures is the most important. The major material for the reactor internal is type 304 and 316 stainless steel and radial support clevis bolts are made of Inconel. The main damage mechanism for reactor internal was IASCC and the effect of IASCC on reactor internal was investigated. The accident for reactor internal was also investigate.

  12. Reaction-rate coefficients, high-energy ions slowing-down, and power balance in a tokamak fusion reactor plasma

    International Nuclear Information System (INIS)

    Tone, Tatsuzo

    1978-07-01

    Described are the reactivity coefficient of D-T fusion reaction, slowing-down processes of deuterons injected with high energy and 3.52 MeV alpha particles generated in D-T reaction, and the power balance in a Tokamak reactor plasma. Most of the results were obtained in the first preliminary design of JAERI Experimental Fusion Reactor (JXFR) driven with stationary neutral beam injection. A manual of numerical computation program ''BALTOK'' developed for the calculations is given in the appendix. (auth.)

  13. Soft x-ray camera for internal shape and current density measurements on a noncircular tokamak

    International Nuclear Information System (INIS)

    Fonck, R.J.; Jaehnig, K.P.; Powell, E.T.; Reusch, M.; Roney, P.; Simon, M.P.

    1988-05-01

    Soft x-ray measurements of the internal plasma flux surface shaped in principle allow a determination of the plasma current density distribution, and provide a necessary monitor of the degree of internal elongation of tokamak plasmas with a noncircular cross section. A two-dimensional, tangentially viewing, soft x-ray pinhole camera has been fabricated to provide internal shape measurements on the PBX-M tokamak. It consists of a scintillator at the focal plane of a foil-filtered pinhole camera, which is, in turn, fiber optically coupled to an intensified framing video camera (/DELTA/t />=/ 3 msec). Automated data acquisition is performed on a stand-alone image-processing system, and data archiving and retrieval takes place on an optical disk video recorder. The entire diagnostic is controlled via a PDP-11/73 microcomputer. The derivation of the polodial emission distribution from the measured image is done by fitting to model profiles. 10 refs., 4 figs

  14. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  15. Fusion materials studies and technological experiments in test thermonuclear reactor-tokamak

    International Nuclear Information System (INIS)

    Altovskij, I.V.; Gornostaev, B.D.; Blinov, I.A.; Gur'ev, V.V.; Zujkov, G.A.; Osadchij, N.A.; Tochenyj, L.V.

    1987-01-01

    Extensive studies of materials (strength, radiation resistance, reliability and safety) should be made to substantiate designs during the initial stage of R and D and the choice of the final version for the design of the blanket and energy conversion system. It is evident that those experimental devices which are available or planned are not capable of simulating effects on the OTR blanket units (high neutron flux, strong electromagnetic fields, stresses due to coolant pressure at operating temperature and own weight, plasma effects, etc.). Therefore, in OTR design, constructive decisions should be taken on the use of the tokamak OTR as a unique experimental device for real full-scale tests of construction, uranium, lithium and protective materials (antidiffusion coatings for radiation safety, etc.) intended for long term reliable and safe operation in future commercial fusion reactors. In the paper, consideration is given to some experimental devices intended for the study of the physical and mechanical properties of materials under loads typical of blanket operation; research into the mutual compatibility of construction materials, with coolant and operating materials; the study of the efficiency and reliability of material coatings; and the determination of optimal operating temperature ranges for materials, etc. (author). 2 refs, 5 figs

  16. Measurements of energy and flux of neutrals at the wall in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Wampler, W.R.; Kilpatrick, S.J.

    1989-01-01

    The energy and flux of charge-exchange neutral deuterium incident on the wall in the Tokamak Fusion Test Reactor (TFTR) were measured using a carbon resistance probe. This device utilizes the increase in electrical resistance of a thin carbon film produced by the impinging particles. Particle energies and fluxes were determined from the stopping effect of thin layers of silicon over the carbon. Film thicknesses were chosen to give a useful energy range from about 50 to 5000 eV. Time-resolved data were obtained for the first time by using a transient digitizer to record the signals every millisecond during the discharge. The deduced flux and Maxwellian temperature were 0.7 x 10 14 /cm 2 s and kT = 350 eV, respectively, during Ohmic heating. Neutral beam heating (10 MW) increased the temperature by a factor of 4 and increased the flux by a factor of 2. The observed flux of neutrals to the wall is large enough for changes in the recycling of these particles to affect plasma fueling significantly

  17. Cryosorption of helium on argon frost in Tokamak Fusion Test Reactor neutral beamlines

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Cropper, M.B.; Dylla, H.F.; Garzotto, V.; Dudek, L.E.; Grisham, L.R.; Martin, G.D.; O'Connor, T.E.; Stevenson, T.N.; von Halle, A.; Williams, M.D.; Kim, J.

    1990-01-01

    Helium pumping on argon frost has been investigated on Tokamak Fusion Test Reactor (TFTR) neutral beam injectors and shown to be viable for limited helium beam operation. Maximum pumping speeds are ∼25% less than those measured for pumping of deuterium. Helium pumping efficiency is low, >20 argon atoms are required to pump each helium atom. Adsorption isotherms are exponential and exhibit a twofold increase in adsorption capacity as the cryopanel temperature is reduced from 4.3 K to 3.7 K. Pumping speed was found to be independent of cryopanel temperature over the temperature range studied. After pumping a total of 2000 Torr l of helium, the beamline base pressure rose to 2x10 -5 Torr from an initial value of 10 -8 Torr. Accompanying this three order of magnitude increase in pressure was a modest 40% decrease in pumping speed. The introduction of 168 Torr l of deuterium prior to helium injection reduced the pumping speed by a factor of two with no decrease in adsorption capacity

  18. Preliminary conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR). Status report, January 1978--March 1978

    International Nuclear Information System (INIS)

    Kelly, J.L.

    1978-03-01

    The DTHR preliminary conceptual design consists of a magnetically confined fusion reactor fitted with a fertile thorium blanket. The fusion driver concept is based on a beam driven plasma, but at sufficiently high plasma densities that neutrons originating from the interactions of bulk plasma ions contribute significantly to the wall loading. The tokamak has a major radius of 5.2 m, a minor radius of 1.2 m, and the elongation is 1.6. All of the magnetic coil systems are superconducting Nb 3 Sn based on the Large Coil Project (LCP) technology. The toroidal field (TF) coils employ an innovative concept, the ''compact D'' configuration. An engineered bundle divertor concept has been developed based on the bundle divertor design techniques developed for TNS and ISX-B. A thermal power of 150MW of 200 keV deuterium is injected into the plasma through six ducts of a positive ion, neutral beam injection system (NBIS). A water cooled, 316 stainless steel vacuum vessel concept was developed and initial scoping analyses look encouraging. The fusile fuel handling system was evaluated and defined. Details of the tritium injection system remain to be developed. Tritium breeding will be assessed in subsequent phases of the DTHR operation. The fusion driver provides a neutron first wall loading of 2MW/m 2 for fissile production in the blanket

  19. A Michelson interferometer/polarimeter on the Tokamak Fusion Test Reactor (TFTR)

    International Nuclear Information System (INIS)

    Park, H.K.; Mansfield, D.K.; Johnson, L.C.; Ma, C.H.

    1987-01-01

    A multichannel interferometer/polarimeter for the Tokamak Fusion Test Reactor (TFTR) has been developed in order to study the time dependent plasma current density (J/sub p/) and electron density (n/sub e/) profile simultaneously. The goal of the TFTR is demonstration of breakeven via dueuterium and tritium (DT) plasma. In order to be operated and maintained during DT operation phase, the system is designed based on the Michelson geometry which possesses intrinsic standing wave problems. So far, there has been no observable signals due to these standing waves. However, a standing wave resulted from the beam path design to achieve a optimum use of the laser power was found. This standing wave has not prevented initial 10 channel interferometer operation. However, a single channel polarimeter test indicated this standing wave was fatal for Faraday notation measurements. Techniques employing 1/2 wave plates and polarizers have been applied to eliminate this standing wave problem. The completion of 10 channel Faraday rotation measurements may be feasible in the near future

  20. Design study of superconducting magnets for tokamak experimental fusion reactor, (1)

    International Nuclear Information System (INIS)

    1980-03-01

    Design study has been made of superconducting magnets for a Tokamak experimental fusion reactor: toroidal field magnet design, poloidal field magnet design, refrigeration system design, magnet safety analysis, and magnet assembling and disassembling system design. A maximum toroidal field in the coil is 11.0 T, providing 5.5 T at plasma center. Nb 3 Sn superconducting cable is used to attain the toroidal field of 11 T. The coil bore is 7.3 x 11.2 m, and the coil shape is deformed constant-tension D-shape. The magnetomotive force is 185.6 MAT, and the operational current is 25.9 kA. In poloidal field magnet design, the coil is pancake-wound Nb 3 Sn conductor. The conductor is enclosed in Ti-alloy sheath, which serves also as helium containment vessel. The conductor is cooled by forced flow supercritical helium of 7 atm and 4.6 K, and the operational current is 25 -- 27 kA. (author)

  1. Tritium production in He-3 gas cells immersed in the Tokamak Fusion Test Reactor neutron field

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.; Gentile, C.A.; Ascione, G.; Kugel, H.W.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Kumar, A. [University of California, Los Angeles, California 90024 (United States)

    1999-01-01

    Tritium generated in an external cell by the reaction {sup 3}He(n,p)T can be used as a gauge of long-term fusion neutron production, because of the 12-year half-life of T and the relative ease of measuring the T content either by sampling or from the saturation current of the cell when operated as an ionization chamber. Two high-pressure {sup 3}He gas cells enclosed in polyethylene neutron moderators were exposed to Tokamak Fusion Test Reactor (TFTR) neutrons during high-power D{endash}T operation. The tritium produced in the cells was assayed by the Princeton Differential Atmospheric Tritium Sampler. The measured tritium generated per 10{sup 19} fusion neutrons was 510 pCi/cc at 2.3 m from the TFTR vessel and 1.3 m below the midplane, and 2020 pCi/cc at 1.0 m from the TFTR vessel in the midplane. Combining these results with previous measurements at a third location, we found 0.11 to 0.23 triton produced per neutron incident on the projected cell cross section, with an asymptotic local tritium breeding ratio of 0.32. {copyright} {ital 1999 American Institute of Physics.}

  2. Preparations for deuterium tritium experiments on the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G. [and others

    1994-04-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR). These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinet{sup {trademark}} system, modification of the vacuum system to handle tritium, preparation and testing of the neutral beam system for tritium operation and a final deuterium-deuterium (D-D) run to simulate expected deuterium-tritium (D-T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D-T experiments using D-D have been performed. The physics objectives of D-T operation are production of {approximately} 10 megawatts (MW) of fusion power, evaluation of confinement and heating in deuterium-tritium plasmas, evaluation of {alpha}-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined {alpha}-particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed.

  3. Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Ja

    1994-05-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. [bold 21], 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert[sup TM] system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of [approx]10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of [alpha]-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined [alpha] particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed.

  4. Measurements of tokamak fusion test reactor D-T radiation shielding efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Ascione, G. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Elwood, S. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Gilbert, J. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Ku, L.P. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Levine, J. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Rule, K. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Azziz, N. [USDOE Environmental Measurements Lab., New York (United States); Goldhagen, P. [USDOE Environmental Measurements Lab., New York (United States); Hajnal, F. [USDOE Environmental Measurements Lab., New York (United States); Shebell, P. [USDOE Environmental Measurements Lab., New York (United States)

    1995-03-01

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

  5. Design of deuterium and tritium pellet injector systems for Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Wysor, R.B.; Baylor, L.R.; Bryan, W.E.

    1985-01-01

    Three pellet injector designs developed by the Oak Ridge National Laboratory (ORNL) are planned for the Tokamak Fusion Test Reactor (TFTR) to reach the goal of a tritium pellet injector by 1988. These are the Repeating Pneumatic Injector (RPI), the Deuterium Pellet Injector (DPI) and the Tritium Pellet Injector (TPI). Each of the pellet injector designs have similar performance characteristics in that they deliver up to 4-mm-dia pellets at velocities up to 1500 m/s with a dsign goal to 2000 m/s. Similar techniques are utilized to freeze and extrude the pellet material. The injector systems incorporate three gun concepts which differ in the number of gun barrels and the method of forming and chambering the pellets. The RPI, a single barrel repeating design, has been operational on TFTR since April 1985. Fabrication and assembly are essentially complete for DPI, and TPI is presently on hold after completing about 80% of the design. The TFTR pellet injector program is described, and each of the injector systems is described briefly. Design details are discussed in other papers at this symposium

  6. Preliminary conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR). Status report, January 1978--March 1978

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.L. (ed.)

    1978-03-01

    The DTHR preliminary conceptual design consists of a magnetically confined fusion reactor fitted with a fertile thorium blanket. The fusion driver concept is based on a beam driven plasma, but at sufficiently high plasma densities that neutrons originating from the interactions of bulk plasma ions contribute significantly to the wall loading. The tokamak has a major radius of 5.2 m, a minor radius of 1.2 m, and the elongation is 1.6. All of the magnetic coil systems are superconducting Nb/sub 3/Sn based on the Large Coil Project (LCP) technology. The toroidal field (TF) coils employ an innovative concept, the ''compact D'' configuration. An engineered bundle divertor concept has been developed based on the bundle divertor design techniques developed for TNS and ISX-B. A thermal power of 150MW of 200 keV deuterium is injected into the plasma through six ducts of a positive ion, neutral beam injection system (NBIS). A water cooled, 316 stainless steel vacuum vessel concept was developed and initial scoping analyses look encouraging. The fusile fuel handling system was evaluated and defined. Details of the tritium injection system remain to be developed. Tritium breeding will be assessed in subsequent phases of the DTHR operation. The fusion driver provides a neutron first wall loading of 2MW/m/sup 2/ for fissile production in the blanket.

  7. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  8. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

    2000-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  9. Reliability tests for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

    1998-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  10. Reactor similarity for plasma–material interactions in scaled-down tokamaks as the basis for the Vulcan conceptual design

    International Nuclear Information System (INIS)

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

    2012-01-01

    Highlights: ► Discussion of similarity scalings for reduced-size tokamaks. ► Proposal of a new set of scaling laws for divertor similarity. ► Discussion of how the new scaling provides fidelity to a reactor. ► The new scaling is used as the basis for the Vulcan conceptual design. - Abstract: Dimensionless parameter scaling techniques are a powerful tool in the study of complex physical systems, especially in tokamak fusion experiments where the cost of full-size devices is high. It is proposed that dimensionless similarity be used to study in a small-scale device the coupled issues of the scrape-off layer (SOL) plasma, plasma–material interactions (PMI), and the plasma-facing material (PFM) response expected in a tokamak fusion reactor. Complete similarity is not possible in a reduced-size device. In addition, “hard” technological limits on the achievable magnetic field and peak heat flux, as well as the necessity to produce non-inductive scenarios, must be taken into account. A practical approach is advocated, in which the most important dimensionless parameters are matched to a reactor in the reduced-size device, while relaxing those parameters which are far from a threshold in behavior. “Hard” technological limits are avoided, so that the reduced-size device is technologically feasible. A criticism on these grounds is offered of the “P/R” model, in which the ratio of power crossing the last closed flux surface (LCFS), P, to the device major radius, R, is held constant. A new set of scaling rules, referred to as the “P/S” scaling (where S is the LCFS area) or the “PMI” scaling, is proposed: (i) non-inductive, steady-state operation; (ii) P is scaled with R 2 so that LCFS areal power flux P/S is constant; (iii) magnetic field B constant; (iv) geometry (elongation, safety factor q * , etc.) constant; (v) volume-averaged core density scaled as n≈n ¯ e ∼R −2/7 ; and (vi) ambient wall material temperature T W,0 constant. It is

  11. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  12. End points in discharge cleaning on TFTR (Tokamak Fusion Test Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, D.; Dylla, H.F.; Bell, M.G.; Blanchard, W.R.; Bush, C.E.; Gettelfinger, G.; Hawryluk, R.J.; Hill, K.W.; Janos, A.C.; Jobes, F.C.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) /approx/ 3 at /approx/ 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H/sub 2/O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after /approx/ 10/sup 5/ pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150/degree/C to about 250/degree/C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H/sub 2/O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab.

  13. End points in discharge cleaning on TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Mueller, D.; Dylla, H.F.; Bell, M.G.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) approx 3 at approx 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H 2 O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after approx 10 5 pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150 degree C to about 250 degree C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H 2 O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab

  14. Power conversion and balance of plant considerations for the STARFIRE commercial tokamak reactor

    International Nuclear Information System (INIS)

    Barry, K.; Graumann, D.

    1981-01-01

    The power conversion and balance of plant facilities for this tenth-of-a-kind tokamak fusion power plant are a combination of both features common to any large power plant, and elements peculiar to the fusion technology. For example, the steam generators, turbine-generator and main condenser components of the power conversion system and the natural draft cooling towers that are used for heat rejection at sites not close to a large body of water are generic to power plants. The tritium reprocessing facilities that minimize the tritium inventory in the plant, the Electrical and RF Power Supply Building that contains the coil and rf power supplies, the cryogenic facilities that provide liquid helium coolant for the superconducting coils, and the Hot Cell in which fully remote repair and maintenance functions are performed are unique to a fusion power plant. One of the major features of the STARFIRE design is steady state operation that maximizes overall facility reliability and eliminates both thermal storage requirements and potential power fluctuations on the grid. The reference reactor power is 4000 MWt with a gross electric power generation of 1440 MW. For STARFIRE, water is the preferred coolant and is utilized in both the first wall/blanket and limiter cooling circuits. Dual parallel primary coolant loops cool the twenty-four first-wall/blanket sectors. The power deposited in the limiter, approximately 5% of the total thermal power, is removed by the separate limiter/feedwater loop and is used for feedwater heating in the steam power conversion system

  15. Rail for inspection/maintenance device in TOKAMAK type container of a thermonuclear reactor

    International Nuclear Information System (INIS)

    Takahashi, Kenji.

    1996-01-01

    A circular rail divided into four arcuate parts for an inspection/maintenance device which runs in a TOKAMAK type container is disposed. Each of the divided rails is supported at the center of the outer surface rotatably by extendable rail supporting shafts. Each of the divided rail is constituted such that it can be contained between limiters disposed at the outer side of the TOKAMAK container when each of the rail support shafts is contracted. With such a constitution, each of the rail support shafts and arcuate rail is contracted and rotated from the outside of the TOKAMAK type container by an actuator. In order to form a circular rail, each of the rail support shafts is extended toward the center of the TOKAMAK type container, and then each of the arcuate rails is rotated into a horizontal state. Then, the joint portions of each of the arcuate rails are connected by using remote controllable locking rods. (I.S.)

  16. Passive vertical stability of ITER (International Thermonuclear Experimental Reactor)

    International Nuclear Information System (INIS)

    Leuer, J.A.; Wesley, J.C.

    1989-01-01

    Next generation tokamaks, such as the International Thermonuclear Experimental Reactor (ITER), are being designed with high plasma current carrying capability. High current is best achieved by increasing the plasma elongation; however, this leads to a plasma configuration which is unstable to axisymmetric vertical motion of the plasma. Passive conductors must be included in the design to stabilize this vertical motion. In this paper, the authors analyze the passive stability property of the Physics Phase (PP) operation of the ITER device. A linearized, rigid-body assumption is used to predict the stability parameter of the proposed passive stabilization structure. The problem is formulated into an eigenvalue problem and the solution predicts the critical stability margin and characteristic growth rate. The proposed stabilization structure and vacuum vessel (VV) provides 50% margin against Alfven time scale vertical motion. The growth rate is less than 50 s -1 for the nominal plasma. The plasma is shown to have low growth rates over a range of plasma internal inductance and poloidal beta. 3 refs., 5 figs., 5 tabs

  17. Creating poloidal flux in a tokamak plasma with low frequency waves

    International Nuclear Information System (INIS)

    Kirkwood, R.K.; Capewell, D.L.; Bellan, P.M.

    1993-01-01

    Using a fully toroidal, collisionless, low frequency model, we show that low amplitude, circularly polarized waves can, depending on antenna geometry (i) drive the toroidal EMF necessary to sustain a tokamak reactor, or (ii) shift the internal current profile. Measurements on a small tokamak to test (ii) agree with the model predictions. (orig.)

  18. Collaboration on Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Text Reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Intrator, T.

    2000-06-01

    This proposal was peer reviewed and funded as a Collaboration on ''Low Phase Speed Radio Frequency Current Drive Experiments at the Tokamak Fusion Test Reactor''. The original plans we had were to carry out the collaboration proposal by including a post doctoral scientist stationed at PPPL. In response to a 60+% funding cut, all expenses were radically pruned. The post doctoral position was eliminated, and the Principal Investigator (T. Intrator) carried out the brunt of the collaboration. Visits to TFTR enabled T. Intrator to set up access to the TFTR computing network, database, and get familiar with the new antennas that were being installed in TFTR during an up to air. One unfortunate result of the budget squeeze that TFTR felt for its last year of operation was that the experiments that we specifically got funded to perform were not granted run time on TFTR., On the other hand we carried out some modeling of the electric field structure around the four strap direct launch Ion Bernstein Wave (IBW) antenna that was operated on TFTR. This turned out to be a useful exercise and shed some light on the operational characteristics of the IBW antenna and its coupling to the plasma. Because of this turn of events, the project was renamed ''Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Test Reactor''.

  19. Collaboration on Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Text Reactor. Final report

    International Nuclear Information System (INIS)

    Intrator, T.

    2000-01-01

    This proposal was peer reviewed and funded as a Collaboration on ''Low Phase Speed Radio Frequency Current Drive Experiments at the Tokamak Fusion Test Reactor''. The original plans we had were to carry out the collaboration proposal by including a post doctoral scientist stationed at PPPL. In response to a 60+% funding cut, all expenses were radically pruned. The post doctoral position was eliminated, and the Principal Investigator (T. Intrator) carried out the brunt of the collaboration. Visits to TFTR enabled T. Intrator to set up access to the TFTR computing network, database, and get familiar with the new antennas that were being installed in TFTR during an up to air. One unfortunate result of the budget squeeze that TFTR felt for its last year of operation was that the experiments that we specifically got funded to perform were not granted run time on TFTR., On the other hand we carried out some modeling of the electric field structure around the four strap direct launch Ion Bernstein Wave (IBW) antenna that was operated on TFTR. This turned out to be a useful exercise and shed some light on the operational characteristics of the IBW antenna and its coupling to the plasma. Because of this turn of events, the project was renamed ''Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Test Reactor''

  20. Resistive internal kink modes in a tokamak with high-pressure plasma

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Mikhajlovskij, A.B.; Tatarinov, E.G.

    1988-01-01

    Theory of resistive internal kink modes in a tokamak with high-pressure plasma is developed. Equation for Fourie-image of disturbed displacment in a resistive layer ie derived with regard to effects of the fourth order by plasma pressure within the framework of single-liquid approach. In its structure this equation coincides with a similar equation for resistive balloon modes and has an exact solution expressed by degenerated hypergeometric function. A general dispersion equation for resistive kink modes is derived with regard to the effects indicated. It is shown that plasma pressure finiteness leads to the reduction of reconnection and tyring-mode increments

  1. High #betta# and toroidal effects on the internal kink mode in tokamaks

    International Nuclear Information System (INIS)

    Schmalz, R.

    1982-09-01

    The inclusion of high-#betta# and first-order toroidal terms in the reduced set of (resistive) MHD equations affords the possibility of improving the study of tokamak plasma behaviour by three-dimensional numerical simulation. A new code, GALA, based on the reduced equations is developed. It is used to analyse the linear and nonlinear behaviour of the internal kink mode in equilibria which are generated by a simple relaxation procedure. We find that the inclusion of toroidal effects in high-#betta# equilibria provides considerable stabilization. (orig.)

  2. The contribution to the energy balance and transport in an advanced-fuel tokamak reactor

    International Nuclear Information System (INIS)

    Atzeni, S.; Vlad, G.

    1985-01-01

    The influence of synchrotron radiation emission on the energy balance of an advanced-fuel (such as D- 3 He, or catalyzed-D) tokamak plasma is considered. It is shown that a region in the β-T space exists, where the fusion energy delivered to the plasma overcomes synchrotron and bremsstrahlung energy losses, and which could then allow for ignited operation. 1-Dimensional codes results are also presented, which illustrate the main features of radial transport in a ignited, D- 3 He tokamak plasma

  3. Reactor applications of the compact fusion advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.; Logan, B.G.; Campbell, R.B. (Lawrence Livermore National Lab., CA (USA))

    1989-05-01

    The preliminary design of a high temperature blanket and compact power conversion system for a D-T reactor utilizing a mercury Rankine cycle and nonequilibrium MHD disk generators is described. The CFAR cycle is described in detail and a blanket geometry is proposed which is designed to give good heat transfer and low pressure drops for the two-phase boiling mercury flow in the presence of strong magnetic fields. The blanket utilizes beryllium and flibe in the low-temperature breeding zone and lithium aluminate in beryllium oxide pebbles in the high-temperature breeding zone to obtain a tritium breeding ratio of 1.07. With careful choice of materials and geometry, it has been possibly to have 46% of the neutron and gamma heat deposited in the high-temperature pebble beds at the back of the blanket. The design utilizes the many advantages of liquid-metal Rankine cycles, including low pressures, mass flow rates and pumping power, and superheating along with internal regeneration to achieve good thermal efficiencies of about 35% using the microwave superheater and about 32% without the microwave superheater. The capital costs are expected to be much lower than past design concepts because the MHD power generators and the mercury condensers and heat exchangers can all be placed inside the reactor vault; preliminary cost estimates indicate a very attractive cost of electricity for this advanced concept. (orig.).

  4. A step towards controlled fusion reactors: Tore Supra tokamak with superconducting magnets

    International Nuclear Information System (INIS)

    Turck, B.

    1988-01-01

    Tore Supra technology has to solve all the problems related to the development and the installaion of superconducting coils and associated cryogenic devices. Tore Supra will allow to get a significative experience to prepare next machines. Specifications and needs of tokamaks concerning the superconducting coils of future machines are recalled [fr

  5. Mode Conversion of High-Field-Side-Launched Fast Waves at the Second Harmonic of Minority Hydrogen in Advanced Tokamak Reactors

    International Nuclear Information System (INIS)

    Sund, R.; Scharer, J.

    2003-01-01

    Under advanced tokamak reactor conditions, the Ion-Bernstein wave (IBW) can be generated by mode conversion of a fast magnetosonic wave incident from the high-field side on the second harmonic resonance of a minority hydrogen component, with near 100% efficiency. IBWs have the recognized capacity to create internal transport barriers through sheared plasma flows resulting from ion absorption. The relatively high frequency (around 200 MHz) minimizes parasitic electron absorption and permits the converted IBW to approach the 5th tritium harmonic. It also facilitates compact antennas and feeds, and efficient fast wave launch. The scheme is applicable to reactors with aspect ratios < 3 such that the conversion and absorption layers are both on the high field side of the magnetic axis. Large machine size and adequate separation of the mode conversion layer from the magnetic axis minimize poloidal field effects in the conversion zone and permit a 1-D full-wave analysis. 2-D ray tracing of the IBW indicates a slightly bean-shaped equilibrium allows access to the tritium resonance

  6. Experience with mechanical segmentation of reactor internals

    International Nuclear Information System (INIS)

    Carlson, R.; Hedin, G.

    2003-01-01

    Operating experience from BWE:s world-wide has shown that many plants experience initial cracking of the reactor internals after approximately 20 to 25 years of service life. This ''mid-life crisis'', considering a plant design life of 40 years, is now being addressed by many utilities. Successful resolution of these issues should give many more years of trouble-free operation. Replacement of reactor internals could be, in many cases, the most favourable option to achieve this. The proactive strategy of many utilities to replace internals in a planned way is a market-driven effort to minimize the overall costs for power generation, including time spent for handling contingencies and unplanned outages. Based on technical analyses, knowledge about component market prices and in-house costs, a cost-effective, optimized strategy for inspection, mitigation and replacements can be implemented. Also decommissioning of nuclear plants has become a reality for many utilities as numerous plants worldwide are closed due to age and/or other reasons. These facts address a need for safe, fast and cost-effective methods for segmentation of internals. Westinghouse has over the last years developed methods for segmentation of internals and has also carried out successful segmentation projects. Our experience from the segmentation business for Nordic BWR:s is that the most important parameters to consider when choosing a method and equipment for a segmentation project are: - Safety, - Cost-effectiveness, - Cleanliness, - Reliability. (orig.)

  7. Ideal MHD stability of internal kinks in circular and shaped tokamaks

    International Nuclear Information System (INIS)

    Luetjens, H.; Bondeson, A.; Vlad, G.

    1992-04-01

    Stability limits for the internal kink mode in tokamaks are calculated for different current profiles and plasma cross sections using ideal magnetohydrodynamics (MHD). The maximum stable poloidal beta at the q = 1 surface (β p ) is sensitive to the current profile, but for circular cross sections, it is typically between 0.1 and 0.2. Large aspect ratio theory gives similar predictions when the appropriate boundary conditions are applied at the plasma-vacuum surface. The pressure driven internal kink is significantly destabilized by ellipticity. For JET geometry, the β p -limit is typically between 0.05 and 0.1, but arbitrarily low limits can result if the shear is reduced at the q=1 surface. A large aspect ratio expansion of the Mercier criterion retaining the effects of ellipticity and triangularity is given to illustrate the destabilizing influence of ellipticity. (author) 17 figs., 16 refs

  8. Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

    International Nuclear Information System (INIS)

    Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

    2015-01-01

    Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor

  9. 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)

  10. Charge-exchange and fusion reaction measurements during compression experiments with neutral beam heating in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.; Chan, A.A.; England, A.C.; Hendel, H.W.; Medley, S.S.; Nieschmidt, E.; Roquemore, A.L.; Scott, S.D.

    1986-04-01

    Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge-exchange neutral particle analyzer. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and /sup 3/He ions, respectively. When the plasma was compressed, the d(d,n)/sup 3/He fusion reaction rate increased a factor of five, and the /sup 3/He(d,p)/sup 4/He rate by a factor of twenty. These data were simulated with a bounce-averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling.

  11. Charge-exchange and fusion reaction measurements during compression experiments with neutral beam heating in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.

    1986-04-01

    Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge-exchange neutral particle analyzer. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and 3 He ions, respectively. When the plasma was compressed, the d(d,n) 3 He fusion reaction rate increased a factor of five, and the 3 He(d,p) 4 He rate by a factor of twenty. These data were simulated with a bounce-averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling

  12. Development of Tokamak reactor system code and conceptual studies of DEMO with He Cooled Molten Li blanket

    International Nuclear Information System (INIS)

    Hong, B.G.; Lee, Dong Won; Kim, Yong Hi

    2007-01-01

    To develop the concepts of fusion power plants and identify the design parameters, we have been developing the tokamak reactor system code. The system code can take into account a wide range of plasma physics and technology effects simultaneously and it can be used to find design parameters which optimize the given figure of merits. The outcome of the system studies using the system code is to identify which areas of plasma physics and technologies and to what extent should be developed for realization of a given fusion power plant concepts. As an application of the tokamak reactor system code, we investigate the performance of DEMO for early realization with a limited extension from the plasma physics and technology used in the design of the ITER. Main requirements for DEMO are selected as: 1) to demonstrate tritium self-sufficiency, 2) to generate net electricity, and 3) for steady-state operation. The size of plasma is assumed to be same as that of ITER and the plasma parameters which characterize the performance, i.e. normalized β value, β N , confinement improvement factor for the H-mode, H and the ratio of the Greenwald density limit n/n G are assumed to be improved beyond those of ITER: β N >2.0, H>1.0 and n/n G >1.0. Tritium self-sufficiency is provided by the He Cooled Molten Lithium (HCML) blanket with the total thickness of 2.5 m including the shield. With n/n G >1.2, net electric power bigger than 500 MW is possible with β N >4.0 andH>1.2. To access operation space for higher electric power, main restrictions are given by the divertor heat load and the steady-state operation requirements. Developments in both plasma physics and technology are required to handle high heat load and to increase the current drive efficiency. (orig.)

  13. Simulations of alpha particle ripple loss from the International Thermonuclear Experimental Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Redi, M.H.; Budny, R.V.; McCune, D.C.; Miller, C.O.; White, R.B.

    1996-05-01

    Calculations of collisional stochastic ripple loss of alpha particles from the new 20 toroidal field (TF) coil International Thermonuclear Experimental Reactor (ITER) predict small alpha ripple losses, less than 0.4%, close to the loss calculated for the full current operation of the earlier 24 TF coil design. An analytic fit is obtained to the ITER ripple data field demonstrating the nonlinear height dependence of the ripple minimum for D shaped ripple contours. In contrast to alpha loss simulations for the Tokamak Fusion Test Reactor (TFTR), a simple Goldston, White, Boozer stochastic loss criterion ripple loss model is found to require an increased renormalization of the stochastic threshold {delta}{sub s}/{delta}{sub GWB} {ge} 1. Effects of collisions, sawtooth broadening and reversal of the grad B drift direction are included in the particle following simulations.

  14. Segmentation and packaging reactor vessels internals

    International Nuclear Information System (INIS)

    Boucau, Joseph

    2014-01-01

    Document available in abstract form only, full text follows: With more than 25 years of experience in the development of reactor vessel internals and reactor vessel segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since disposal cost is a key factor, it is important to plan and optimize waste segmentation and packaging. The choice of the optimum cutting technology is also important for a successful project implementation and depends on some specific constraints. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. The usual method is to start at the end of the process, by evaluating handling of the containers, the waste disposal requirements, what type and size of containers are available for the different disposal options, and working backwards to select a cutting method and finally the cut geometry required. The 3-D models can include intelligent data such as weight, center of gravity, curie content, etc, for each segmented piece, which is very useful when comparing various cutting, handling and packaging options. The detailed 3-D analyses and thorough characterization assessment can draw the attention to material potentially subject to clearance, either directly or after certain period of decay, to allow recycling and further disposal cost reduction. Westinghouse has developed a variety of special cutting and handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals

  15. Internal helical modes with m > 1 in a tokamak with a small shear and high plasma pressure

    International Nuclear Information System (INIS)

    Mikha lovskij, A.B.; Aburdzhaniya, G.D.; Krymskij, A.M.

    1979-01-01

    Internal helical modes with m>1 in a circular cross-section tokamak with a small shear and large value of the parameter β (β is the ratio between the mean plasma pressure and the mean pressure of the poloidal magnetic field) are investigated. The equations obtained are used to study the destabilizing effects leading to helical instabilities. The role of destabilizing effects is regarded both in local and in a nonlocal approximations on the assumption that the radial plasma pressure is distributed parabolically and that the radial current distribution is also parabolic though slightly varying. It has been established that the profiling of current may lead to the tokamak plasma stability with respect to the modes under investigation. A tokamak with a small shear has been shown to be more stable relative to these modes than that with a large shear

  16. Stability analysis of internal ideal modes in low-shear tokamaks

    International Nuclear Information System (INIS)

    Wahlberg, C.; Graves, J. P.

    2007-01-01

    The stability of internal, ideal modes in tokamaks with low magnetic shear in the plasma core is analyzed. For equilibria with large aspect ratio, a parabolic pressure profile and a flat q profile in the core, an exact solution of the ideal magnetohydrodynamic (MHD) stability equations is found. The solution includes the eigenfunctions and the complete spectra of two distinctly different MHD phenomena: A family of fast-growing, Mercier-unstable global eigenmodes localized in a low-shear region with q 1 in the core. In the latter case the solution in addition includes one unstable eigenmode, if beta is larger than a critical value depending on the width of the low-shear region and on the q-profile in the edge region

  17. Efficiency of LH current drive in tokamaks featuring an internal transport barrier

    International Nuclear Information System (INIS)

    Oliveira, C I de; Ziebell, L F; Rosa, P R da S

    2005-01-01

    In this paper, we study the effects of the occurrence of radial transport of particles in a tokamak on the efficiency of the current drive by lower hybrid (LH) waves, in the presence of an internal transport barrier. The results are obtained by numerical solution of the Fokker-Planck equation which rules the evolution of the electron distribution function. We assume that the radial transport of particles can be due to magnetic or electrostatic fluctuations. In both cases the efficiency of the current drive is shown to increase with the increase of the fluctuations that originate the transport. The dependence of the current drive efficiency on the depth and position of the barrier is also investigated

  18. Culham conceptual Tokamak reactor MkII. Conceptual layout of buildings for a twin reactor power station

    International Nuclear Information System (INIS)

    Guthrie, J.A.S.; Harding, N.H.

    1981-01-01

    This paper discusses the conceptual design of the nuclear complex of a 2400 MWe twin fusion reactor power station utilising common services and a single containment building. The design is based upon environmental and maintenance logistical requirements, the provision of adequate storage, workshop and construction facilities and the constraints imposed by the geometry of the main and auxiliary reactor coolant systems. (author)

  19. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

    Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

    1979-01-01

    This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). However, the results of this investigation are believed to be generally applicable to the broad class of the next generation of experimental tokamak facilities such as ETF. The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties are compared to the benefits and conclusions and recommendations are developed on resolving the issue

  20. Electron cyclotron heating of a tokamak reactor at down-shifted frequencies

    International Nuclear Information System (INIS)

    Fidone, I.; Giruzzi, G.; Mazzucato, E.

    1985-01-01

    The absorption of electron cyclotron waves in a hot and dense tokamak plasma is investigated for the case of the extraordinary mode for outside launching. It is shown that, for electron temperatures T/sub e/ greater than or equal to 5 keV, strong absorption occurs for oblique propagation at frequencies significantly below the electron gyrofrequency at the plasma center. A new density dependence of the wave absorption is found which is more favorable for plasma heating than the familiar n/sub e/ -1 scaling

  1. Overview of the STARFIRE reference commercial tokamak fusion power reactor design

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Barry, K.

    1980-01-01

    The purpose of the STARFIRE study is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup, superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield

  2. Cryogenic distillation: a fuel enrichment system for near-term tokamak-type D-T fusion reactors

    International Nuclear Information System (INIS)

    Misra, B.; Davis, J.F.

    1980-02-01

    The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio

  3. Method for temporary shielding of reactor vessel internals

    International Nuclear Information System (INIS)

    Grimm, N.P.; Sejvar, J.

    1991-01-01

    This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

  4. A preliminary study of a D-T tokamak fusion reactor with advanced blanket using compact fusion advanced Brayton (CFAB) cycle

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, K.; Ohnishi, M.; Yamamoto, Y. [Kyoto Univ. (Japan)] [and others

    1994-12-31

    Key issues on a D-T Tokamak fusion reactor with advanced blanket concept using CFAB (Compact Fusion Advanced Brayton) cycle are presented. Although the previously proposed and studied compact fusion advanced Rankine cycle using mercury liquid metal has shown, in general, excellent performance characteristics in extracting energy and electricity with high efficiency by the {open_quotes}in-situ{close_quotes} nonequilibrium MHD disk generator, and in enhancing safety potential, there was a fear about uses of hazardous mercury as primary coolant as well as its limited natural resources. To overcome these disadvantages while retaining the advantage features of a ultra-high temperature coolant inherent in the synchrotron energy-enhanced D-T tokamak reactor, a compact fusion advanced Brayton cycle using helium was reexamined which was once considered relatively not superior in the CFAR study, at the expense of high, but acceptable circulation power, lower heat transfer characteristics, and probably of a little bit reduced safety.

  5. International Reactor Innovative and Secure (IRIS) summary

    International Nuclear Information System (INIS)

    Carelli, Mario D.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) reactor is described in the first part of the presentation. IRIS is a light water cooled reactor with an integral configuration, where steam generators, pumps and pressurizer are inside the reactor vessel. Partially funded by the DOE NERI program, IRIS is being developed by an international consortium of 16 organizations from seven countries. A key IRIS characteristic is its 'safety by design' approach which strives to eliminate, by design, as many accidents as possible rather than coping with their consequences. Initial returns are very positive; out of the eight Class IV accidents considered in the AP600 only one remains as a Class IV in IRIS, and at much reduced probability. Small-to-medium LOCAs have minimal consequences as the core remains safely under water for days, without the need for safety injection or water makeup. In spite of its novelty IRIS is firmly grounded on proven LWR technology and therefore a prototype is not needed to assure design certification. Rather, very extensive scaled tests will be performed to investigate the performance of in-vessel components such as steam generators and pumps, both individually and as interactive systems. Accident sequences will also be simulated and tested to prove IRIS safety by design claims. The first core fuel is less than 5% enriched and the fuel assembly is very similar to existing PWR assemblies, so there is no licensing challenge regarding the fuel. Because of the safety by design approach, yielding simplifications In design and accident management (e.g., IRIS does not have an emergency core cooling system), some accident scenarios are eliminated and others have lesser consequences. Thus, simplification and streamlining of the regulatory process might be possible. Risk informed regulation will be coupled with safety by design to show lower accident and damage probabilities. This could lead to a relaxation of siting regulatory requirements. It is

  6. The development of beryllium plasma spray technology for the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Castro, R.G.; Elliott, K.E.; Hollis, K.J.; Watson, R.D.

    1999-01-01

    Over the past five years, four international parties, which include the European Communities, Japan, the Russian Federation and the United States, have been collaborating on the design and development of the International Thermonuclear Experimental Reactor (ITER), the next generation magnetic fusion energy device. During the ITER Engineering Design Activity (EDA), beryllium plasma spray technology was investigated by Los Alamos National Laboratory as a method for fabricating and repairing and the beryllium first wall surface of the ITER tokamak. Significant progress has been made in developing beryllium plasma spraying technology for this application. Information will be presented on the research performed to improve the thermal properties of plasma sprayed beryllium coatings and a method that was developed for cleaning and preparing the surface of beryllium prior to depositing plasma sprayed beryllium coatings. Results of high heat flux testing of the beryllium coatings using electron beam simulated ITER conditions will also be presented

  7. 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

  8. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  9. ICRF boronization. A new technique towards high efficiency wall coating for superconducting tokamak reactors

    International Nuclear Information System (INIS)

    Li Jiangang; Zhao Yan Ping; Gu Xue Mao

    1999-01-01

    A new technique for wall conditioning that will be especially useful for future larger superconducting tokamaks, such as ITER, has been successfully developed and encouraging results have been obtained. Solid carborane powder, which is non-toxic and non-explosive, was used. Pulsed RF plasma was produced by a non-Faraday shielding RF antenna with RF power of 10 kW. The ion temperature was about 2 keV with a toroidal magnetic field of 1.8 T and a pressure of 3x10 -1 Pa. Energetic ions broke up the carborane molecules, and the resulting boron ions struck and were deposited on the first wall. In comparison with glow discharge cleaning boronization, the B/C coating film shows higher adhesion, more uniformity and longer lifetime during plasma discharges. The plasma performance was improved after ICRF boronization. (author). Letter-to-the-editor

  10. Steady-state temperature profile of the tokamak reactor in the neoclassic mode

    International Nuclear Information System (INIS)

    Volkov, T.F.; Igitkhanov, Yu.L.; Tokar, M.Z.

    It has been shown that when transfer processes are neoclassic in nature, it is possible to have a thermonuclear reactor combustion mode during which there is formed in the near-wall area a protective zone of dense cold plasma which is capable of preventing the wall from being eroded. The fact that it is possible to establish this mode is due to the ''opacity'' of reactor plasma with regard to neutrals

  11. Demountable low stress high field toroidal field magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.; Hsieh, D.; Lehner, J.; Suenaga, M.

    1977-01-01

    A new type of superconducting magnet system for large fusion reactors is described in this report. Instead of winding large planar or multi-axis coils, as has been proposed in previous fusion reactor designs, the superconducting coils are made by joining together several prefabricated conductor sections. The joints can be unmade and sections removed if they fail. Conductor sections can be made at a factory and shipped to the reactor site for assembly. The conductor stress level in the assembled coil can be kept small by external support of the coil at a number of points along its perimeter, so that the magnetic forces are transmitted to an external warm reinforcement structure. This warm reinforcement structure can also be the primary containment for the fusion reactor, constructed similar to a PCRV (Prestressed Concrete Reactor Vessel) used in fission reactors. Low thermal conductivity, high strength supports are used to transfer the magnetic forces to the external reinforcement through a hydraulic system. The hydraulic supports are movable and can be programmed to accommodate thermal contraction and to minimize stress in the superconducting coil

  12. Demountable low stress high field toroidal field magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.; Hsieh, D.; Lehner, J.; Suenaga, M.

    1978-01-01

    A new type of superconducting magnet system for large fusion reactors is described. Instead of winding large planar or multi-axis coils, as has been proposed in previous fusion reactor designs, the superconducting coils are made by joining together several prefabricated conductor sections. The joints can be unmade and sections removed if they fail. Conductor sections can be made at a factory and shipped to the reactor site for assembly. The conductor stress level in the assembled coil can be kept small by external support of the coil at a number of points along its perimeter, so that the magnetic forces are transmitted to an external warm reinforcement structure. This warm reinforcement structure can also be the primary containment for the fusion reactor, constructed similar to a PCRV (Prestressed Concrete Reactor Vessel) used in fission reactors. Low thermal conductivity, high strength supports are used to transfer the magnetic forces to the external reinforcement through a hydraulic system. The hydraulic supports are movable and can be programmed to accommodate thermal contraction and to minimize stress in the superconducting coil. (author)

  13. Nuclear reactor internals construction and failed fuel rod detection system

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  14. Physics design of advanced steady-state tokamak reactor A-SSTR2

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Ushigusa, Kenkichi

    2000-10-01

    Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

  15. RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

    International Nuclear Information System (INIS)

    Ferrada, Juan J.; Reiersen, Wayne T.

    2011-01-01

    U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment

  16. A D-T tokamak fusion reactor with advanced blanket using compact fusion advanced Rankine (CFAR) cycle

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, K.; Shimohiro, D.; Yamamoto, Y.; Toku, H. (Inst. of Atomic Energy, Kyoto Univ., Uji (Japan)); Inui, Y.; Ishikawa, M.; Umoto, J.; Miki, N. (Dept. of Electrical Engineering, Kyoto Univ. (Japan)); Fukuyama, A. (Dept. of Electrical and Electronic Engineering, Okayama Univ. (Japan)); Mitarai, O. (Dept. of Electrical Engineering, Kumamoto Inst. of Tech. (Japan)); Okamoto, M.; Sekimoto, H.; Fujii, Y.; Kim, Hu; Watanabe, K.; Ishii, T. (Research Lab. for Nuclear Engineering, Tokyo Inst. of Tech., Meguro (Japan)); Takagi, T. (Inst. of Fluid Science, Univ. Tohoku, Sendai (Japan)); Suzuki, T.; Mutoh, I. (National Research Inst. for Metals, Tsukuba, Ibaraki (Japan)); Miyazaki, K.; Saitoh, M. (Dept. of Nuclear Engineering, Osaka Univ., Suita (Japan)); Logan, B.G.; Barr, W.L. (Lawrence Livermore National Lab., CA (United States)); Hoffman, M.A. (Univ. of California, Davis, CA (United States)); Campbell, R.B. (Sandia National Lab., Albuquerque, NM (United States))

    1991-12-01

    Recent progress of a D-T tokamak fusion reactor with advanced blanket concept using CFAR (Compact Fusion Advanced Rankine) cycle is presented. High enthalpy extraction in excess of 50% was found achievable by a nonequilibrium disk-type MHD generator for gas stagnation temperatures of 3000 K superheated by synchrotron radiation, leading to a 40% overall plant efficiency with recuperative heat cycle and advanced thermoelectric converters. The stored magnetic energy of a 10 T MHD three-pairs Helmholtz-type magnet was estimated to be about 600 MJ per MHD generator. A mixture of mercury and lithium liquid flow was examined and found applicable to reduce the mercury inventory. The tritium permeation loss inherent in the relatively high temperature CFAR cycle, neutronics relating to tritium breeding and streaming through the waveguide and the MHD generator duct are discussed. Results of a 2000 h high temperature (600-700deg C) mercury corrosion test of Type 304 stainless steel, a TZM alloy and four kinds of ceramic rods in refluxing mercury with 0.004 mol fraction of potassium admixture have shown no noticeable corrosion except for Type 304 stainless steel. (orig.).

  17. A conceptual composite blanket design for the Tokamak type of thermonuclear reactor incorporating thermoelectric pumping of liquid lithium

    International Nuclear Information System (INIS)

    Dutta Gupta, P.B.

    1981-01-01

    The conceptual liquid lithium blanket design for the tokamak type of thermonuclear reactor put forward is a modification of the initial simple but novel design concept enunciated earlier that exploits the availability of suitably oriented magnetic fields and temperature gradients within the blanket to pump the liquid as has been shown feasible by laboratory model experiments. The modular construction of the blanket cells is retained but the earlier simple back to back double spiralling channel module is replaced by a composite unit of three radially nested layer-structures to optimise heat and tritium extraction from the blanket. The layer-structure at the first wall generates liquid lithium circulation by thermoelectric magnetohydrodynamic forces and the segregated double spiralling channels serve as inlet-outlet driving devices. The outermost layer-structure is cooled by helium. Liquid lithium in the intermediate layer-structure is pumped at a very slow rate. The choice of the relative dimensional proportions of the three layer-structure and the channel cross-section, material property and the spiralling contour is of critical importance for the design. This paper presents the design data for a conceptual design of such a blanket with a 5000 MW (th) rating. (author)

  18. Hydrogen isotope trapping on graphite collectors during an isotope exchange experiment in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Kilpatrick, S.J.; Ulrickson, M.; Dylla, H.F.; Manos, D.M.; Ramsey, A.T.; Nygren, R.; Hirooka, Y.; Wampler, W.R.

    1989-03-01

    A rotatable collector probe was used to expose several graphite samples to a deuterium-to-hydrogen-to-deuterium exchange experiment in the Tokamak Fusion Test Reactor (TFTR) at the start of the 1988 operations period. This experiment proved the utility of helium conditioning discharges in accelerating the changeover process. Samples included portions of a tile taken from the inner bumper limiter (POCO AXF-5Q graphite) of TFTR during the recent machine opening, and coupons which had been conditioned in the Plasma Interactive Surface Component Experimental Station (PISCES) by exposure to a helium plasma. The samples were exposed to different groups of the /approximately/100 1.4 MA discharges that comprised the experiment. They were removed and analyzed for retained deuterium and impurities by nuclear reaction analysis and Rutherford backscattering spectrometry. Codeposited carbon layers had been formed with thicknesses up to several tenths of a micron. The inferred percentages of trapped hydrogenic species were in general agreement with spectroscopic data. The computed carbon fluence per D + discharge, 1.2 /times/ 10 17 C/cm 2 , is compared to recent measurements on limiter tiles removed from TFTR. 22 refs., 3 figs., 1 tab

  19. First wall/blanket/shield design and power conversion for the ARIES-IV tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Conn, R.W.; Najmabadi, F.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10 MPa base pressure. The coolant flows poloidally in two loops, one inboard and one outboard. The coolant channels are circular tubes that form shells and are placed between two purge plates; the space between two adjacent tubes and the plate is purge gas flow area. The solid breeder is Li 2 O, and Be is used as neutron multiplier to ensure adequate TBR. Beryllium and Li 2 O are placed in between the adjacent tube shells. A computer code was developed to perform and optimize thermal-hydraulic design. Minimization of blanket thickness and the amount of Be, and the maximization of breeder zone thickness were done by iteration with neutronics. The gross thermal efficiency is 49%. The cost of electricity is 68 mills/kWh. The use of low activation SiC composite as the structural material, Li 2 O as the solid breeder, and avoidance of tungsten in the divertor has resulted in a good safety performance, and LSA rating of 1. Overall, SiC/He/Li 2 O ARIES-IV design is expected to have attractive economic and safety advantages

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

    International Nuclear Information System (INIS)

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

    1989-03-01

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

  1. Reactor applications of the compact fusion advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, H.A.; Logan, B.G.; Campbell, R.B.

    1988-03-01

    We have made a preliminary design of a D-T fusion reactor blanket and MHD power conversion system based on the CFAR concept, and found that the performance and costs for the reference cycle are very attractive. While much remains to be done, the potential advantage of liquid metal Rankine cycles for fusion applications are much clearer now. These include low pressures and mass flow rates, a nearly isothermal module shell which minimizes problems of thermal distortion and stresses, and an insensitivity to pressure losses in the blanket so that the two-phase MHD pressure drops in the boilling part of the blanket and the ordinary vapor pressure drops in the pebble-bed superheating zones are acceptable (the direct result of pumping a liquid rather than having to compress a gas). There are no moving parts in the high-temperature MHD power generators, no steam bottoming plant is required, only small vapor precoolers and condensers are needed because of the high heat rejection trmperatures, and only a relatively small natural-draft heat exhanger is required to reject the heat to the atmosphere. The net result is a very compact fusion reactor and power conversion system which fit entirely inside an 18 meter radius reactor vault. Although we have not yet performed a detailed cost analysis, preliminary cost estimates indicate low capital costs and a very attractive cost of electricity. 11 refs., 5 figs., 2 tabs.

  2. A study on the fusion reactor - Study of ICRF coupling in the KAIST tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Duk In; Chang, Hong Young; Lee, Sun Chil; Jun, Sang Jin; Kwon, Gi Chung; Seo, Sung Hun; Heo, Sung Hoi; You, Kwang Il; Song, Soo Bin; Lee, Sung Chul; Kim, Min Chul; Lee, Chan Hui [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-09-01

    Research objectives are to design and fabricate the antenna, measure t property of absorption transmitted to the plasma, and research the physical phenomena about the ICRF coupling, Main heating method is ohmic heating at the KAIST tokamak. The power of the plasma produced by ohmic heating is about 100 kW. Because the toroidal field is 5 {approx} 8 kG, the RF system`s output power is about 10 kW and frequency range is 7 {approx} 30 MHz. In the first year, a 1 kW RF preamplifier was bought. In this year, a CW 2 kW RF main amp. and RF power monitoring system was bought. In the research on antenna, we study the method how to measure electric field emitted from antenna using piezo elements. The matching network composed of two VVC (35 kV), 100 {approx} 1000 pF match firmly up to 50 kW power. We studied the measurement method of antenna impedance theoretically, and measured power efficiency and antenna impedance in the helicon plasma. 32 refs., 5 tabs., 29 figs. (author)

  3. Evaluation of neutron streaming through injection ports in a tokamak-type fusion reactor

    International Nuclear Information System (INIS)

    Ide, Takahiro; Seki, Yasushi; Iida, Hiromasa

    1976-03-01

    The effects of neutron streaming through injection ports in the fusion reactor designed in JAERI have been studied, especially those on tritium breeding ratio and the shielding of the superconducting magnet. In placement of the injection ports in the blanket, the tritium breeding ratio decreases by up to 1.3%, and shielding problem of the superconducting magnet is very important. (auth.)

  4. Lithium adsorption by the first wall of fusion reactor-tokamak

    International Nuclear Information System (INIS)

    Bakunin, O.G.

    1989-01-01

    Lithium adsorption by the first wall of fusion reactor under stationary conditions and in the absence of chemical reactions is considered. Possibility of achieving 70% coating of the wall with lithium which can lead to sufficient decrease of sputtering is shown. 5 refs.; 5 figs

  5. On the survivability of diagnostic windows in the CIT [Compact Ignition Tokamak] reactor

    International Nuclear Information System (INIS)

    Taylor, A.

    1988-11-01

    The problem of radiation induced stresses in CIT diagnostic windows is discussed. Existing data indicate windows of existing design will probably survive if placed on the periphery of the reactor. There is a lack of adequate engineering data upon which the design and survivability of windows can be based. 22 refs., 5 figs., 2 tabs

  6. Tokamak Fusion Test Reactor. Final conceptual design report. [Overall cost and scheduling program

    Energy Technology Data Exchange (ETDEWEB)

    1976-02-01

    The TFTR is the first U.S. magnetic confinement device planned to demonstrate the fusion of D-T at reactor power levels. This report addresses the physics objectives and the engineering goals of the TFTR project. Technical, cost, and schedule aspects of the project are included. (MOW)

  7. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

    Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

    1979-01-01

    This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties to the benefits and conclusions and recommendations on resolving the issue are discussed

  8. The impact of confinement scaling on ITER [International Thermonuclear Experimental Reactor] parameters

    International Nuclear Information System (INIS)

    Reid, R.L.; Galambos, J.D.; Peng, Y.K.M.

    1988-09-01

    Energy confinement scaling is a major concern in the design of the International Thermonuclear Experimental Reactor (ITER). The existing database for tokamaks can be fitted with a number of different confinement scaling expressions that have similar degrees of approximation. These scaling laws predict confinement times for ITER that vary by over an order of magnitude. The uncertainties in the form and magnitude of these scaling laws must be substantially reduced before the plasma performance of ITER can be predicted with adequate reliability. The TETRA systems code is used to calculate the dependence of major ITER parameters on the scaling laws currently in use. Design constraints of interest in the present phase of ITER consideration are used, and the minimum-cost devices arising from these constraints are reviewed. 9 refs., 13 figs., 4 tabs

  9. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  10. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  11. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

    1980-06-01

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  12. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  13. Anomalous loss of DT alpha particles in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, Hans W. [Princeton Univ., NJ (United States)

    1997-09-01

    An escaping alpha collector probe has been developed for TFTR`s DT phase. Energy distributions of escaping alphas have been determined by measuring the range of α-particles implanted into nickel foils located within the alpha collector. Results at 1.0 MA of plasma current are in good agreement with predictions for first orbit alpha loss. Results at 1.8 MA, however, show a significant anomalous loss of partially thermalized alphas (in addition to the expected first orbit loss), which is not observed with the lost alpha scintillator detectors in DT plasmas, but does resemble the anomalous delayed loss seen in DD plasmas. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations. An experiment designed to study the effect of plasma major radius shifts on α-particle loss has led to a better understanding of α-particle dynamics in tokamaks. Intuitively, one might suppose that confined marginally passing α-particles forced to move toward higher magnetic field during an inward major radius shift (i.e., compression) would mirror and become trapped particles, leading to increased alpha loss. Such an effect was looked for during the shift experiment, however, no significant changes in alpha loss to the 90° lost alpha scintillator detector were observed during the shifts. It is calculated that the energy gained by an α-particle during the inward shift is sufficient to explain this result. However, an unexpected loss of partially thermalized α-particles near the passing/trapped boundary was observed to occur between inward and outward shifts at an intermediate value of plasma current (1.4 MA). This anomalous loss feature is not yet understood.

  14. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    Willem Janssens, Ir.; Crutzen, Y.; Farfaletti-Casali, F.; Matera, R.

    1991-01-01

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

  15. Tritium system for a tokamak reactor with a self-pumped limiter

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Sze, D.K.

    1986-01-01

    The self-pumping concept was proposed as a means of simplifying the impurity control system in a fusion reactor. The idea is to remove helium in-situ by trapping in freshly deposited metal surface layers of a limiter or divertor. Trapping material is added to the plasma scrape-off or edge region where it is transported to the wall. Some of the key issues for this concept are the tritium inventory in the trapping material and the permeation of protium and recycling of tritium. These quantities are shown to be acceptable for the reference design. The tritium issues for a helium-cooled solid breeder reactor design with vanadium alloy as a structural material are also examined. Models are presented for tritium permeation and inventory calculation for structure materials with the effect of a thin layer of coating material

  16. Design of a tokamak fusion reactor first wall armor against neutral beam impingement

    International Nuclear Information System (INIS)

    Myers, R.A.

    1977-12-01

    The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiation damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem

  17. The system of underwater CCTV inspection for reactor internal components

    International Nuclear Information System (INIS)

    Zhu Rong

    1997-12-01

    During the operation of nuclear power plant, the reactor internal components are greatly scoured and vibrated by flowing water. So the structural integrity and surface sludge for reactor internal components are needed to be inspected during refuelling. Thus an inspection system is developed, in which the camera inspects underwater at different height and different direction by mechanical elevator and the image of closed-circuit television (CCTV) is mixed with digital coordinate of the camera position for re-inspection. It is the first system for inspection of reactor internal components in China. This system has been used 4 times in the inspection of Daya Bay Nuclear Power Plant successfully

  18. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J.A.

    1994-11-30

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

  19. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11-12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of

  20. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    International Nuclear Information System (INIS)

    Azizov, E A

    2012-01-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined. (conferences and symposia)

  1. Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

    International Nuclear Information System (INIS)

    Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

    2002-01-01

    these works is also a part of further Kazakhstan contribution in ITER project. JSC 'UMP' provides feasibility study of manufacturing possibility of complete details from beryllium together with superconductor components taking into consideration ITER project requirements. It is supposed that new grades of beryllium, such as SS65C or DSHG-200, will be developed or alternative grade of beryllium with its further verification will be used. As a support for project ITER participation as well as for development of new contemporary trends in science and technology development, and for training the high skilled personnel a unique and the only one in the world KTM experimental tokamak installation is being developed in the country. It will be used for testing structural materials and important elements of thermonuclear reactors. At present a draft of scientific program for KTM reactor researches has been worked out; also feasibility study has been conducted, preliminary design is completed and the work of design documentation development and modeling of the installation is got under way. KTM tokamak, which is constructed in NNC RK in partnership with Russian organizations, will allow to perform predicative tests of future ITER reactor assemblies being produced at JSC 'UMP'. It will promote international cooperation in the field of thermonuclear material science and new material production. It will also contribute greatly in new technologies and modern designs of thermonuclear fusion reactors debugging. ITER and KTM projects are the examples of Successful Eurasia cooperation in the field of peaceful atomic energy use

  2. Conceptual Design of Tokamak High Power Reactor (A-SSTR2)

    Science.gov (United States)

    Nishio, Satoshi; Tobita, Kenji; Ushigusa, Kenkichi; Konishi, Satoshi; Reactor Design Team

    Based on the last decade of JAERI reactor design studies, an advanced commercial reactor concept (A-SSTR2) that meets both economical and environmental requirements has been proposed. The A-SSTR2 is a compact power reactor (Rp = 6.2 m, ap = 1.5 m, Ip = 12 MA) with a high fusion power (Pf = 4 GW) and a net thermal efficiency of 51%. The machine configuration is simplified by eliminating the center solenoid (CS) coil system. An SiC/SiC composite for the blanket structure material, helium gas cooling with a pressure of 10 MPa and an outlet temperature of 900°C, and TiH2 for the bulk shield material are introduced. For the toroidal field (TF) coil, a high temperature (TC) superconducting wire made of bismuth with a maximum field of 23 T and a critical current density of 1,000 A/mm2 at a temperature of 20 K is applied. In spite of the CS-less configuration, a computer simulation gives satisfactory plasma equilibria, plasma initiation process, and current ramp-up scenario. The current rampup time is about 22 hours. The MHD stabilities for the ballooning mode and the ideal low n kink-modes are confirmed. The stabilization of n = 1 and n = 2 kink modes requires a shell position closer than 1.4 times and 1.2 times the plasma minor radius, respectively. With regard to the divertor thermal condition, it was found that a neon gas-seeded divertor plasma with a fraction of ˜ 2.5% gives a thermal load reduction at the divertor plate from 460 MW to 100 MW and a plasma temperature drop at the divertor plate from 200 eV to 20 ˜ 30 eV. By increasing the shield thickness by about 15 cm, the total radwaste is dramatically reduced. The radwaste percentage relative to the total waste is reduced from 92 wt.% to 17 wt.%.

  3. Materials problems and possible solutions for near term Tokamak fusion reactors

    International Nuclear Information System (INIS)

    Kulcinski, G.L.

    1978-01-01

    It is the purpose of this paper to clarify the magnitude of the problems that might arise from neutron damage and to put into perspective the methods and facilities that might be used to solve these problems. First, a brief review of some of the fundamental aspects of radiation damage from neutrons will be given for the non-materials scientist followed by a current listing of the anticipated radiation environment of the various near term (TFTR, JET, T-20), EPR, and DPR designs. The reader should note that such designs are highly fluid and may change considerably in the future (in fact due to the very problem we will be discussing). Next, the present and future facilities that could be used to test CTR materials will be reviewed and their utility in providing pertinent fundamental and engineering data will be discussed. Finally, some conclusions and recommendations on the near term reactor materials problems will be presented

  4. Comparison of explicit calculations for n = 3 to 8 dielectronic satellites of the FeXXV Kα resonance line with experimental data from the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Decaux, V.; Bitter, M.; Hsuan, H.; Hill, K.W.; von Goeler, S.; Park, H.; Bhalla, C.P.

    1991-12-01

    Dielectronic satellite spectra of the FeXXV Kα resonance line observed from the Tokamak Fusion Test Reactor (TFTR) plasmas have been compared with recent explicit calculations for the n = 3 to 8 dielectronic satellites as well as the earlier theoretical predictions, which were based on the 1/n 3 scaling law for n > 4 satellites. The analysis has been performed by least-squares fits of synthetic spectra to the experimental data. The synthetic spectra constructed from both theories are in good agreement with the observed data. However, the electron temperature values obtained from the fit of the present explicit calculations are in better agreement with independent measurements. 20 refs., 4 figs

  5. Resistive edge mode instability in stellarator and tokamak geometries

    Science.gov (United States)

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-01

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

  6. Current drive studies for the ARIES steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Ehst, D.A.; Mandrekas, J.

    1994-01-01

    Steady-state plasma operating scenarios are designed for three versions of the ARIES reactor, using non-inductive current drive techniques that have an established database. R.f. waves, including fast and lower hybrid waves, are the reference drivers for the D-T burning ARIES-I and ARIES-II/IV, while neutral beam injection is employed for ARIES-III which burns D- 3 He. Plasma equilibria with a high bootstrap-current component have been used, in order to minimize the recirculating power fraction and cost of electricity. To maintain plasma stability, the driven current profile has been aligned with that of equilibrium by proper choices of the plasma profiles and power launch parameters. Except for ARIES-III, the current-drive power requirements and the relevant technology developments are found to be quite reasonable. The wave-power spectrum and launch requirements are also considered achievable with a modest development effort. Issues such as an improved database for fast-wave current drive, lower-hybrid power coupling to the plasma edge, profile control in the plasma core, and access to the design point of operation remain to be addressed. ((orig.))

  7. Measurement of local, internal magnetic fluctuations via cross-polarization scattering in the DIII-D tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Barada, K., E-mail: kshitish@ucla.edu; Rhodes, T. L.; Crocker, N. A.; Peebles, W. A. [University of California-Los Angeles, P.O. Box 957099, Los Angeles, California 90095 (United States)

    2016-11-15

    We present new measurements of internal magnetic fluctuations obtained with a novel eight channel cross polarization scattering (CPS) system installed on the DIII-D tokamak. Measurements of internal, localized magnetic fluctuations provide a window on an important physics quantity that we heretofore have had little information on. Importantly, these measurements provide a new ability to challenge and test linear and nonlinear simulations and basic theory. The CPS method, based upon the scattering of an incident microwave beam into the opposite polarization by magnetic fluctuations, has been significantly extended and improved over the method as originally developed on the Tore Supra tokamak. A new scattering geometry, provided by a unique probe beam, is utilized to improve the spatial localization and wavenumber range. Remotely controllable polarizer and mirror angles allow polarization matching and wavenumber selection for a range of plasma conditions. The quasi-optical system design, its advantages and challenges, as well as important physics validation tests are presented and discussed. Effect of plasma beta (ratio of kinetic to magnetic pressure) on both density and magnetic fluctuations is studied and it is observed that internal magnetic fluctuations increase with beta. During certain quiescent high confinement operational regimes, coherent low frequency modes not detected by magnetic probes are detected locally by CPS diagnostics.

  8. 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

  9. Requirements for US regulatory approval of the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Petti, D.A.; Haire, J.C.

    1993-12-01

    The International Thermonuclear Experimental Reactor (ITER) is the first fusion machine that will have sufficient decay heat and activation product inventory to pose potential nuclear safety concerns. As a result, nuclear safety and environmental issues will be much more important in the approval process for the design, siting, construction, and operation of ITER in the United States than previous fusion devices, such as the Tokamak Fusion Test Reactor. The purpose of this report is (a) to provide an overview of the regulatory approval process for a Department of Energy (DOE) nuclear facility; (b) to present the dose limits used by DOE to protect workers, the public, and the environment from the risks of exposure to radiation and hazardous materials; (c) to discuss some key nuclear safety-related issues that must be addressed early in the Engineering Design Activities (EDA) to obtain regulatory approval; and (d) to provide general guidelines to the ITER Joint Central Team (JCT) concerning the development of a regulatory framework for the ITER project

  10. Hydrodynamic study of an internal airlift reactor for microalgae culture.

    Science.gov (United States)

    Rengel, Ana; Zoughaib, Assaad; Dron, Dominique; Clodic, Denis

    2012-01-01

    Internal airlift reactors are closed systems considered today for microalgae cultivation. Several works have studied their hydrodynamics but based on important solid concentrations, not with biomass concentrations usually found in microalgae cultures. In this study, an internal airlift reactor has been built and tested in order to clarify the hydrodynamics of this system, based on microalgae typical concentrations. A model is proposed taking into account the variation of air bubble velocity according to volumetric air flow rate injected into the system. A relationship between riser and downcomer gas holdups is established, which varied slightly with solids concentrations. The repartition of solids along the reactor resulted to be homogenous for the range of concentrations and volumetric air flow rate studied here. Liquid velocities increase with volumetric air flow rate, and they vary slightly when solids are added to the system. Finally, liquid circulation time found in each section of the reactor is in concordance with those employed in microalgae culture.

  11. International Harmonization of Reactor Licensing Regulations

    International Nuclear Information System (INIS)

    Kuhnt, Dietmar.

    1977-01-01

    The purpose of a harmonization policy for reactor licensing regulations on the basis of already considerable experience is to attain greater rationalisation in this field, in the interest of economic policy and healthy competition, and most important, radiation protection and safety of installations. This paper considers the legal instruments for such harmonization and the conditions for their implementation, in particular within the Communities framework. (NEA) [fr

  12. International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

    International Nuclear Information System (INIS)

    2007-01-01

    Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

  13. Driftwave-based modeling of poloidal spin-up precursor and step-wise expansion of internal transport barriers in tokamaks

    International Nuclear Information System (INIS)

    Staebler, G.M.; Waltz, R.E.; Kinsey, J.E.; Bateman, G.; Kritz, A.H.; Onjun, T.; Pankin, A.; Zhu, P.; Horton, W.

    2001-01-01

    The rich phenomenology of internal transport barriers observed in tokamaks includes a poloidal spin-up precursor for balanced injection neutral beam heating and step-wise expansion of the barrier for unbalanced injection. Examples of numerical simulations of these phenomena are presented. Two driftwave-based predictive transport models (GLF23 and Multi-Mode) are used. Both models include the suppression of ion temperature gradient modes as the EB shear approaches the computed maximum linear growth rate. Modeling of discharges with internal transport barriers from the DIII-D, JET and TFTR tokamaks are compared. (author)

  14. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

  15. An Overview of the International Reactor Physics Experiment Evaluation Project

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gulliford, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-09

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  16. Proceedings of the international symposium on materials testing reactors

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Kawamura, Hiroshi

    2009-01-01

    This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

  17. International students workshop on innovative light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anglart, H. [Royal Inst. of Tech., Stockholm (Sweden). Nuclear Reactor Technology; Laurien, E. [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme; Schulenberg, T. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Kern- und Energietechnik

    2008-06-15

    Nuclear reactor design is still one of the most fascinating subjects of mechanical engineering. Thirty students from 10 worldwide nations demonstrated this impressively in a recent workshop on supercritical water cooled reactors of the 4{sup th} generation, held from March 31 to April 3, 2008, in Karlsruhe, Germany, hosted by the Karlsruhe Institute of Technology. Bachelor and master students as well as young scientists working on their doctorate presented their own particular contribution to design and analyses of innovative reactor components, including its safety systems and other plant design. Their presentations were accompanied by lectures of leading scientists working in the European project of the 'High Performance Light Water Reactor' which is sponsored by the European Commission as part of its 6{sup th} Framework Programme. The workshop is an initiative of the Generation IV International Forum. (orig.)

  18. International students workshop on innovative light water reactors

    International Nuclear Information System (INIS)

    Anglart, H.; Laurien, E.; Schulenberg, T.

    2008-01-01

    Nuclear reactor design is still one of the most fascinating subjects of mechanical engineering. Thirty students from 10 worldwide nations demonstrated this impressively in a recent workshop on supercritical water cooled reactors of the 4 th generation, held from March 31 to April 3, 2008, in Karlsruhe, Germany, hosted by the Karlsruhe Institute of Technology. Bachelor and master students as well as young scientists working on their doctorate presented their own particular contribution to design and analyses of innovative reactor components, including its safety systems and other plant design. Their presentations were accompanied by lectures of leading scientists working in the European project of the 'High Performance Light Water Reactor' which is sponsored by the European Commission as part of its 6 th Framework Programme. The workshop is an initiative of the Generation IV International Forum. (orig.)

  19. Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

    Science.gov (United States)

    Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

    2018-04-01

    1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

  20. Tokamak reactor systems studies

    International Nuclear Information System (INIS)

    1992-01-01

    A summary of work completed on the ARIES project during this report period is given. The main areas of effort were: neutronics, shield optimization and design, safety, systems, startup and shutdown, and ripple loss

  1. Long Pulse Technology Tokamak

    International Nuclear Information System (INIS)

    Jernigan, T.C.

    1978-01-01

    The LPTT tokamak is a non-circular tokamak (R = 1.5 m, a = .45 m) proposed by ORNL for extended pulse operation at high β (5%) and reactor level wall power loading (40 w/cm 2 ). The toroidal field coils are superconducting and a super-conducting bundle divertor is proposed for active impurity control. All systems are designed for continuous operation which will provide pulse lengths > 20 seconds with a 6 to 10 weber flux swing. Experimental access and flexibility in operation are primary design goals

  2. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and The International Reactor Reactor Physics Experiment Evaluation Project (IRPhEP)

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, J.B.; Bess, J. [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Gulliford, J. [Organization for Economic Cooperation and Development (OECD),Nuclear Energy Agency, Paris, (France)

    2011-07-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) are sources of evaluated integral benchmark data that may be used for validation of reactor physics / nuclear criticality safety analytical methods and data, nuclear data testing, and safety analysis licensing activities. The IRPhEP is patterned after its predecessor, the ICSBEP, but focuses on other integral measurements such as buckling, spectral characteristics, reactivity effects, reactivity coefficients, kinetics measurements, reaction-rate and power distributions, nuclide compositions and other miscellaneous types of measurements in addition to the critical configuration. Both projects will be discussed.

  3. Upper internals arrangement for a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

    2013-07-09

    In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

  4. Internals segmentation at Shoreham 850 MWe boiling water reactor

    International Nuclear Information System (INIS)

    Hirsch, D.J.; Garvey, T.F.

    1993-01-01

    The decommissioning of the Shoreham Nuclear Power Station required the removal and segmentation of the reactor internals. PCI Energy Services, a contractor specialising in remote cutting, machining, and welding performed this work. To make the required cuts, twenty-five different tools were designed, fabricated, assembled, and qualified. The tools included an internals work platform for cutting components in the reactor pressure vessel (RPV), a Wet Cutting Station with a 15' long Plasma Arc Cutting torch delivery mast for segmenting components after removal from the Reactor Pressure Vessel, and a Dry Cutting Station for in-air segmenting of less contaminated components. Other tools included remote-operated torch delivery tools, remote vision systems, filtration systems, and airborne contaminant catchment systems. The work started in mid-June 1992 and was completed in November 9 1992, ahead of schedule and below the ALARA goal. (author)

  5. The physics of tokamak start-up

    International Nuclear Information System (INIS)

    Mueller, D.

    2013-01-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current

  6. Lower internals for pressurized water reactor

    International Nuclear Information System (INIS)

    Chevereau, G.; Babin, M.

    1989-01-01

    The lower internals for PWR has a separating plate mounted beneath its lower core plate and defining a distribution chamber with it, peripheral mechanical connectors joining the plates separated by coolant passage and apertures in the separation plate connected to a coolant pipe [fr

  7. Underwater Remote Handling Equipment for Reactor Internals Maintenance

    International Nuclear Information System (INIS)

    Motohiko Kimura; Mitsuaki Shimamura; Tomoyuki Itoh; Nobuhiko Tanaka; Yasuhiro Yuguchi; Katsuhiko Naruse

    2002-01-01

    More than fifty nuclear reactors generate about thirty-five percent of electricity in Japan. The need to operate these reactors safely and in a stable manner constitutes a very important issue. On the other hand, aged reactors are increasing and they are not necessarily designed and constructed using the latest technology. Stress Corrosion Cracking (SCC) on reactor internal components has become a major concern regarding aged reactors in recent years. Usually maintenance work such as inspection, repair, and preventive maintenance for core components is done by using underwater remote handling and robotic technology. It becomes very important to develop not only new efficient technology for inspection, repair, and preventive maintenance for all suspect components and but also the associated application technology for execution in a reactor. We have been developing several kind of remote handling equipment for underwater maintenance work. This paper describes some results obtained in the area of underwater remote handling that can contribute to the progress of plant reliability. (authors)

  8. Nuclear reactor internals with control elements guides

    International Nuclear Information System (INIS)

    Baujat, J.; Chevereau, G.

    1991-01-01

    The internals have a lower plate, a superior plate, support columns and guide tubes for the control rods displacements. The lower section of the control rod guide tube have a base that fits into a bevelled seat in the lower plate. The guide tube is held into the seat by a spring, compressed between the base of the upper section of the tube and the lower plate

  9. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  10. Formation of an internal transport barrier in the ohmic H-mode in the TUMAN-3M tokamak

    International Nuclear Information System (INIS)

    Andrejko, M.V.; Askinazi, L.G.; Golant, V.E.; Zhubr, N.A.; Kornev, V.A.; Krikunov, S.V.; Lebedev, S.V.; Levin, L.S.; Razdobarin, G.T.; Rozhdestvensky, V.V.; Smirnov, A.I.; Tukachinsky, A.S.; Yaroshevich, S.P.

    2000-01-01

    In experiments on studying the ohmic H-mode in the TUMAN-3M tokamak, it is found that, in high-current (I p ∼ 120-170 kA) discharges, a region with high electron-temperature and density gradients is formed in the plasma core. In this case, the energy confinement time τ E attains 9-18 ms, which is nearly twice as large as that predicted by the ELM-free ITER-93H scaling. This is evidence that the internal transport barrier in a plasma can exist without auxiliary heating. Calculations of the effective thermal diffusivity by the ASTRA transport code demonstrate a strong suppression of heat transport in the region where the temperature and density gradients are high

  11. International standardization of nuclear reactor designs - the way forward

    International Nuclear Information System (INIS)

    Raetzke, Christian

    2010-01-01

    The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

  12. ITER at the international conference on fusion reactor materials

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.; Matera, R.

    1998-01-01

    The reports summarizes the topics of the eighth International Conference on Fusion Reactor Materials (ICFRM-8) which was held in Sendai, Japan, on 26-31 October 1997. The ICFRM is focused on the whole spectrum of materials and technologies to be applied in fusion reactors and related facilities. The total number of conference participants was over 500, representing 24 countries and about 600 oral and poster papers were presented at the conference. Three sessions were devoted to ITER materials: (i) Design-Materials Interface and ITER (oral session); (ii) ITER, Irradiation Facility and Technology, (poster session); (iii) ITER and Beyond (discussion session)

  13. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  14. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  15. Moderator behaviour and reactor internals integrity at Atucha I NPP

    International Nuclear Information System (INIS)

    Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

    1996-01-01

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

  16. International Conference on Physics and Technology of Reactors and Applications

    International Nuclear Information System (INIS)

    2007-01-01

    The first international conference on physics and technology of reactors and applications (PHYTRA 1) which took place in Marrakech (Morocco) from 14 to 16 March 2007, was designed to bring together scientists, teachers and students from universities, research centres and industry and other institutions to exchange knowledge and to discuss ideas and future issues. The programmes of the PHYTRA 1 conference covers a wide variety topics, the conference was organised in three plenary sessions, ten oral technical sessions and two poster sessions. The plenary sessions covers the following topics : The prospects of nuclear energy, The situation of nuclear sciences and energy in Morocco and Africa, and the new development in reactor physics and reactor design [fr

  17. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  18. International standardization of safety requirements for fast reactors

    International Nuclear Information System (INIS)

    2011-06-01

    Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

  19. Dimensional control and check of field machining parts for reactor internals installation

    International Nuclear Information System (INIS)

    Zhang Caifang

    2010-01-01

    Some key issues of dimensional control for reactor internals installation are analyzed, and important technical requirements of crucial quality control elements on the measurement, machining, and checking of reactor internals filed machining parts are discussed. Moreover, provisions on quality control and risk prevention of reactor internals filed machining parts are presented in this paper. (author)

  20. World's largest DC flywheel generator for the toroidal field power supply of JAERI's JFT-2M Tokamak nuclear fusion reactor

    International Nuclear Information System (INIS)

    Tani, Takashi; Nakanishi, Yuji; Horita, Tsuyoshi; Kawase, Chiharu; Oyabu, Isao; Kishimoto, Takeshi.

    1996-01-01

    Mitsubishi Electric has delivered the world's largest DC generator for the toroidal field coil power supply of the JFT-2M Tokamak at the Japan Atomic Energy Research Institute. The unit rotates at 225 or 460 rpm, providing a maximum rated output of 2,700 V, 19,000 A and 51.3 MW. The toroidal field is a DC field, so use of a DC generator permits a simpler design consuming less floor space than an AC drive system. The generator was manufactured following extensive studies on commutation, mechanical strength and insulation. (author)

  1. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    Energy Technology Data Exchange (ETDEWEB)

    Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

  2. Improving nuclear safety at international research reactors: The Integrated Research Reactor Safety Enhancement Program (IRRSEP)

    International Nuclear Information System (INIS)

    Huizenga, David; Newton, Douglas; Connery, Joyce

    2002-01-01

    Nuclear energy continues to play a major role in the world's energy economy. Research and test reactors are an important component of a nation's nuclear power infrastructure as they provide training, experiments and operating experience vital to developing and sustaining the industry. Indeed, nations with aspirations for nuclear power development usually begin their programs with a research reactor program. Research reactors also are vital to international science and technology development. It is important to keep them safe from both accident and sabotage, not only because of our obligation to prevent human and environmental consequence but also to prevent corresponding damage to science and industry. For example, an incident at a research reactor could cause a political and public backlash that would do irreparable harm to national nuclear programs. Following the accidents at Three Mile Island and Chernobyl, considerable efforts and resources were committed to improving the safety posture of the world's nuclear power plants. Unsafe operation of research reactors will have an amplifying effect throughout a country or region's entire nuclear programs due to political, economic and nuclear infrastructure consequences. (author)

  3. 3. International conference on catalysis in membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

  4. Gamma dose from activation of internal shields in IRIS reactor.

    Science.gov (United States)

    Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

  5. Gamma dose from activation of internal shields in IRIS reactor

    International Nuclear Information System (INIS)

    Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

  6. International Reactor Physics Experiment Evaluation (IRPhE) Project

    International Nuclear Information System (INIS)

    2013-01-01

    The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

  7. Calculation of the absolute detection efficiency of a moderated 235U neutron detector on the tokamak fusion test reactor

    Science.gov (United States)

    Ku, L. P.; Hendel, H. W.; Liew, S. L.

    1989-08-01

    Neutron transport simulations have been carried out to calculate the absolute detection efficiency of a moderated 235U neutron detector which is used on the TFTR as part of the primary fission detector diagnostic system for measuring fusion power yields. Transport simulations provide a means by which the effects of variations in various shielding and geometrical parameters can be explored. These effects are difficult to study in calibration experiments. The calculational model, benchmarked against measurements, can be used to complement future detector calibrations, when the high level of radioactivity resulting from machine operation may severely restrict access to the tokamak. We present a coupled forward-adjoint algorithm, employing both the deterministic and Monte Carlo sampling methods, to model the neutron transport in the complex tokamak and detector geometries. Sensitivities of the detector response to the major and minor radii, and angular anisotropy of the neutron emission are discussed. A semiempirical model based on matching the calculational results with a small set of experiments produces good agreement (± 15%) for a wide range of source energies and geometries.

  8. International benchmark on the natural convection test in Phenix reactor

    International Nuclear Information System (INIS)

    Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

    2013-01-01

    Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

  9. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

    2017-11-15

    One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

  10. Steady-state plasma and reactor parameters for elliptical cross section tokamaks with very large power ratings

    International Nuclear Information System (INIS)

    Usher, J.L.; Powell, J.R.

    1975-06-01

    In previous studies only circular cross section reactor plasmas were considered. The purpose of this research is to examine the effects of elliptical plasma cross sections. Several technological benefits have been determined. Maximum magnetic field strength requirements are 30 to 65 percent less than for 5000 MW (th) reactors and may be as much as 40 percent less than for circular cross section reactors of comparable size. Very large n tau values are found (10 15 to 10 17 sec/cm 3 ), which produce large burn-up fractions (15 to 60 percent). There is relatively little problem with impurity build-up. Long confinement times (60 to 500 seconds) are found. Finally, the elliptical cross section reactors exhibit a major toroidal radius reduction of as large as 30 percent over circular reactors operating at comparable power levels

  11. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR

  12. Time evolution of the energy confinement time, internal inductance and effective edge safety factor on IR-T1 tokamak

    International Nuclear Information System (INIS)

    Salar Elahi, A; Ghoranneviss, M

    2010-01-01

    An attempt is made to investigate the time evolution of the energy confinement time, internal inductance and effective edge safety factor on IR-T1 tokamak. For this purpose, four magnetic pickup coils were designed, constructed and installed on the outer surface of the IR-T1 and then the Shafranov parameter (asymmetry factor) was obtained from them. On the other hand, also a diamagnetic loop was designed and installed on IR-T1 and poloidal beta was determined from it. Therefore, the internal inductance and effective edge safety factor were measured. Also, the time evolution of the energy confinement time was measured using the diamagnetic loop. Experimental results on IR-T1 show that the maximum energy confinement time (which corresponds to minimum collisions, minimum microinstabilities and minimum transport) is at low values of the effective edge safety factor (2.5 eff (a) i <0.72). The results obtained are in agreement with those obtained with the theoretical approach [1-5].

  13. Electromagnetic simulations of tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Michael; Mishchenko, Alexey [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Assoziation, Wendelsteinstrasse 1, 17491 Greifswald (Germany)

    2014-07-01

    A practical fusion reactor will require a plasma β of around 5%. In this range Alfvenic effects become important. Since a practical reactor will also produce energetic alpha particles, the interaction between Alfvenic instabilities and fast ions is of particular interest. We have developed a fluid electron, kinetic ion hybrid model that can be used to study this problem. Compared to fully gyrokinetic electromagnetic codes, hybrid codes offer faster running times and greater flexibility, at the cost of reduced completeness. The model has been successfully verified against the worldwide ITPA Toroidal Alfven Eigenmode (TAE) benchmark, and the ideal MHD code CKA for the internal kink mode in a tokamak. Use of the model can now be turned toward cases of practical relevance. Current work focuses on simulating fishbones in a tokamak geometry, which may be of relevance to ITER, and producing the first non-perturbative self-consistent simulations of TAE in a stellarator, which may be of relevance both to Wendelstein 7-X and any future stellarator reactor. Preliminary results of these studies are presented.

  14. Tritium Experience in Large Tokamaks: Application to ITER

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D; Gentile, C.; Federici, G.; Haanges, R.

    1998-05-01

    Recent experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports on an IEA workshop that brought together scientists and engineers to share experience and expertise on all fusion-related tritium issues. Extensive discussion periods were devoted to exploring outstanding issues and identifying potential R ampersand D avenues to address them. This paper summarizes the presentations, discussions, and recommendations

  15. Advanced commercial Tokamak optimization studies

    International Nuclear Information System (INIS)

    Whitley, R.H.; Berwald, D.H.; Gordon, J.D.

    1985-01-01

    Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units

  16. 3D Simulation of a Loss of Vacuum Accident (LOVA in ITER (International Thermonuclear Experimental Reactor: Evaluation of Static Pressure, Mach Number, and Friction Velocity

    Directory of Open Access Journals (Sweden)

    Jean-François Ciparisse

    2018-04-01

    Full Text Available ITER (International Thermonuclear Experimental Reactor is a magnetically confined plasma nuclear reactor. Inside it, due to plasma disruptions, the formation of neutron-activated powders, which are essentially made out of tungsten and beryllium, occurs. As many windows for diagnostics are present on the reactor, which operates at very low pressure, a LOVA (Loss of Vacuum Accident could be possible and may lead to dust mobilisation and a toxic and radioactive fallout inside the plant. This study is aimed at reproducing numerically the first seconds of a LOVA in ITER, in order to get information about the dust resuspension risk. This work has been carried out by means of a CFD (Computational Fluid Dynamics simulation of the beginning of the pressurisation transient inside the whole Tokamak. It has been found that the pressurization transient is extremely slow, and that the friction speed on the walls is very high, and therefore a high mobilization risk of the dust is expected on the entire internal surface of the reactor. It has been observed that a LOVA in a real-scale reactor is more severe than the one reproduced in reduced-scale facilities, as STARDUST-U, because the speeds are higher, and the dust resuspension capacity of the flow is greater.

  17. Inspection and repair of reactor pressure vessel (RPV) internals

    International Nuclear Information System (INIS)

    Bohmann, W.; Poetz, F.; Nicolai, M.

    1996-01-01

    The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (author)

  18. 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

  19. International project GT-MHR - New generation of nuclear reactors

    International Nuclear Information System (INIS)

    Vasyaev, A.; Kodochigov, N.; Kuzavkov, N.; Kuznetsov, L.

    2001-01-01

    Gas turbine-modular helium reactor (GT-MHR) is the reactor of new generation, which satisfies the requirements of the progressing large-scale nuclear power engineering. The activities in GT-MHR Project started in 1995. In 1997 the Conceptual Design was developed under four-side Agreement (MINATOM, General Atomics, FRAMATOME, Fuji Electric); it has passed through the internal and international reviews, has been approved and recommended for further development as one of new trends in creation of new generation plants. Starting from 1999, the activities in the development of the Preliminary Design of the plant were deployed under the Agreement between the Government of the United States of America and the Government of the Russian Federation on Scientific and Technical Cooperation in the Management of Plutonium That Has Been Withdrawn From Nuclear Military Programs dated July 24, 1998. The activities are established under the Contract between MINATOM and OKBM Russia, and under the General Agreement between Department of Energy (DOE), USA and OKBM. The GT-MHR Project is included into 'Development Strategy of Russian Nuclear Power in the first Half of the XXI-st Century' providing for 'the participation in an international project on the development and construction of GT-MHR nuclear power plant till year 2010 and 'operation of GT-MHR prototype unit and creation of fuel fabrication facility (within framework of International Project) till year 2030'. (author)

  20. 8th International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Leotta, G G; Muon-catalyzed fusion and fusion with polarized nuclei

    1988-01-01

    The International School of Fusion Reactor Technology started its courses 15 years ago and since then has mantained a biennial pace. Generally, each course has developed the subject which was announced in advance at the closing of the previous course. The subject to which the present proceedings refer was chosen in violation of that rule so as to satisfy the recent and diffuse interest in cold fusion among the main European laboratories involved in controlled thermonuclear research (CTR). In the second half of 1986 we started to prepare a workshop aimed at assessing the state of the art and possibly of the perspectives of muon- catalyzed fusion. Research in this field has recently produced exciting experimental results open to important practical applications. We thought it worthwhile to consider also the beneficial effects and problems of the polarization ofthe nuclei in both cold and thermonuclear fusion. In preparing the 8th Course on Fusion Reactor Technology, it was necessary to abandon the tradi...

  1. Magnet systems for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Miller, J.R.

    1988-01-01

    The definition phase for the International Thermonuclear Experimental Reactor (ITER) has been nearly completed, thus beginning a three-year design effort by teams from the European Community (EC), Japan, US, and USSR. Preliminary parameters for the superconducting magnet system have been established to guide more detailed design work. Radiation tolerance of the superconductors and insulators has been important because it sets requirements for the neutron-shield dimension and sensitively influences reactor size. Major levels of mechanical stress appear in the structural cases of the inboard legs of the toroidal-field (TF) coils. The winding packs of the TF coils include significant fractions of steel that provide support against in-plane separating loads, but they offer little support against out-of-plane loads unless shear-bonding of the conductors can be maintained. Heat removal from nuclear and ac loads has not limited the fundamental design, but it has nonnegligible economic consequences. 3 refs., 3 figs., 5 tabs

  2. IAEA designated international centre based on research reactors (ICERR)

    Energy Technology Data Exchange (ETDEWEB)

    Di Tigliole, Andrea Borio; Bradley, Edward; Khoroshev, Mikhail; Marshall, Frances; Morris, Charles; Tozser, Sandor [International Atomic Energy Agency, Vienna (Austria). Dept. of Nuclear Energy

    2016-04-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. These programmes will soon have achieved their goals. However, the needs of the nuclear community dictate that the majority of the research reactors continues to operate using low enriched uranium (LEU) fuel in order to meet the varied mission objectives. As a result, inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. In view of this fact, the IAEA drew up a report presenting available reprocessing and recycling services for RR SNF.

  3. Magnet systems for the International Thermonuclear Experimental Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henning, C.D.; Miller, J.R.

    1988-09-22

    The definition phase for the International Thermonuclear Experimental Reactor (ITER) has been nearly completed, thus beginning a three-year design effort by teams from the European Community (EC), Japan, US, and USSR. Preliminary parameters for the superconducting magnet system have been established to guide more detailed design work. Radiation tolerance of the superconductors and insulators has been important because it sets requirements for the neutron-shield dimension and sensitively influences reactor size. Major levels of mechanical stress appear in the structural cases of the inboard legs of the toroidal-field (TF) coils. The winding packs of the TF coils include significant fractions of steel that provide support against in-plane separating loads, but they offer little support against out-of-plane loads unless shear-bonding of the conductors can be maintained. Heat removal from nuclear and ac loads has not limited the fundamental design, but it has nonnegligible economic consequences. 3 refs., 3 figs., 5 tabs.

  4. Erosion/redeposition analysis of the ITER [International Tokamak Engineering Reactor] divertor

    International Nuclear Information System (INIS)

    Brooks, J.N.

    1989-07-01

    Sputtering erosion of the proposed ITER divertor has been analyzed using the REDEP computer code. A carbon coated plate at medium and low plasma edge temperatures, as well as beryllium and tungsten plates, have been examined. Peak net erosion rates for C and Be are very high (∼20--80 cm/burn·yr) though an order of magnitude less than the gross rates. Tritium buildup rates in codeposited carbon surface layers may also be high (∼50--250 kg/burn·yr). Plasma contamination, however, from divertor sputtering is low (approx lt.5%). Operation with low Z divertor plates, at high duty factors, therefore appears unacceptable due to erosion, but may work for low duty factor (∼2%) ''physics phase'' operation. Sweeping of the poloidal field lines at the divertor can reduce erosion, by typical factors of ∼2--8. A tungsten coated plate works well, from the erosion standpoint, for plasma plate temperatures of ∼40 eV or less. 18 refs., 11 figs., 3 tabs

  5. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

    Chan, V.S.

    1994-09-01

    In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

  6. International centres of excellence based on research reactors

    International Nuclear Information System (INIS)

    Alldred, K.; Tozser, S.M.; Adelfang, P.

    2013-01-01

    A number of high flux research reactors were, or will be constructed. Each of these high flux facilities has the potential to be an important regional or International Centre of Excellence based on Research Reactors (ICERR) and scientific hub for research and materials investigations. Some are so organized currently, but for many there is a strongly national focus and scope for a significant expansion of their international role. There are manifold benefits of an expanded international role both for the ICERR's themselves and for the institutes that affiliate with them. These benefits include increased utilization and financial stability, increased international prestige, and enhanced scientific resources and capabilities. There are significant hurdles to obtaining the benefits from an expanded international role. For example, to achieve its full potential an ICERR must accommodate scientists from other nations, and include the plans and aspirations of the international community in the ICERR governance. The ICERR must also fully meet the national responsibilities for safety and security. Balancing these potentially conflicting requirements and finding a path through the organisational and legal issues is a significant challenge for any institute. The existing ICERR's therefore provide important case studies and examples of best practice that could inform the actions of other potential ICERR's. This paper describes an IAEA initiative to encourage and support the formation of new ICERR's, strengthen existing ones, and increase training resources available to Member States. The initiative will seek to share best practice and facilitate meetings and technical exchanges between the existing and potential ICERRs, and between the potential ICERR's and potential subscribing or affiliating institutes. (orig.)

  7. International Centers of Excellence based on Research Reactors

    International Nuclear Information System (INIS)

    Alldred, K.; Tozser, S. M.; Adelfang, P.

    2012-01-01

    A number of high flux research reactors were, or will be constructed. Each of these high flux facilities has the potential to be an important regional or International Centre of Excellence based on Research Reactors (ICERR) and scientific hub for research and materials investigations. Some are so organized currently, but for many there is a strongly national focus and scope for a significant expansion of their international role. There are manifold benefits of an expanded international role both for the ICERR's themselves and for the institutes that affiliate with them. These benefits include increased utilization and financial stability, increased international prestige, and enhanced scientific resources and capabilities. There are significant hurdles to obtaining the benefits from an expanded international role. For example, to achieve its full potential an ICERR must accommodate scientists from other nations, and include the plans and aspirations of the international community in the ICERR governance. The ICERR must also fully meet the national responsibilities for safety and security. Balancing these potentially conflicting requirements and finding a path through the organisational and legal issues is a significant challenge for any institute. The existing ICERR's therefore provide important case studies and examples of best practice that could inform the actions of other potential ICERR's. This paper describes an IAEA initiative to encourage and support the formation of new ICERR's, strengthen existing ones, and increase training resources available to Member States. The initiative will seek to share best practice and facilitate meetings and technical exchanges between the existing and potential ICERRs, and between the potential ICERR's and potential subscribing or affiliating institutes. (authors)

  8. Upgraded prototype-reactor internal pump for ABWR

    International Nuclear Information System (INIS)

    Kumagai, Mikio; Amemori, Shiro; Saito, Takehiko

    1988-01-01

    In 1983, Toshiba, using their own technology, manufactured a commercial grade reactor internal pump (RIP). Recently, however, a licensing agreement with KSB of West Germany covering the RIP technology, has combined the know-how of KSB with Toshiba's technology to produce a truly high-quality prototype RIP. The pump produces the required coreflow for ABWR at low speed and with high efficiency, and simply by increasing the pump speed to the prior level, the coreflow can be further increased for such advantages as improved fuel cycle economy. Here, the advanced features and test results of the RIP are summarized. (author)

  9. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  10. 'Wall Lapping Plasma' with rotating helical resonant islands for impurity control and mechanical valves for ash exhaust in a reactor-grade tokamak without a divertor

    International Nuclear Information System (INIS)

    Tazima, Teruhiko; Sugihara, Masayoshi

    1979-09-01

    An alternative conception of the divertor, called ''Wall Lapping Plasma'' is proposed for impurity control and ash exhaust which are one of the most serious problems in reactor-grade tokamaks. Resonant helical islands formed in the boundary region rotate when we add rotating helical field by two sets of external helical coils whose current changes alternately. Consequently the plasna surface in contact with the wall by the islands rotates along the whole wall surface, so that the plasma contamination by evaporation of wall surfaces due to local heat deposition can be avoided. Plasma particles flow along the magnetic force lines intersecting the wall by islands. Intersecting angle is very small, so that mechanical valves with small height of opening located on the wall can exhaust ash easily, since backflow of neutralized helium is small because of the narrow opening. The necessary helical field is only 1/500 of the toroidal magnetic field, the total valve area is less than several percent of the wall surface area: besides the valves are easily repairable. ''Wall Lapping Plasma'' will be interesting as an alternative of the divertor because of the simple technology. (author)

  11. The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

    2012-01-01

    This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

  12. Process and apparatus for adjusting a new upper reactor internals in a reactor vessel of a PWR

    International Nuclear Information System (INIS)

    Frizot, A.; Cadaureille, G.; Lalere, C.; Machuron, J.Y.

    1987-01-01

    On the new upper reactor internals is mounted devices for alignment and clearances, before introducing in the reactor vessel. After introducing alignment and clearances are measured. Adjustment pieces are provided for optimum clearances and alignment and fixed after removal from vessel. Decontamination is made by using water jets prior to fitting recess parts [fr

  13. Fusion reactor development: A review

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This paper is a review of the current prospects for fusion reactor development based upon the present status in plasma physics research, fusion technology development and reactor conceptual design for the tokamak magnetic confinement concept. Recent advances in tokamak plasma research and fusion technology development are summarized. The direction and conclusions of tokamak reactor conceptual design are discussed. The status of alternate magnetic confinement concept research is reviewed briefly. A feasible timetable for the development of fusion reactors is presented

  14. Functional model and general principles of structural materials selection for fusion tokamak reactor first wall and blanket

    International Nuclear Information System (INIS)

    Vinokurov, V.F.; Glukhikh, V.A.; Gorynin, I.V.; Kazantsev, A.N.; Parshin, A.M.; Saksaganskij, G.L.

    1987-01-01

    The functional mathematical model of the most energy-stressed components of the reactor (the first wall and blanket) is developed. This model is the basis for the quantitative estimate of various concepts and design options from promise view points, physical and technological requirements may be formulated in general and limiting properties may be found. In this way the first wall and blanket may be directly introduced within the framework of multi-parametric system analysis and fusion reactor optimization. The concept of base structure is suggested as a generalized representation of ''the first-wall-blanket structure'' and its functions. A set of structural, thermophysical and hydraulic parameters is introduced to describe the base structure. The calculated model is presented, permitting by variation of structural and physical parameters to determine technical and economic reactor parameters as the functions of plasma parameters and structural material properties. The mechanisms of fusion reactor destructive effects on structural materials characteristics are analyzed. Comparative estimates of promising steels and alloys of various classes and modifications are given. The effect of neutron irradiation dose and temperature on strength, plasticity and swelling of materials and their compatibility with gas and liquid metal coolants are shown. Structural, physical and technological methods to improve operating characteristics and to raise radiation material resistance are discussed. (author). 7 refs, 11 figs, 2 tabs

  15. Progress toward international agreement to improve reactor safety

    International Nuclear Information System (INIS)

    Lieberman, J.I.; Graham, B.

    1993-01-01

    Representatives of nearly one-half of the 114 member states of the International Atomic Energy Agency (IAEA), including the United States, have participated in the development of an international nuclear safety conventions proposed multilateral treaty to improve civil nuclear power reactor safety. A preliminary draft of the convention has been developed (referred to as the draft convention for this report), but discussions are continuing, and when the final convention text will be completed and presented to IAEA member states for signature is uncertain. This report responds to the former and current Chairman's request that we provide information on the development of the nuclear safety convention, including a discussion of (1) the draft convention's scope and objectives, (2) how the convention will be implemented and monitored, (3) the views of selected country representatives on what provisions should be included in the draft convention, and (4) the convention's potential benefits and limitations

  16. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  17. Computer predictions for future Tokamaks

    International Nuclear Information System (INIS)

    Duechs, D.F.

    1978-01-01

    Proceeding from a reasonable agreement with existing experimental results, this lecture presents radial particle and energy transport computations which extrapolate to large (up to reactor dimensions) future Tokamaks. Special consideration is given to the behavior of alpha-particles, the influence of high-z impurities, and the thermal stability of the plasma

  18. Development of a cross-polarization scattering system for the measurement of internal magnetic fluctuations in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, T. L., E-mail: trhodes@ucla.edu; Peebles, W. A.; Crocker, N. A.; Nguyen, X. [Physics and Astronomy Department, University of California, Los Angeles, California 90098 (United States)

    2014-11-15

    The design and performance of a new cross-polarization scattering (CPS) system for the localized measurement of internal magnetic fluctuations is presented. CPS is a process whereby magnetic fluctuations scatter incident electromagnetic radiation into a perpendicular polarization which is subsequently detected. A new CPS design that incorporates a unique scattering geometry was laboratory tested, optimized, and installed on the DIII-D tokamak. Plasma tests of signal-to-noise, polarization purity, and frequency response indicate proper functioning of the system. CPS data show interesting features related to internal MHD perturbations known as sawteeth that are not observed on density fluctuations.

  19. The physics design of the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Batchelor, D.B.; Bialek, J.

    1994-01-01

    The physics approaches to improved, steady-state tokamak reactors, as evolved through reactor design studies, ideas based on experimental results, and better theoretical understanding, are the foundation for the mission and physics design of the Tokamak Physics Experiment (TPX). The mission of TPX is to develop the scientific basis for cost-competitive, continuously operating tokamak power plants. We report here the design status of TPX, a device optimized to achieve improved performance through strong plasma shaping, recycling control, and current profile shaping, while operating continuously. The design incorporates poloidal field flexibility for a wide range of operation in normalized beta and internal inductance, a double-null ''Vee'' divertor configuration for power and particle control, internal and external n ≠ 0 coils, as well as passive stabilizers, for control of MHD activity, and remote maintenance for continuous high-power operation in deuterium. Having superconducting poloidal and toroidal coils, the TPX device itself is capable of continuous operation, although initially auxiliary equipment limits the pulse length to 1000 sec

  20. Impact of an integrated core/SOL description on the R and B T optimization of tokamak fusion reactors

    Science.gov (United States)

    Siccinio, M.; Fable, E.; Angioni, C.; Saarelma, S.; Scarabosio, A.; Zohm, H.

    2018-01-01

    An updated and improved version of the 0D divertor and scrape-off layer (SOL) model published in Siccinio et al (2016 Plasma Phys. Control. Fusion 58 125011) was coupled with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Pereverzev and Yushmanov 2002 IPP Report 5/98 and Fable et al 2013 Plasma Phys. Control. Fusion 55 124028). The resulting numerical tool was employed for various scans in the major radius R and in the toroidal magnetic field B T—for different safety factors q, allowable loop voltages V loop and H factors—in order to identify the most convenient choices for an electricity producing tokamak. Such a scenario analysis was carried out evaluating self-consistently, and simultaneously, the core profile and transport effects, which significantly impact on the fusion power outcome, and the divertor heat loads, which represent one of the most critical issues in view of the realization of fusion power plants (Zohm et al 2013 Nucl. Fusion 53 073019 and Wenninger et al 2017 Nucl. Fusion 57 046002). The main result is that, when divertor limits are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, and a maximum achievable power exists—i.e. no benefits would be obtained from a further increase in R and B T once the optimum is reached. This result appears as an intrinsic physical limit for all those devices where a radiative SOL is needed to deal with the power exhaust, and where a lower limit on the power crossing the separatrix (e.g. because of the L–H transition) is present.

  1. Importance of the fine structure in a tokamak for the abnormal transport and the internal disruptions; Importance de la structure magnetique fine dans un Tokamak pour le transport anormal et les disruptions internes

    Energy Technology Data Exchange (ETDEWEB)

    Sabot, R.

    1996-02-28

    The problem of energy transport in a Tokamak, in presence of magnetic islets, has been treated by decomposing this problem in different bricks. To assembly the different bricks the model of dynamic percolation, which couples by the intermediate of scattering coefficient, the activity of transport sites (islets size) to the profile of transported quantity (temperature profile) has been chosen. The results, got with this model, results connected to the hypothesis of a limited number of islets, agree with the different observations. A possible application of this model could be the exploration of different operating conditions of Tokamak and a research of improved confinement running. (N.C.). 149 refs., 85 figs.

  2. Conceptual design of advanced steady-state tokamak reactor (A-SSTR2) - Compact and safety oriented commercial power plant

    International Nuclear Information System (INIS)

    Nishio, S.; Ushigusa, K.; Ueda, S.

    2001-01-01

    Based on the last decade JAERI reactor design studies, the advanced commercial reactor concept (A-SSTR2) which meets both economical and environmental requirements has been proposed. The A-SSTR2 is a compact power reactor (R p =6.2m, a p =1.5m, I p =12MA) with a high fusion power (P f =4GW) and a net thermal efficiency of 51%. The machine configuration is simplified by eliminating a center solenoid (CS) coil system. SiC/SiC composite for blanket structure material, helium gas cooling with pressure of 10MPa and outlet temperature of 900 deg. C, and TiH 2 for bulk shield material are introduced. For the toroidal field (TF) coil, a high temperature (T C ) superconducting wire made of bismuth with the maximum field of 23T and the critical current density of 1000A/mm 2 at a temperature of 20K is applied. In spite of the CS-less configuration, a computer simulation gives a satisfactory plasma equilibria, plasma initiation process and current ramp up scenario. (author)

  3. 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

  4. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report, January 1, 1977--September 30, 1977

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1977-01-01

    The tokamak engineering test reactor report was completed and the work is described. Studies of a small, high power density tokamak was started and very early progress is summarized. The problems of rf heating in tokamak reactors are discussed

  5. Nuclear fusion research at Tokamak Energy Ltd

    International Nuclear Information System (INIS)

    Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

    2017-01-01

    Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

  6. International project on innovative nuclear reactors and fuel cycles

    International Nuclear Information System (INIS)

    Mourogov, V. M.; Juhn, P. E.

    2003-01-01

    In response to two IAEA General Conference Resolutions in September 2000, the IAEA has launched the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) in May 2001. As of February 2003, 12 IAEA Member States and the European Commission have become members of INPRO. In total, 19 cost-free experts have been nominated by these Member States and the European Commission to work for the INPRO project at the IAEA. Four meetings of the INPRO Steering Committee (SC), which is the decision and review body of INPRO, were held, two in 2001 and another two in 2002. The objective of INPRO, which is composed of two phases (Phase 1 and Phase 2), is to support safe, economic and proliferation resistant use of nuclear technology, in a sustainable manner, to meet the global energy needs in the next 50 years and beyond. During Phase 1, work is also subdivided in two sub phases: The currently on-going Phase 1A is focussing on the selection of criteria and development of methodologies and guidelines for the comparison of different reactor and fuel cycle concepts and approaches, taking into account the compilation and review of such concepts and approaches, and determination of user requirements in the areas of economics; environment; safety; proliferation-resistance; and cross cutting issues. The preliminary results of Phase 1A with respect to user requirements are summarized in the paper

  7. IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

    International Nuclear Information System (INIS)

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

  8. Creating an international forum - The success story of the International Forum for Reactor Aging Management (IFRAM)

    International Nuclear Information System (INIS)

    Brenchley, D.L.; Bond, L.J.; Carpenter, C.E.; Hwang, I.S.; Martin, O.; Reister, R.; Shoji, T.; Tilley, R.

    2012-01-01

    As the benefits of extending the safe operating life of nuclear power plants has become more evident, so has the need to increase international cooperation in reactor aging management research. This paper describes how individuals and organizations from Asia, Europe, and North America teamed to create the International Forum for Reactor Aging Management (IFRAM). The mission of IFRAM is to facilitate the appropriate exchange of information among those parties and organizations around the world that are presently, or are planning to, address issues of nuclear power plant (NPP) systems, structures and components aging management. The main purpose of this paper is to describe the steps to success in creating IFRAM. The desired attributes, charter, and operational methods of IFRAM are described and the key contributors for creating IFRAM are identified. (author)

  9. Fast nuclear reactors. Associated international projects. State of the art and assessment of the concepts

    International Nuclear Information System (INIS)

    Azpitarte, O.; Ramilo, L.

    2013-01-01

    The recognition of the strategic importance of nuclear energy as a source of sustainable energy may be perceived in the continuous development, in many countries, of the technology of fast nuclear reactors with an associated closed fuel cycle, assuming that these Generation IV innovative systems will be required in the future. These reactors fulfill international requirements for safety and reliability, economic competitiveness, sustainability and proliferation resistance. They have the potential of using more efficiently the natural resources of Uranium and of reducing the volume and radiotoxicity of the nuclear waste by partitioning and transmutation of Minor Actinides. The national and international programs being carried out today are concentrated in the following concepts: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR), Super Critical Water Reactor (SCWR) and Molten Salt Reactor (MSR). This article presents a short review of the technology of the mentioned concepts and details the current state of the main national and international related projects. (author)

  10. Coatings for fusion reactor environments

    International Nuclear Information System (INIS)

    Mattox, D.M.

    1979-01-01

    The internal surfaces of a tokamak fusion reactor control the impurity injection and gas recycling into the fusion plasma. Coating of internal surfaces may provide a desirable and possibly necessary design flexibility for achieving the temperatures, ion densities and containment times necessary for net energy production from fusion reactions to take place. In this paper the reactor environments seen by various componentare reviewed along with possible materials responses. Characteristics of coating-substrate systems, important to fusion applications, are delineated and the present status of coating development for fusion applications is reviewed. Coating development for fusion applications is just beginning and poses a unique and important challenge for materials development

  11. Modelling of tokamak plasmas with internal transport barriers using ion temperature gradient based models

    International Nuclear Information System (INIS)

    Kinsey, J.E.; Waltz, R.E.; Staebler, G.M.; St.John, H.

    1999-01-01

    The dynamic formation of an internal transport barrier resulting from an ExB driven bifurcation has been demonstrated in simulations of a DIII-D NCS discharge with an L-mode edge using the GLF23 model. Taking the sources, sinks, equilibrium, and the density profile from a power balance analysis, the thermal and toroidal momentum transport were simultaneously evolved while computing the effects of rotational shear stabilization. As the bifurcation point is approached, the mode growth rate and ExB shear become tangential and the profiles exhibit a dithering behavior. The input power, in particular the toroidal momentum, is sufficient to overcome the ITB threshold and a bifurcation in both the momentum and heat flux is observed. While the ion thermal and momentum transport show an significant improvement in confinement, the electron thermal transport remains anomalous due to the continued presence of the ETG mode. Comparing JET OS discharges to DIII-DNCS discharges, it is found that the ExB shear rate needs to be enhanced by a factor 2.65 in order to obtain profiles consistent with experimental data. Future work will focus on improving the robustness and efficiency of numerical techniques to allow wider application and increased predictive capability needed to conduct power threshold studies of internal transport barriers. (author)

  12. Installation technology of reactor internals on shroud replacement work

    International Nuclear Information System (INIS)

    Miyano, Hiroshi

    1999-01-01

    Since the replacement of large welded reactor internals much as a core shroud did not have a precedent in the world, quite a few technologies had to be developed. Especially for the installation of new core shroud, jet pumps, core plate and top guide, the accurate weld and fit-up techniques for large structures was required to secure their integrity. The vessel shielding system was utilized to reduce general area dose rate such that all replacement work. For jet pump installation, automatic remote welding machines were used for high radiation area. As for the core shroud, shroud support weld prep machining tool with high accuracy, jacking system to support fit-up, new weld machine for small work space and low heat input weld joint were developed. Shroud replacement work in Fukushima Dai-ichi NPS Unit 3 (1F-3) with application of these development techniques, was successfully accomplished. The technology is applied for 1F-2 replacement work also. (author)

  13. 1st International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Knoepfel, Heinz; Safety, Environmental Impact and Economic Prospects of Nuclear Fusion

    1990-01-01

    This book contains the lectures and the concluding discussion of the "Seminar on Safety, Environmental Impact, and Economic Prospects of Nuclear Fusion", which was held at Erice, August 6-12, 1989. In selecting the contributions to this 9th meeting held by the International School of Fusion Reactor Technology at the E. Majorana Center for Scientific Cul­ ture in Erice, we tried to provide a comprehensive coverage of the many interre­ lated and interdisciplinary aspects of what ultimately turns out to be the global acceptance criteria of our society with respect to controlled nuclear fusion. Consequently, this edited collection of the papers presented should provide an overview of these issues. We thus hope that this book, with its extensive subject index, will also be of interest and help to nonfusion specialists and, in general, to those who from curiosity or by assignment are required to be informed on these as­ pects of fusion energy.

  14. Deterministic Hydraulic Load Analysis on Reactor Internals of APR1400

    International Nuclear Information System (INIS)

    Kim, Kyu Hyung; Ko, Do Young; Gu, Ja Yeong

    2011-01-01

    The structural integrity of the reactor vessel internals (RVI) of the nuclear power plants that have been constructed should be verified in accordance with the US Nuclear Regulatory Commission Regulatory Guide 1.20 (RG1.20) comprehensive vibration assessment program (CVAP) during preoperational and initial startup testing. The program consists of a vibration and stress analysis, a vibration measurement, an inspection, and an assessment of each program. The vibration and stress analysis program is comprised of a hydraulic load analysis and a structural response analysis. The hydraulic loads include the random hydraulic loads induced by turbulent flow and deterministic hydraulic loads induced by pump pulsation. This paper describes a developed full scope 3-D model and the deterministic hydraulic loads for the RVI of the APR1400

  15. Computer system for International Reactor Pressure Vessel Materials Database support

    International Nuclear Information System (INIS)

    Arutyunjan, R.; Kabalevsky, S.; Kiselev, V.; Serov, A.

    1997-01-01

    This report presents description of the computer tools for support of International Reactor Pressure Vessel Materials Database developed at IAEA. Work was focused on raw, qualified, processed materials data, search, retrieval, analysis, presentation and export possibilities of data. Developed software has the following main functions: provides software tools for querying and search of any type of data in the database; provides the capability to update the existing information in the database; provides the capability to present and print selected data; provides the possibility of export on yearly basis the run-time IRPVMDB with raw, qualified and processed materials data to Database members; provides the capability to export any selected sets of raw, qualified, processed materials data

  16. IRIS International Reactor Innovative and Secure Final Technical Progress Report

    International Nuclear Information System (INIS)

    Carelli, M.D.

    2003-01-01

    OAK-B135 This NERI project, originally started as the Secure Transportable Autonomous Light Water Reactor (STAR-LW) and currently known as the International Reactor Innovative and Secure (IRIS) project, had the objective of investigating a novel type of water-cooled reactor to satisfy the Generation IV goals: fuel cycle sustainability, enhanced reliability and safety, and improved economics. The research objectives over the three-year (1999-2002) program were as follows: First year: Assess various design alternatives and establish main characteristics of a point design; Second year: Perform feasibility and engineering assessment of the selected design solutions; Third year: Complete reactor design and performance evaluation, including cost assessment These objectives were fully attained and actually they served to launch IRIS as a full fledged project for eventual commercial deployment. The program did not terminate in 2002 at the end of the NERI program, and has just entered in its fifth year. This has been made possible by the IRIS project participants which have grown from the original four member, two-countries team to the current twenty members, nine countries consortium. All the consortium members work under their own funding and it is estimated that the value of their in-kind contributions over the life of the project has been of the order of $30M. Currently, approximately 100 people worldwide are involved in the project. A very important constituency of the IRIS project is the academia: 7 universities from four countries are members of the consortium and five more US universities are associated via parallel NERI programs. To date, 97 students have worked or are working on IRIS; 59 IRIS-related graduate theses have been prepared or are in preparation, and 41 of these students have already graduated with M.S. (33) or Ph.D. (8) degrees. This ''final'' report (final only as far as the NERI program is concerned) summarizes the work performed in the first four

  17. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  18. Tokamak formation and sustainment by tokamak injection

    International Nuclear Information System (INIS)

    Farengo, R.; Jarboe, T.R.

    1991-01-01

    The authors propose here a new helicity injection method for tokamak formation and sustainment that has high efficiency, conserves toroidal symmetry and is inductively driven. The basic idea is to inject a small tokamak (source tokamak) into a larger tokamak (steady tokamak). This current drive scheme eliminates the need for the ohmic heating transformer in the steady tokamak allowing the formation of very small aspect ratio tokamaks (Spherical Tori). Thus, steady state operation and high beta can be realized simultaneously. The method can also be applied to a larger aspect ratio tokamak and used in conjunction with the standard inductive formation technique. In order to allow for translation the ohmic heating coil used to produce the source tokamaks must be fed from one end (as in the CSS device) and the toroidal field coil must link both tokamaks. After formation the source tokamaks are accelerated towards the steady tokamak by a mirror field and the tension of the field lines that wrap around both tokamaks (producing a doublet type configuration). In a tokamak the helicity is proportional to the current. This indicates that (assuming helicity is conserved during the merging process) a steady state situation will result if the helicity supplied by the source tokamaks is equal to the helicity dissipated by the steady tokamak. Assuming that source tokamaks of helicity K s are injected with frequency f, the steady state condition can be written as: fK s = 2V t Ψ t = K t /τ K where V t , Ψ t , K t and τ K are the ohmic loop voltage, toroidal flux, helicity and helicity decay time of the steady tokamak. A simple calculation shows that the DIII-D tokamak could be sustained by injecting source tokamaks with R = 1.20 m, a = 0.23 m and I = 151 kA at a frequency of 120 Hz. 1 ref

  19. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System.

  20. Tokamak burn control

    International Nuclear Information System (INIS)

    Sager, G.T.

    1988-06-01

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  1. The IRIS consortium: international cooperation in advanced reactor development

    International Nuclear Information System (INIS)

    Carelli, M.; Petrovic, B.; Miller, K.; Lombardi, C.; Ricotti, M.E.

    2005-01-01

    Besides its many outstanding technical innovations in the design and safety, the most innovative feature of the International Reactor Innovative and Secure (IRIS), is perhaps the international cooperation which carries on its development. IRIS is designed by an international consortium which currently numbers 21 organizations from ten countries across four continents. It includes reactor, fuel and fuel cycle vendors, component manufacturers, laboratories, academia, architect engineers and power producers. The defining organizational characteristics of IRIS is that while Westinghouse has overall lead and responsibility, this lead is of the type of 'primus inter pares' (first among equals) rather than the traditional owner versus suppliers/contractors relationship. All members of the IRIS consortium contribute and expect to have a return, should IRIS be successfully deployed, commensurate to their investment. The nature of such return will be tailored to the type of each organization, because it will of course be of a different nature for say a component manufacturer, university, or architect engineer. One fundamental tenet of the consortium is that all members, regardless of their amount of contribution, have equal access to all information developed within the project. Technical work is thus being coordinated by integrated subgroups and the whole team meets twice a year to perform an overall review of the work, discuss policy and strategy and plan future activities. Personnel from consortium members have performed internships, mostly at Westinghouse locations in Pittsburgh, Pennsylvania, and Windsor, Connecticut, but also at other members, as it has been the case for several graduate students. In fact, more than one hundred students at the various universities have been working on IRIS, most of them conducting graduate theses at the master or doctoral level. The IRIS experience has proved very helpful to the students in successfully landing their employment choice

  2. 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

  3. Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Bruhn, M.L.

    1988-04-01

    Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs.

  4. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  5. International project on innovative nuclear reactors and fuel cycles (INPRO)

    International Nuclear Information System (INIS)

    Omoto, A.

    2006-01-01

    The IAEA's project INPRO was initiated in order to provide a forum for discussion of experts and policy makers on all aspects of nuclear energy planning as well as on the development and deployment of innovative nuclear energy systems (INS). It brings together technology holders users and potential users to consider jointly the international and national actions required for achieving desired innovations in nuclear reactors and fuel cycles, but it pays particular attention to the needs of developing countries. Currently INPRO members count 24 including even three countries, which are not yet operating nuclear reactors. Its initial phase has produced an outlook into the future of the energy markets and defined basic principles, user requirements and criteria in the following areas as TECDOC1362 in June 2003; Economics, Environment, Fuel Cycle and Waste, Safety, Proliferation Resistance and Crosscutting Issues. This assessment methodology can be applied for screening an INS, comparing different INS to find a preferred INS consistent with the needs of a given state, and identifying RD and D needs. The methodology has be validated through case studies and updated as TECDOC1434 in December 2004. Currently, besides producing a manual for each chapter of TECDOC1434, six assessment studies of various INS options are being carried out and the number of such studies is increasing. Further several tasks are ongoing including modeling and analysis of global and regional balance of resources and INS deployment scenarios in order to gain the better perspective of future implication of INS deployment as well as to identify challenges and opportunities of INS. It is envisioned that INPRO will continue to develop with three planned major pillars of activity; methodology, infrastructure and coordination for planning of R and D activities. The paper discusses the progress and status of INPRO as well as the future prospect of INPRO activities

  6. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    1981-01-01

    Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

  7. 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.)

  8. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    1992-01-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

  9. Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    1998-04-01

    Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997

  10. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-09-01

    A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  11. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-01-01

    A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  12. ENS RRFM 2005: 9th international topical meeting on research reactor fuel management. Transactions

    International Nuclear Information System (INIS)

    2005-01-01

    The ENS topical meeting on research reactor fuel management is an annual conference launched successfully in 1997. It has since then grown into well established international forum for the exchange and expertise on all significant aspects of the nuclear fuel cycle of research reactors. Oral presentations at this meeting were divided in the following four sessions: International Topics; Fuel Development, Qualification, Fabrication and Licensing; Reactor Operation, Fuel Safety and Core Conversion; Spent Fuel Management, Back-end Options, Transportation. The three poster sessions were devoted to fuel development, qualification, fabrication and licensing; reactor operation, fuel safety, core conversion, spent fuel; spent fuel management, fuel cycle back-end options, transportation

  13. The engineering design of the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1994-01-01

    A mission and supporting physics objectives have been developed, which establishes an important role for the Tokamak Physics Experiment (TPX) in developing the physic basis for a future fusion reactor. The design of TPX include advanced physics features, such as shaping and profile control, along with the capability of operating for very long pulses. The development of the superconducting magnets, actively cooled internal hardware, and remote maintenance will be an important technology contribution to future fusion projects, such as ITER. The Conceptual Design and Management Systems for TPX have been developed and reviewed, and the project is beginning Preliminary Design. If adequately funded the construction project should be completed in the year 2000

  14. Analysis of quench-vent pressures for present design of ITER [International Thermonuclear Experimental Reactor] TF [toroidal field] coils

    International Nuclear Information System (INIS)

    Slack, D.S.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Union of the Soviet Union, and the United States. This paper examines the effects of a quench within the toroidal field (TF) coils based on current ITER design. It is a preliminary, rough analysis. Its intent is to assist ITER designers while more accurate computer codes are being developed and to provide a check against these more rigorous solutions. Rigorous solutions to the quench problem are very complex involving three-dimensional heat transfer, extreme changes in heat capacities and copper resistivity, and varying flow dynamics within the conductors. This analysis addresses all these factors in an approximate way. The result is much less accurate than a rigorous analysis. Results here could be in error as much as 30 to 40 percent. However, it is believed that this paper can still be very useful to the coil designer. Coil pressures and temperatures vs time into a quench are presented. Rate of helium vent, energy deposition in the coil, and depletion of magnetic stored energy are also presented. Peak pressures are high (about 43 MPa). This is due to the very long vent path length (446 m), small hydraulic diameters, and high current densities associated with ITER's cable-in-conduit design. The effects of these pressures as well as the ability of the coil to be self protecting during a quench are discussed. 3 refs., 3 figs., 1 tab

  15. Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

    International Nuclear Information System (INIS)

    Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

    2012-01-01

    With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

  16. Mathematical modeling plasma transport in tokamaks

    International Nuclear Information System (INIS)

    Quiang, Ji

    1995-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%

  17. Proceedings of the international symposium on research reactor safety operations and modifications

    International Nuclear Information System (INIS)

    1990-03-01

    The International Symposium on Research Reactor Safety, Operations and Modifications was organized by the International Atomic Energy Agency in cooperation with Atomic Energy of Canada Limited-Research Company. The main objectives of this Symposium were: (1) to exchange information and to discuss current perspectives and concerns relating to all aspects to research reactor safety, operations, and modifications; and, (2) to present views and to discuss future initiatives and directions for research reactor design, operations, utilization, and safety. The symposium topics included: research reactor programmes and experience; research reactor design safety and analysis; research reactor modifications and decommissioning; research reactor licensing; and new research reactors. These topics were covered during eight oral sessions and three poster sessions. These Proceedings include the full text of the 93 papers presented. The subject of Symposium was quite wide-ranging in that it covered essentially all aspects of research reactor safety, operations, and modifications. This was considered to be appropriate and timely given the 326 research reactors currently in operation in some 56 countries; given the degree of their utilization which ranges from pure and applied research to radioisotopes production to basic training and manpower development; and given that many of these reactors are undergoing extensive modifications, core conversions, power upratings, and are becoming the subject of safety reassessment and regulatory reviews. Although the Symposium covered many topics, the majority of papers and discussions tended to focus mainly on research reactor safety. This was seen as a clear sign of the continuing recognition of the fundamental importance of identifying and addressing, particularly through international cooperation, issues and concerns associated with research reactor safety

  18. Physical model study of neutron noise induced by vibration of reactor internals

    International Nuclear Information System (INIS)

    Liu Jinhui; Gu Fangyu

    1999-01-01

    The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

  19. Tokamak advanced pump limiter experiments and analysis

    International Nuclear Information System (INIS)

    Conn, R.W.

    1983-06-01

    Experiments with pump limiter modules on several operating tokamaks establish such limiters as efficient collectors of particles and has demonstrated the importance of ballistic scattering as predicted theoretically. Plasma interaction with recycling neutral gas appears to become important as the plasma density increases and the effective ionization mean free path within the module decreases. In limiters with particle collection but without active internal pumping, the neutral gas pressure is found to vary nonlinearly with the edge plasma density at the highest densities studies. Both experiments and theory indicate that the energy spectrum of gas atoms in the pump ducting is non-thermal, consistent with the results of Monte Carlo neutral atom transport calculations. The distribution of plasma power over the front surface of such modules has been measured and appears to be consistent with the predictions of simple theory. Initial results from the latest experiment on the ISX-B tokamak with an actively pumped limiter module demonstrates that the core plasma density can be controlled with a pump limiter and that the scrape-off layer plasma can partially screen the core plasma from gas injection. The results from module pump limiter experiments and from the theory and design analysis of advanced pump limiters for reactors are used to suggest the major features of a definitive, axisymmetric, toroidal belt pump limiter experiment

  20. CAT-D-T tokamaks

    International Nuclear Information System (INIS)

    Greenspan, E.; Blue, T.; Miley, G.H.

    1981-01-01

    The domains of plasma fuel cycles bounded by the D-T and Cat-D, and by the D-T and SCD modes of operation are examined. These domains, referred to as, respectively, the Cat-D-T and SCD-T modes of operation, are characterized by the number (γ) of tritons per fusion neutron available from external (to the plasma) sources. Two external tritium sources are considered - the blankets of the Cat-D-T (SCD-T) reactors and fission reactors supported by the Cat-D-T (SCD-T) driven hybrid reactors. It is found that by using 6 Li for the active material of the control elements of the fission reactors, it is possible to achieve γ values close to unity. Cat-D-T tokamaks could be designed to have smaller size, higher power density, lower magnetic field and even lower plasma temperature than Cat-D tokamaks; the difference becomes significant for γ greater than or equal to .75. The SCD-T mode of operation appears to be even more attractive. Promising applications identified for these Cat-D-T and SCD-T modes of operation include hybrid reactors, fusion synfuel factories and fusion reactors which have difficulty in providing all their tritium needs

  1. Lower hybrid heating experiments in tokamaks: an overview

    International Nuclear Information System (INIS)

    Porkolab, M.

    1985-10-01

    Lower hybrid wave propagation theory relevant to heating fusion grade plasmas (tokamaks) is reviewed. A brief discussion of accessibility, absorption, and toroidal ray propagation is given. The main part of the paper reviews recent results in heating experiments on tokamaks. Both electron and ion heating regimes will be discussed. The prospects of heating to high temperatures in reactor grade plasmas will be evaluated

  2. Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database.

  3. Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    1997-08-01

    This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database

  4. An investigation of structural design methodology for HTGR reactor internals with ceramic materials (Contract research)

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro

    2008-03-01

    To advance the performance and safety of HTGR, heat-resistant ceramic materials are expected to be used as reactor internals of HTGR. C/C composite and superplastic zirconia are the promising materials for this purpose. In order to use these new materials as reactor internals in HTGR, it is necessary to establish a structure design method to guarantee the structural integrity under environmental and load conditions. Therefore, C/C composite expected as reactor internals of VHTR is focused and an investigation on the structural design method applicable to the C/C composite and a basic applicability of the C/C composite to representative structures of HTGR were carried out in this report. As the results, it is found that the competing risk theory for the strength evaluation of the C/C composite is applicable to design method and C/C composite is expected to be used as reactor internals of HTGR. (author)

  5. IGORR-IV - Proceedings of the fourth meeting of the International Group on Research Reactors

    International Nuclear Information System (INIS)

    Rosenbalm, K.F.

    1995-01-01

    The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results

  6. IGORR-IV -- Proceedings of the fourth meeting of the International Group on Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbalm, K.F. [comp.

    1995-12-31

    The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results.

  7. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  8. MODAL CHARACTERISTIC ANALYSIS OF THE APR1400 NUCLEAR REACTOR INTERNALS FOR SEISMIC ANALYSIS

    OpenAIRE

    JONG-BEOM PARK; YOUNGIN CHOI; SANG-JEONG LEE; NO-CHEOL PARK; KYOUNG-SU PARK; YOUNG-PIL PARK; CHAN-IL PARK

    2014-01-01

    Reactor internals are sensitive to dynamic loads such as earthquakes and flow induced vibration. Thus, it is essential to identify the dynamic characteristics to evaluate the seismic integrity of the structures. However, a full-sized system is too large to perform modal experiments, making it difficult to extract data on its modal characteristics. In this research, we constructed a finite element model of the APR1400 reactor internals to identify their modal characteristics. The commercial re...

  9. HEU core conversion of Russian production reactors: a major threat to the international RERTR regime

    International Nuclear Information System (INIS)

    Kuperman, Alan J.; Leventhal, Paul L.

    1998-01-01

    This paper calls the attention for the major threat to the International Reduced Enrichment for Research and Test Reactors (RERTR) program, represented by the HEU core conversion of russian production reactors. This program aims to reduce and eventually eliminate international civilian commerce in nuclear weapons-usable, highly enriched uranium , and thereby significantly lower risks of the material being stolen or diverted by terrorist or states for producing nuclear weapons

  10. Overview of the US stellarator reactor study

    International Nuclear Information System (INIS)

    Lyon, J.F.; Gulec, K.; Miller, R.L.; El-Guebaly, L.

    1993-01-01

    This study, which uses a cost-minimization code that incorporates the ARIES costing and reactor component models with a I-D energy transport calculation, shows that a torsatron reactor could be competitive with a tokamak reactor

  11. International Working Group on Fast Reactors Second Annual Meeting. Summary Report

    International Nuclear Information System (INIS)

    1969-01-01

    The Agenda of the Meeting was as follows: Opening of the meeting. 2. Appraisal of the IWGFB's activity for the period from the first annual meeting of the Group. 3. Comments on national programmes on fast breeder reactors. 4. Presentation of general findings and conclusions of national and regional meetings on fast reactor problems held in represented countries and international organisations last year. 5. Comments on the programmes of international meetings on fast reactors to be held in 1969. 6. Consideration of a schedule for meetings on fast reactors in 1970. 7. Suggestions for the topics and location of specialists' meetings in 1969-1970. 8. Suggestions for reviews and studies in the field of fast reactors. 9. The time and place of the third annual meeting of the IWGFR. 10. Closing of the meeting

  12. Status of national programmes on fast reactors 1997/98. 31. annual meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    1998-01-01

    The objective of the meeting was to co-ordinate the exchange of information on the status of fast reactor development and operational experience, including experience with experimental types of reactor; to consider meeting arrangements for 1998 and 1999; and to review the IAEA co-ordinated research activities in the field of fast reactor, as well as co-ordination of the International Working Group on Fast Reactors activities with other organizations

  13. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Baker, C.C.; Roberts, M.

    1976-01-01

    Three designs of an EPR are briefly described. The plasma physics involved in the EPR design is discussed. Brief discussions on the following topics are included: (1) MHD equilibrium, (2) impurity control, (3) transport scaling, (4) burn cycle dynamics, (5) first wall system, (6) blanket/shield systems, (7) plasma heating, (8) toroidal field coils, (9) poloidal field coils, (10) energy storage and transfer, (11) tritium fuel cycle, (12) radiation damage, and (13) costs

  14. Catalyzed deuterium fueled tokamak reactors

    International Nuclear Information System (INIS)

    Southworth, F.H.

    1977-01-01

    Catalyzed deuterium fuel presents several advantages relative to D-T. These are, freedom from tritium breeding, high charged particle power fraction and lowered neutron energy deposition in the blanket. Higher temperature operation, lower power densities and increased confinement are simultaneously required. However, the present study has developed designs which have capitalized upon the advantages of catalyzed deuterium to overcome the difficulties associated with the fuel while obtaining high efficiency

  15. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

    1979-01-01

    The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

  16. Research into controlled fusion in tokamaks

    International Nuclear Information System (INIS)

    Zacek, F.

    1992-01-01

    During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

  17. Dismantling of JPDR reactor internals by underwater plasma arc cutting technique using robotic manipulator

    International Nuclear Information System (INIS)

    Yokota, M.

    1988-01-01

    The actual dismantling of JPDR started on December 4, 1986. As of now, equipment that surrounds the reactor has mostly been removed to provide working space in reactor containment prior to the dismantling of reactor internals. Some reactor internals have been successfully dismantled using the underwater arc cutting system with a robotic manipulator during the period of January to March 1988. The cutting system is composed of an underwater plasma arc cutting device and a robotic manipulator. The cut off reactor internals were core spray block, feedwater sparger and stabilizers for fuel upper grid tube. The plasma arc cutting device was developed to dismantle the reactor internals underwater. It mainly consists of a plasma torch, power and gas supply systems for the torch, and by-product treatment systems. It has the cutting ability of 130 mm thickness stainless steel underwater. The robotic manipulator has seven degrees of freedom of movement, enabling it to move in almost the same way as the arm of a human being. The arm of the robot is mounted on a supporting device which is suspended by three chains from the support structure set on a service floor. A plasma torch is griped by the robotic hand; its position to the structure to be cut is controlled from a remote control room, about 100 meters outside the reactor containment

  18. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  19. Tritium-management requirements for D-T fusion reactors (ETF, INTOR, FED)

    International Nuclear Information System (INIS)

    Finn, P.A.; Clemmer, R.G.; Misra, B.

    1981-10-01

    The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design

  20. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    OpenAIRE

    KO, DO-YOUNG; KIM, KYU-HYUNG

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

  1. Computing radiation dose to reactor pressure vessel and internals

    International Nuclear Information System (INIS)

    1996-01-01

    Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

  2. International aspects of high-temperature reactor development

    International Nuclear Information System (INIS)

    Balthesen, E.

    1985-01-01

    The development of high-temperature reactor technology is under way in the Federal Republic of Germany, the United States, the Soviet Union and Japan. Access of nuclear energy to the heat market is the common goal in the long run. HTR applications for oil recovery and refining are under investigation in the Republic of China. The German pebble-bed reactor concept is beginning to be accepted worldwide. Due to extensive development work, market maturity for the technology for electricity and process-steam production has been achieved in the Federal Republic. (orig.) [de

  3. 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)

  4. KeproVt : underwater robotic system for visual inspection of nuclear reactor internals

    International Nuclear Information System (INIS)

    Cho, Byung-Hak; Byun, Seung-Hyun; Shin, Chang-Hoon; Yang, Jang-Bum; Song, Sung-Il; Oh, Jung-Mook

    2004-01-01

    An underwater robotic system for visual inspection of reactor vessel internals has been developed. The Korea Electric Power Robot for Visual Test (KeproVt) consists of an underwater robot, a vision processor based measuring unit, a master control station and a servo control station. The vision processor based measuring unit employs a first-of-a-kind engineering technology in nuclear robotics. The vision processor makes use of a camera located at the top of the water level referenced to the reactor center line to get an image of the robot, and computes the location and orientation of the robot. The robot guided by the control station with the measuring unit can be controlled to have any motion at any position in the reactor vessel with ±1 cm positioning and ±2 deg. heading accuracies with enough precision to inspect reactor internals. A simple and fast installation process is emphasized in the developed system. The installation process consists of hooking a vision camera on the guide rail of the refueling machine and putting a small robot (14.5 kg in weight) in the reactor cavity pool. The easy installation and automatic operation meet the demand of shortening the reactor outage and reducing the number of inspection personnel. The developed robotic system was successfully deployed at the Yonggwang Nuclear Unit 1 for the visual inspection of reactor internals

  5. Elevated temperature design of KALIMER reactor internals accounting for creep and stress-rupture effects

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Yoo, Bong

    2000-01-01

    In most LMFBR (Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER (Korea Advanced Liquid Metal Reactor) reactor internal structures is carried out for normal operating conditions which have the operating temperature 530 deg. C and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME code case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects. (author)

  6. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    1999-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  7. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    2001-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  8. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    International Nuclear Information System (INIS)

    Iracane, Daniel; Bignan, Gilles; Lindbaeck, Jan-Erik; Blomgren, Jan

    2010-01-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  9. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, Daniel; Bignan, Gilles [CEA Atomic Energy Commission Saclay Batiment 121- 91191 Gif Sur Yvette (France); Lindbaeck, Jan-Erik; Blomgren, Jan [VATTENFALL AB Nuclear Power Jaemtlandsgatan 99 SE-16287 Stockholm (Sweden)

    2010-07-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  10. On Brazil's participation in the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO)

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho

    2007-01-01

    In response to a resolution of its 44th General Conference (GC(44)/RES/21) held in September 2000, the International Atomic Energy Agency launched in May 2001 the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO) with the objective of supporting the safe, sustainable, economic and proliferation-resistant use of nuclear technology to meet the global energy needs of the 21st century. Brazil joined the project from its beginnings and in 2005 submitted a proposal for the screening assessment using INPRO methodology of two small-size light-water reactors as potential components of an innovative nuclear reactor system (INS) completed with a conventional open nuclear fuel cycle. The INS reactor components currently being assessed are the International Reactor Innovative and Secure (IRIS) that is being developed by an international consortium made of 21 organizations from 10 countries (Brazil included) led by the Westinghouse Company, and the Fixed Bed Nuclear Reactor (FBNR) that is being developed at the Federal University of Rio Grande do Sul. This paper gives an overview of Brazil's participation in INPRO, highlighting the objective, scope and intermediate results of the assessment study being performed, and the possibilities for participation in one or two collaborative research projects under INPRO Phase 2 Action Plan for 2008-2009. (author)

  11. International Thermonuclear Experimental Reactor (ITER). Engineering Design Activities (EDA). Agreement and protocol 1

    International Nuclear Information System (INIS)

    1992-01-01

    This document contains protocol 1 to the agreement among the European Atomic Energy Community, the government of Japan, the Government of the Russian Federation, and the Government of the United States of America on cooperation in the engineering design activities for the International Thermonuclear Experimental Reactor, which activities shall be conducted under the auspices of the International Atomic Energy Agency

  12. A process and container for transport and storage of irradiated internal equipment from a nuclear reactor

    International Nuclear Information System (INIS)

    Dubourg, M.

    1996-01-01

    A transfer container is assembled (from at least two components) around the highly irradiated internal equipment, in the reactor pool; the entire internal equipment is then transported, as a whole, in the container to a storage zone. The storage container walls are fitted with biological protection and cooling means

  13. Load bearing capacities and elastic-plastic behavior of reactor vessel internals

    International Nuclear Information System (INIS)

    Watanabe, Keita; Nagase, Ryuichi

    2017-01-01

    Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

  14. Investigation of internal elements impaction on particles circulation in a fluidized bed reactor

    Science.gov (United States)

    Solovev, S. A.; Soloveva, O. V.; Antipin, A. V.; Shamsutdinov, E. V.

    2018-01-01

    A numerical study of the fluidized bed apparatus in the presence of various internal elements is carried out. A chemical reaction for temperature-dependent processes with heat absorption is considered. The task of incoming heated catalyst granules to the reactor is investigated. The main emphasis is focused on the circulation flows of the catalyst particles, heating of the reactor, and the efficiency of the chemical reaction. The analysis of the impact of various design elements on the efficiency of the reactor is carried out. The influence of feeding heated catalyst device design on the effectiveness of whole reactor heating is educed. The influence of the presence of fine particles on the efficiency of the reaction for different reactor design features is also educed.

  15. The OECD Halden Reactor Project - international research on safety and reliability of nuclear power generation

    International Nuclear Information System (INIS)

    Broy, Y.; Wiesenack, W.; Moen, L.A.

    2001-01-01

    The Halden Reactor was one of the first experimental reactors in the world, built into a mountain cave in the middle of the town of Halden (120 km south of Oslo). The reactor went critical for the first time in June 1959. The experimental programme is run under the auspices of the OECD Halden Reactor Project (HRP). This was established in 1958 as a joint undertaking of the OECD Nuclear Energy Agency. The Project is an international collaboration and today 21 participating countries are sponsoring the experimental programme. This paper gives an overview of the Halden Reactor and its associated experimental facilities; the current research program on fuels and materials testing, including some selected results; and finally an outline of the planned research activities in the period 2000 - 2002. (authors)

  16. MHD stability of vertically asymmetric tokamak equilibria

    International Nuclear Information System (INIS)

    Dalhed, H.E.; Grimm, R.C.; Johnson, J.L.

    1981-03-01

    The ideal MHD stability properties of a special class of vertically asymmetric tokamak equilibria are examined. The calculations confirm that no major new physical effects are introduced and the modifications can be understood by conventional arguments. The results indicate that significant departures from up-down symmetry can be tolerated before the reduction in β becomes important for reactor operation

  17. Maintenance considerations of the STARFIRE commercial tokamak

    International Nuclear Information System (INIS)

    Trachsel, C.A.; Zahn, H.S.; Field, R.E.; Stevens, H.C.

    1979-01-01

    This paper presents the maintenance approach, the commercial tokamak design features that enhance maintenance and preliminary repair time and required mean-time-between-failures for major subsystems. Reactor hall building and maintenance equipment requirements including hot cells, coil rewinding, and cranes are discussed

  18. Proceedings of 18th international conference on structural mechanics in reactor technology

    International Nuclear Information System (INIS)

    2005-07-01

    The 18th International Conference on Structural Mechanics in Reactor Technology was held on August 7-12, 2005 in Beijing, China, and Sponsored by International Association for Structural Mechanics in Reactor Technology, Chinese Nuclear Society, Chinese Society of Theoretical and Applied Mechanics, and Tsinghua University. 486 abstracts are Collected. The contents includes: opening, plenary and keynote presentations; computational mechanics; fuel and core structures; aging, life extension, and license renewal; design methods and rules for components; fracture mechanics; concrete material, containment and other structures; analysis and design for dynamic and extreme loads; seismic analysis, design and qualification; structural reliability and probabilistic safety assessment (PSA); operation, inspection and maintenance; severe accident management and structural evaluation; advanced reactors and generation IV reactors; decommissioning of nuclear facilities and waste management.

  19. International collaboration between nuclear research centres and the role of research reactors

    International Nuclear Information System (INIS)

    Dodd, B.

    2001-01-01

    A research reactor is a core facility in many nuclear research centres (NRCs) of Member States and it is logical that it should be the focus of any international collaboration between such centres. There are several large and sophisticated research reactors in operation in both developed and developing Member States, such as Belgium, China, Egypt, France, Hungary, Indonesia, India, Japan, ROK, Netherlands, South Africa and the USA. There are also several new, large reactors under construction or being planned such as those in Australia, Canada, China, France, Germany, and Thailand. It is felt that the utilization of these reactors can be enhanced by international co-operation to achieve common goals in research and applications. (author)

  20. Proceedings of the International Workshop on Boiling Water Reactor Stability

    International Nuclear Information System (INIS)

    1991-01-01

    With regard to technical understanding of the phenomena, the participants agreed that the causes of instability appear to be well understood, but there are many variables involved, and their correlation with instability conditions is not always certain. Most codes claimed to be capable of predicting oscillations and unstable conditions, based on post-test analyses of data from actual events, but there do not seem to be any blind predictions available which accurately predict an instability event before the actual test results are released. As a result, reactor owners have decided that the best course is to avoid, with sufficient margin, certain regions in the power-flow map where regions of instability are known to exist, rather than try to predict them very accurately. The meeting concluded that the safety significance of BWR instability is rather limited, and current estimates of plant risk do not show it to be a dominant contributor. This is because the installed plant protection systems will shut a reactor down when the oscillations exceed power limits, and any fuel damage which might occur will be localized and containable. However, it was also agreed that an instability event could increase uncertainties in the human error rate, because operators who have never seen an unstable reactor may take actions which are not necessarily the best for the particular situation. In addition, although an instability event may not cause any harm to the public, it may cause some fuel failures, and these are certainly a concern to a reactor owner, for economic and radiation protection reasons. Finally, it was also agreed that BWR instability is certainly considered to be significant by the public, where acceptance of the technology would erode if a plant is perceived to be in an uncontrolled state, regardless of the actual risk inherent in the situation

  1. Characterization of decommissioned reactor internals: Monte Carlo analysis technique

    International Nuclear Information System (INIS)

    Reid, B.D.; Love, E.F.; Luksic, A.T.

    1993-03-01

    This study discusses computer analysis techniques for determining activation levels of irradiated reactor component hardware to yield data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. The study recommends the Monte Carlo Neutron/Photon (MCNP) computer code as the best analysis tool for this application and compares the technique to direct sampling methodology. To implement the MCNP analysis, a computer model would be developed to reflect the geometry, material composition, and power history of an existing shutdown reactor. MCNP analysis would then be performed using the computer model, and the results would be validated by comparison to laboratory analysis results from samples taken from the shutdown reactor. The report estimates uncertainties for each step of the computational and laboratory analyses; the overall uncertainty of the MCNP results is projected to be ±35%. The primary source of uncertainty is identified as the material composition of the components, and research is suggested to address that uncertainty

  2. Comparison of tokamak burn cycle options

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1985-01-01

    Experimental confirmation of noninductive current drive has spawned a number of suggestions as to how this technique can be used to extend the fusion burn period and improve the reactor prospects of tokamaks. Several distinct burn cycles, which employ various combinations of Ohmic and noninductive current generation, are possible, and we will study their relative costs and benefits for both a commerical reactor as well as an INTOR-class device. We begin with a review of the burn cycle options

  3. Structural integrity and management of aging in internal components of BWR reactors

    International Nuclear Information System (INIS)

    Arganis J, C.R.

    2004-01-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  4. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

    International Nuclear Information System (INIS)

    2010-01-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

  5. Development and application of a welding procedure for remote repair of Magnox reactor internal components

    International Nuclear Information System (INIS)

    Morgan-Warren, E.J.

    1988-01-01

    This paper summarises the development and application of an all-welding repair method for reinforcing magnox reactor internal components. The development was dominated by the necessity for remote operation and the environmental constraints, in particular the oxide covering on the steel reactor structure. The choice of welding process is described, together with the development of the procedure for remote operation. The quality assurance procedure, including the verification of the technique and monitoring of the repair operation, is discussed. (author)

  6. IGORR-IV: Proceedings of the fourth meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    Rosenbalm, K.F.

    1995-01-01

    The fourth meeting of the International Group on Research Reactors (IGORR-IV) was attended by was good 55 registered participants from 28 organizations in 13 countries, which compares well with the previous meetings. Twenty-nine papers were presented in five sessions over the two-day meeting. Session subjects were: Operating Research Reactors and Facility Upgrades; Research Reactors in Desin and Construction; Research, Development, and Analysis Results of Thermal Hydraulic Calculations, U 3 Si 2 Fuel Performance and Faibrication; Structural Materials Performance; Neutronics; Severe Accident analysis. Written versions of the papers or hard copies of the viewgraphs used are published in these Proceedings

  7. Ukrainian Nuclear Society International Conference 'Modernization of the NPP with VVER reactor' (abstracts)

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1999-01-01

    Abstracts of the papers presented at International conference of the Ukrainian Nuclear Society 'Modernization of the NPP with VVER reactor'. The following problems are considered: improving the NPP's safety and reliability; reactor modernization, the lifetime prolongation; increasing of the reactor operating characteristics; methods of capacity factor increasing: refueling control, maintenance control; technical and economical aspects of NPP modernization; modernization of the automated control system of the fuel process at the NPP's; technical features and methods for the continued radiation and technology control at the NPP's; training, increasing the staff qualification and NPP modernization

  8. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-10-03

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

  9. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    International Nuclear Information System (INIS)

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-01-01

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies

  10. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  11. Status of tokamak research

    Energy Technology Data Exchange (ETDEWEB)

    Rawls, J.M. (ed.)

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design. (MOW)

  12. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

  13. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned.

  14. Technical Requirements For Reactors To Be Deployed Internationally For the Global Nuclear Energy Partnership

    International Nuclear Information System (INIS)

    Ingersoll, Daniel T.

    2007-01-01

    The Global Nuclear Energy Partnership (GNEP) seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy. Fully meeting the GNEP vision may require the deployment of thousands of reactors in scores of countries, many of which do not use nuclear energy currently. Some of these needs will be met by large-scale Generation III and III+ reactors (>1000 MWe) and Generation IV reactors when they are available. However, because many developing countries have small and immature electricity grids, the currently available Generation III(+) reactors may be unsuitable since they are too large, too expensive, and too complex. Therefore, GNEP envisions new types of reactors that must be developed for international deployment that are 'right sized' for the developing countries and that are based on technologies, designs, and policies focused on reducing proliferation risk. The first step in developing such systems is the generation of technical requirements that will ensure that the systems meet both the GNEP policy goals and the power needs of the recipient countries. Reactor systems deployed internationally within the GNEP context must meet a number of requirements similar to the safety, reliability, economics, and proliferation goals established for the DOE Generation IV program. Because of the emphasis on deployment to nonnuclear developing countries, the requirements will be weighted differently than with Generation IV, especially regarding safety and non-proliferation goals. Also, the reactors should be sized for market conditions in developing countries where energy demand per capita, institutional maturity and industrial infrastructure vary considerably, and must utilize fuel that is compatible with the fuel recycle technologies being developed by GNEP. Arrangements are already underway to establish Working Groups jointly with Japan and Russia to develop requirements for reactor systems. Additional bilateral and multilateral

  15. Gas reactor international coope--ative program. Interim report: assessment of gas-cooled reactor economics

    Energy Technology Data Exchange (ETDEWEB)

    1979-08-01

    A computer analysis of domestic economic incentive is presented. Included are the sample computer data set for ten combinations of reprocessing and reactor assumptions; basic data set and computer output; higher uranium availability computer output; 50 percent higher GCR fabrication cost computer output; 50 percent higher GCR reprocessing cost computer output; year 1990 and year 2000 GCR introduction scenario computer outputs; 75 percent perceived capacity factor for PBR computer output; and capital cost of GCRs 1.2 times that of LWRs.

  16. Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

    Science.gov (United States)

    Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

    2017-10-01

    The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

  17. Dual-cell reduction and group effect in an internal microelectrolysis reactor

    International Nuclear Information System (INIS)

    Ying, Diwen; Peng, Juan; Li, Kan; Wang, Yalin; Pan, Siwen; Jia, JinPing

    2013-01-01

    Highlights: ► 2,4-DCP was removed simultaneously in cathode and anode cells in IME reactor. ► Mechanism of dual-cell reduction gave an insight of cathode and anode cells. ► Significant V oc increase with one/two electrodes couples being installed in series. ► Group effect reveals possible high redox potential in IME reactor. -- Abstract: To address some of the fundamental questions regarding the functions of cathodes and anodes (e.g., iron and granular active carbon) and what happens when numerous particles (microscopic galvanic cells) are combined in the widely used and efficient wastewater treatment of an internal microelectrolysis (IME) reactor, we employed a specifically designed dual-cell IME reactor with single-couple electrodes and a narrow cell IME reactor with multi-couple electrodes. The simultaneous removal of 6.36 mg L −1 and 2.93 mg L −1 of 2,4-dichlorophenol within 30 min from the graphite cathode and the iron anode cells, respectively, was observed in the dual-cell reactor. The innovative concept behind this phenomenon is that the anode, which is generally believed to be oxidative, could probably be reductive in an IME reactor. Thus, it is important to understand the unique performance of IME reactors. The group effect, which provides a 45% increase of open-circuit potential with just two additional electrode couples in aqueous solution, was tested and verified in a multi-couple electrode reactor. It suggests a complex potential distribution in the IME reactor and that compounds even with high redox potential could possibly be reduced, which was generally believed to be difficult to accomplish

  18. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  19. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  20. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    International Nuclear Information System (INIS)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately