WorldWideScience

Sample records for fusion reactor concept

  1. SABR fusion-fission hybrid transmutation reactor design concept

    Science.gov (United States)

    Stacey, Weston

    2009-11-01

    A conceptual design has been developed for a sub-critical advanced burner reactor (SABR) consisting of i) a sodium cooled fast reactor fueled with the transuranics (TRU) from spent nuclear fuel, and ii) a D-T tokamak fusion neutron source based on ITER physics and technology. Subcritical operation enables more efficient transmutation fuel cycles in TRU fueled reactors (without compromising safety), which may be essential for significant reduction in high-level waste repository requirements. ITER will serve as the prototype for the fusion neutron source, which means SABRs could be implemented to help close the nuclear fuel cycle during the 2^nd quarter of the century.

  2. First wall and blanket concepts for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Biggio, M.; Cardella, A.; Daenner, W.; Farfaletti-Casali, F.; Ponti, C.; Rieger, M.; Vieider, G.

    1985-07-01

    The paper describes the progress of the studies on first wall and liquid breeder blankets for tritium production in the Next European Torus (NET). Two concepts of first wall/blanket segments are described, using 17Li83Pb as breeder and water as coolant. In both concepts the first wall is integrated in a steel box enveloping the breeder units which are cylindrical vessels with an inside heat transfer system. The thermomechanical and neutronics features of the two concepts are evaluated. Finally, the questions related to tritium permeation into coolant and tritium recovery from breeder are discussed on the basis of the analysis in progress in Europe.

  3. Compact fusion reactors

    CERN Document Server

    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.

  4. Hydrogen Production in Fusion Reactors

    OpenAIRE

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M; Uenosono, C.

    1993-01-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  5. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  6. Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.; Werner, R.W.; Carlson, G.A.; Cornish, D.N.

    1976-04-01

    Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements. Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated.

  7. A spherical torus nuclear fusion reactor space propulsion vehicle concept for fast interplanetary travel

    Science.gov (United States)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1999-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a>5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including payload, central truss, nuclear reactor (including diverter and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, and component design.

  8. A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

    Science.gov (United States)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1998-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

  9. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  10. Tritium management in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II.

  11. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  12. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  13. Adapting computational optimization concepts from aeronautics to nuclear fusion reactor design

    Directory of Open Access Journals (Sweden)

    Baelmans M.

    2012-10-01

    Full Text Available Even on the most powerful supercomputers available today, computational nuclear fusion reactor divertor design is extremely CPU demanding, not least due to the large number of design variables and the hybrid micro-macro character of the flows. Therefore, automated design methods based on optimization can greatly assist current reactor design studies. Over the past decades, “adjoint methods” for shape optimization have proven their virtue in the field of aerodynamics. Applications include drag reduction for wing and wing-body configurations. Here we demonstrate that also for divertor design, these optimization methods have a large potential. Specifically, we apply the continuous adjoint method to the optimization of the divertor geometry in a 2D poloidal cross section of an axisymmetric tokamak device (as, e.g., JET and ITER, using a simplified model for the plasma edge. The design objective is to spread the target material heat load as much as possible by controlling the shape of the divertor, while maintaining the full helium ash removal capabilities of the vacuum pumping system.

  14. Review of the safety concept for fusion reactor concepts and transferability of the nuclear fission regulation to potential fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Raeder, Juergen; Weller, Arthur; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik (IPP), Garching (Germany); Jin, Xue Zhou; Boccaccini, Lorenzo V.; Stieglitz, Robert; Carloni, Dario [Karlsruher Institute fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany); Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany); Herb, Joachim [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany)

    2016-01-15

    This paper summarizes the current state of the art in science and technology of the safety concept for future fusion power plants (FPPs) and examines the transferability of the current nuclear fission regulation to the concepts of future fusion power plants. At the moment there exist only conceptual designs of future fusion power plants. The most detailed concepts with regards to safety aspects were found in the European Power Plant Conceptual Study (PPCS). The plant concepts discussed in the PPCS are based on magnetic confinement of the plasma. The safety concept of fusion power plants, which has been developed during the last decades, is based on the safety concepts of installations with radioactive inventories, especially nuclear fission power plants. It applies the concept of defence in depth. However, there are specific differences between the implementations of the safety concepts due to the physical and technological characteristics of fusion and fission. It is analysed whether for fusion a safety concept is required comparable to the one of fission. For this the consequences of a purely hypothetical release of large amounts of the radioactive inventory of a fusion power plant and a fission power plant are compared. In such an event the evacuation criterion outside the plant is exceeded by several orders of magnitude for a fission power plant. For a fusion power plant the expected radiological consequences are of the order of the evacuation criterion. Therefore, a safety concept is also necessary for fusion to guarantee the confinement of the radioactive inventory. The comparison between the safety concepts for fusion and fission shows that the fundamental safety function ''confinement of the radioactive materials'' can be transferred directly in a methodical way. For a fusion power plant this fundamental safety function is based on both, physical barriers as well as on active retention functions. After the termination of the fusion

  15. Review of the safety concept for fusion reactor concepts and transferability of the nuclear fission regulation to potential fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Raeder, Juergen; Weller, Arthur; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik (IPP), Garching (Germany); Jin, Xue Zhou; Boccaccini, Lorenzo V.; Stieglitz, Robert; Carloni, Dario [Karlsruher Institute fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany); Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany); Herb, Joachim [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany)

    2016-01-15

    This paper summarizes the current state of the art in science and technology of the safety concept for future fusion power plants (FPPs) and examines the transferability of the current nuclear fission regulation to the concepts of future fusion power plants. At the moment there exist only conceptual designs of future fusion power plants. The most detailed concepts with regards to safety aspects were found in the European Power Plant Conceptual Study (PPCS). The plant concepts discussed in the PPCS are based on magnetic confinement of the plasma. The safety concept of fusion power plants, which has been developed during the last decades, is based on the safety concepts of installations with radioactive inventories, especially nuclear fission power plants. It applies the concept of defence in depth. However, there are specific differences between the implementations of the safety concepts due to the physical and technological characteristics of fusion and fission. It is analysed whether for fusion a safety concept is required comparable to the one of fission. For this the consequences of a purely hypothetical release of large amounts of the radioactive inventory of a fusion power plant and a fission power plant are compared. In such an event the evacuation criterion outside the plant is exceeded by several orders of magnitude for a fission power plant. For a fusion power plant the expected radiological consequences are of the order of the evacuation criterion. Therefore, a safety concept is also necessary for fusion to guarantee the confinement of the radioactive inventory. The comparison between the safety concepts for fusion and fission shows that the fundamental safety function ''confinement of the radioactive materials'' can be transferred directly in a methodical way. For a fusion power plant this fundamental safety function is based on both, physical barriers as well as on active retention functions. After the termination of the fusion

  16. Research on the HYLIFE liquid-first-wall concept for future laser-fusion reactors. Final report No. 5

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.

    1980-09-01

    It has been proposed to protect the structural walls of a future laser fusion reactor with a curtain or fluid-wall of liquid lithium jets. As part of the investigation of this concept, experiments have been performed on planar sheet water jets issuing vertically downward from slit nozzles. The nozzles were subjected to transverse forced harmonic excitation to simulate the vibrational environment of the laser fusion reactor, and experiments were run at both 1 atm and at lower ambient pressures. Linear temporal stability theory is shown to predict the onset of the unstable regime and the initial spatial growth rates quite well for the cases where the amplitudes of the nozzle vibration are not too large and the waveform is nearly sinusoidal. In addition, both the linear theory and a simplified trajectory theory are shown to predict the initial wave envelope amplitudes very well. For larger amplitude nozzle excitation, the waveform becomes highly nonlinear and non-sinusoidal and can resemble a sawtooth waveform in some cases; these latter experimental results can only be partially explained by existing theories at the present time.

  17. Accelerator based fusion reactor

    Science.gov (United States)

    Liu, Keh-Fei; Chao, Alexander Wu

    2017-08-01

    A feasibility study of fusion reactors based on accelerators is carried out. We consider a novel scheme where a beam from the accelerator hits the target plasma on the resonance of the fusion reaction and establish characteristic criteria for a workable reactor. We consider the reactions d+t\\to n+α,d+{{}3}{{H}\\text{e}}\\to p+α , and p+{{}11}B\\to 3α in this study. The critical temperature of the plasma is determined from overcoming the stopping power of the beam with the fusion energy gain. The needed plasma lifetime is determined from the width of the resonance, the beam velocity and the plasma density. We estimate the critical beam flux by balancing the energy of fusion production against the plasma thermo-energy and the loss due to stopping power for the case of an inert plasma. The product of critical flux and plasma lifetime is independent of plasma density and has a weak dependence on temperature. Even though the critical temperatures for these reactions are lower than those for the thermonuclear reactors, the critical flux is in the range of {{10}22}-{{10}24}~\\text{c}{{\\text{m}}-2}~{{\\text{s}}-1} for the plasma density {ρt}={{10}15}~\\text{c}{{\\text{m}}-3} in the case of an inert plasma. Several approaches to control the growth of the two-stream instability are discussed. We have also considered several scenarios for practical implementation which will require further studies. Finally, we consider the case where the injected beam at the resonance energy maintains the plasma temperature and prolongs its lifetime to reach a steady state. The equations for power balance and particle number conservation are given for this case.

  18. Status and problems of fusion reactor development.

    Science.gov (United States)

    Schumacher, U

    2001-03-01

    Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

  19. Magnetic fusion reactor economics

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A.

    1995-12-01

    An almost primordial trend in the conversion and use of energy is an increased complexity and cost of conversion systems designed to utilize cheaper and more-abundant fuels; this trend is exemplified by the progression fossil fission {yields} fusion. The present projections of the latter indicate that capital costs of the fusion ``burner`` far exceed any commensurate savings associated with the cheapest and most-abundant of fuels. These projections suggest competitive fusion power only if internal costs associate with the use of fossil or fission fuels emerge to make them either uneconomic, unacceptable, or both with respect to expensive fusion systems. This ``implementation-by-default`` plan for fusion is re-examined by identifying in general terms fusion power-plant embodiments that might compete favorably under conditions where internal costs (both economic and environmental) of fossil and/or fission are not as great as is needed to justify the contemporary vision for fusion power. Competitive fusion power in this context will require a significant broadening of an overly focused program to explore the physics and simbiotic technologies leading to more compact, simplified, and efficient plasma-confinement configurations that reside at the heart of an attractive fusion power plant.

  20. Materials issues in fusion reactors

    Science.gov (United States)

    Suri, A. K.; Krishnamurthy, N.; Batra, I. S.

    2010-02-01

    The world scientific community is presently engaged in one of the toughest technological tasks of the current century, namely, exploitation of nuclear fusion in a controlled manner for the benefit of mankind. Scientific feasibility of controlled fusion of the light elements in plasma under magnetic confinement has already been proven. International efforts in a coordinated and co-operative manner are presently being made to build ITER - the International Thermonuclear Experimental Reactor - to test, in this first step, the concept of 'Tokamak' for net fusion energy production. To exploit this new developing option of making energy available through the route of fusion, India too embarked on a robust fusion programme under which we now have a working tokamak - the Aditya and a steady state tokamak (SST-1), which is on the verge of functioning. The programme envisages further development in terms of making SST-2 followed by a DEMO and finally the fusion power reactor. Further, with the participation of India in the ITER program in 2005, and recent allocation of half - a - port in ITER for placing our Lead - Lithium Ceramic Breeder (LLCB) based Test Blanket Module (TBM), meant basically for breeding tritium and extracting high grade heat, the need to understand and address issues related to materials for these complex systems has become all the more necessary. Also, it is obvious that with increasing power from the SST stages to DEMO and further to PROTOTYPE, the increasing demands on performance of materials would necessitate discovery and development of new materials. Because of the 14.1 MeV neutrons that are generated in the D+T reaction exploited in a tokamak, the materials, especially those employed for the construction of the first wall, the diverter and the blanket segments, suffer crippling damage due to the high He/dpa ratios that result due to the high energy of the neutrons. To meet this challenge, the materials that need to be developed for the tokamaks

  1. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  2. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  3. Research on the wetted first wall concept for future laser fusion reactors. Final report No. 1, October 1, 1974--January 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.; Munir, Z.A.

    1976-01-01

    Research is in progress to determine the feasibility of the wetted first wall concept for a future laser fusion reactor. The basic idea involves the use of a thin coating of lithium on the inner wall of the laser fusion containment vessel to protect it from the micro-explosion blast debris. This report contains a review of the available information on contact angles and wettability of alkali metals on various metal substrates as well as a review of literature on thin falling liquid films. A proposed experiment to measure the contact angles of lithium on stainless steel and niobium is described. The requirements for a second experiment to measure certain key characteristics of thin falling films are also included.

  4. (Meeting on fusion reactor materials)

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.H. (Pacific Northwest Lab., Richland, WA (USA)); Klueh, R.L.; Rowcliffe, A.F.; Wiffen, F.W. (Oak Ridge National Lab., TN (USA)); Loomis, B.A. (Argonne National Lab., IL (USA))

    1990-11-01

    During his visit to the KfK, Karlsruhe, F. W. Wiffen attended the IEA 12th Working Group Meeting on Fusion Reactor Materials. Plans were made for a low-activation materials workshop at Culham, UK, for April 1991, a data base workshop in Europe for June 1991, and a molecular dynamics workshop in the United States in 1991. At the 11th IEA Executive Committee on Fusion Materials, discussions centered on the recent FPAC and Colombo panel review in the United States and EC, respectively. The Committee also reviewed recent progress toward a neutron source in the United States (CWDD) and in Japan (ESNIT). A meeting with D. R. Harries (consultant to J. Darvas) yielded a useful overview of the EC technology program for fusion. Of particular interest to the US program is a strong effort on a conventional ferritic/martensitic steel for fist wall/blanket operation beyond NET/ITER.

  5. Nuclear data requirements for fusion reactor nucleonics

    Energy Technology Data Exchange (ETDEWEB)

    Bhat, M.R.; Abdou, M.A.

    1980-01-01

    Nuclear data requirements for fusion reactor nucleonics are reviewed and the present status of data are assessed. The discussion is divided into broad categories dealing with data for Fusion Materials Irradiation Test Facility (FMIT), D-T Fusion Reactors, Alternate Fuel Cycles and the Evaluated Data Files that are available or would be available in the near future.

  6. Reactor potential for magnetized target fusion

    Energy Technology Data Exchange (ETDEWEB)

    Dahlin, J.E

    2001-06-01

    Magnetized Target Fusion (MTF) is a possible pathway to thermonuclear fusion different from both magnetic fusion and inertial confinement fusion. An imploding cylindrical metal liner compresses a preheated and magnetized plasma configuration until thermonuclear conditions are achieved. In this report the Magnetized Target Fusion concept is evaluated and a zero-dimensional computer model of the plasma, liner and circuit as a connected system is designed. The results of running this code are that thermonuclear conditions are achieved indeed, but only during a very short time. At peak compression the pressure from the compressed plasma and magnetic field is so large reversing the liner implosion into an explosion. The time period of liner motion reversal is termed the dwell time and is crucial to the performance of the fusion system. Parameters as liner thickness and plasma density are certainly of significant importance to the dwell time, but it seems like a reactor based on the MTF principle hardly can become economic if not innovative solutions are introduced. In the report two such solutions are presented as well.

  7. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2004-06-14

    A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.

  8. Compound cryopump for fusion reactors

    CERN Document Server

    Kovari, M; Shephard, T

    2013-01-01

    We reconsider an old idea: a three-stage compound cryopump for use in fusion reactors such as DEMO. The helium "ash" is adsorbed on a 4.5 K charcoal-coated surface, while deuterium and tritium are adsorbed at 15-22 K on a second charcoal-coated surface. The helium is released by raising the first surface to ~30 K. In a separate regeneration step, deuterium and tritium are released at ~110 K. In this way, the helium can be pre-separated from other species. In the simplest design, all three stages are in the same vessel, with a single valve to close the pump off from the tokamak during regeneration. In an alternative design, the three stages are in separate vessels, connected by valves, allowing the stages to regenerate without interfering with each other. The inclusion of the intermediate stage would not affect the overall pumping speed significantly. The downstream exhaust processing system could be scaled down, as much of the deuterium and tritium could be returned directly to the reactor. This could reduce ...

  9. Generic Magnetic Fusion Reactor Revisited

    Science.gov (United States)

    Sheffield, John; Milora, Stanley

    2015-11-01

    The original Generic Magnetic Fusion Reactor paper was published in 1986. This update describes what has changed in 30 years. Notably, the construction of ITER is providing important benchmark numbers for technologies and costs. In addition, we use a more conservative neutron wall flux and fluence. But these cost-increasing factors are offset by greater optimism on the thermal-electric conversion efficiency and potential availability. The main examples show the cost of electricity (COE) as a function of aspect ratio and neutron flux to the first wall. The dependence of the COE on availability, thermo-electric efficiency, electrical power output, and the present day's low interest rates is also discussed. Interestingly, at fixed aspect ratio there is a shallow minimum in the COE at neutron flux around 2.5 MW/m2. The possibility of operating with only a small COE penalty at even lower wall loadings (to 1.0 MW/m2 at larger plant size) and the use of niobium-titanium coils are also investigated. J. Sheffield was supported by ORNL subcontract 4000088999 with the University of Tennessee.

  10. Data Fusion Concepts and Ideas

    CERN Document Server

    Mitchell, H B

    2012-01-01

    “Data Fusion: Concepts and Ideas” provides a comprehensive introduction to the concepts and idea of multisensor data fusion. This textbook is an extensively revised second edition of the author's successful book: "Multi-Sensor Data Fusion: An Introduction". The book is self-contained and no previous knowledge of multi-sensor data fusion is assumed. The reader is made familiar with tools taken from a wide range of diverse subjects including: neural networks, signal processing, statistical estimation, tracking algorithms, computer vision and control theory which are combined by using a common statistical framework. As a consequence, the underlying pattern of relationships that exists between the different methodologies is made evident. The book is illustrated with many real-life examples taken from a diverse range of applications and contains an extensive list of modern references. The new completely revised and updated edition includes nearly 70 pages of new material including a full new chapter as well as...

  11. Radiochemical problems of fusion reactors. 1. Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Crespi, M.B.A.

    1984-02-01

    A list of fusion reactor candidate materials is given, for use in connection with blanket structure, breeding, moderation, neutron multiplication, cooling, magnetic field generation, electrical insulation and radiation shielding. The phenomena being studied for each group of materials are indicated. Suitable irradiation test facilities are discussed under the headings (1) accelerator-based neutron sources, (2) fission reactors, and (3) ion accelerators.

  12. Advanced fusion concepts: project summaries

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-12-01

    This report contains descriptions of the activities of all the projects supported by the Advanced Fusion Concepts Branch of the Office of Fusion Energy, US Department of Energy. These descriptions are project summaries of each of the individual projects, and contain the following: title, principle investigators, funding levels, purpose, approach, progress, plans, milestones, graduate students, graduates, other professional staff, and recent publications. Information is given for each of the following programs: (1) reverse-field pinch, (2) compact toroid, (3) alternate fuel/multipoles, (4) stellarator/torsatron, (5) linear magnetic fusion, (6) liners, and (7) Tormac. (MOW)

  13. Overview of the Lockheed Martin Compact Fusion Reactor (CFR) Program

    Science.gov (United States)

    McGuire, Thomas

    2015-11-01

    The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. An overview of the concept and its diamagnetic, high beta magnetically encapsulated linear ring cusp confinement scheme will be given. The analytical model of the major loss mechanisms and predicted performance will be discussed, along with the major physics challenges. Key features of an operational CFR reactor will be highlighted. The proposed developmental path following the current experimental efforts will be presented. ©2015 Lockheed Martin Corporation. All Rights Reserved.

  14. Cold nuclear fusion reactor and nuclear fusion rocket

    Directory of Open Access Journals (Sweden)

    Huang Zhenqiang

    2013-10-01

    Full Text Available "Nuclear restraint inertial guidance directly hit the cold nuclear fusion reactor and ion speed dc transformer" [1], referred to as "cold fusion reactor" invention patents, Chinese Patent Application No. CN: 200910129632.7 [2]. The invention is characterized in that: at room temperature under vacuum conditions, specific combinations of the installation space of the electromagnetic field, based on light nuclei intrinsic magnetic moment and the electric field, the first two strings of the nuclei to be bound fusion on the same line (track of. Re-use nuclear spin angular momentum vector inherent nearly the speed of light to form a super strong spin rotation gyro inertial guidance features, to overcome the Coulomb repulsion strong bias barrier to achieve fusion directly hit. Similar constraints apply nuclear inertial guidance mode for different speeds and energy ion beam mixing speed, the design of ion speed dc transformer is cold fusion reactors, nuclear fusion engines and such nuclear power plants and power delivery systems start important supporting equipment, so apply for a patent merger

  15. Packed fluidized bed blanket for fusion reactor

    Science.gov (United States)

    Chi, John W. H.

    1984-01-01

    A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

  16. The first fusion reactor: ITER

    Science.gov (United States)

    Campbell, D. J.

    2016-11-01

    Established by the signature of the ITER Agreement in November 2006 and currently under construction at St Paul-lez-Durance in southern France, the ITER project [1,2] involves the European Union (including Switzerland), China, India, Japan, the Russian Federation, South Korea and the United States. ITER (`the way' in Latin) is a critical step in the development of fusion energy. Its role is to provide an integrated demonstration of the physics and technology required for a fusion power plant based on magnetic confinement.

  17. Optimization of the fission--fusion hybrid concept

    Energy Technology Data Exchange (ETDEWEB)

    Saltmarsh, M.J.; Grimes, W.R.; Santoro, R.T.

    1979-04-01

    One of the potentially attractive applications of controlled thermonuclear fusion is the fission--fusion hybrid concept. In this report we examine the possible role of the hybrid as a fissile fuel producer. We parameterize the advantages of the concept in terms of the performance of the fusion device and the breeding blanket and discuss some of the more troublesome features of existing design studies. The analysis suggests that hybrids based on deuterium--tritium (D--T) fusion devices are unlikely to be economically attractive and that they present formidable blanket technology problems. We suggest an alternative approach based on a semicatalyzed deuterium--deuterium (D--D) fusion reactor and a molten salt blanket. This concept is shown to emphasize the desirable features of the hybrid, to have considerably greater economic potential, and to mitigate many of the disadvantages of D--T-based systems.

  18. Evolution of the tandem mirror reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, G.A.; Logan, B.G.

    1982-03-09

    We discuss the evolution of the tandem mirror reactor concept from the original conceptual reactor design (1977) through the first application of the thermal barrier concept to a reactor design (1979) to the beginning of the Mirror Advanced Reactor Study (1982).

  19. Fusion reactors for hydrogen production via electrolysis

    Science.gov (United States)

    Fillo, J. A.; Powell, J. R.; Steinberg, M.

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and supplement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of 40 to 60% and hydrogen production efficiencies by high temperature electrolysis of 50 to 70% are projected for fusion reactors using high temperature blankets.

  20. Burning high-level TRU waste in fusion fission reactors

    Science.gov (United States)

    Shen, Yaosong

    2016-09-01

    Recently, the concept of actinide burning instead of a once-through fuel cycle for disposing spent nuclear fuel seems to get much more attention. A new method of burning high-level transuranic (TRU) waste combined with Thorium-Uranium (Th-U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper. The thorium-based TRU fuel burns all of the long-lived actinides via a hard neutron spectrum while outputting power. A one-dimensional model of the reactor concept was built by means of the ONESN_BURN code with new data libraries. The numerical results included actinide radioactivity, biological hazard potential, and much higher burnup rate of high-level transuranic waste. The comparison of the fusion-fission reactor with the thermal reactor shows that the harder neutron spectrum is more efficient than the soft. The Th-U cycle produces less TRU, less radiotoxicity and fewer long-lived actinides. The Th-U cycle provides breeding of 233U with a long operation time (>20 years), hence significantly reducing the reactivity swing while improving safety and burnup.

  1. Materials needs for compact fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A.

    1983-01-01

    The economic prospects for magnetic fusion energy can be dramatically improved if for the same total power output the fusion neutron first-wall (FW) loading and the system power density can be increased by factors of 3 to 5 and 10 to 30, respectively. A number of compact fusion reactor embodiments have been proposed, all of which would operate with increased FW loadings, would use thin (0.5 to 0.6 m) blankets, and would confine quasi-steady-state plasma with resistive, water-cooled copper or aluminum coils. Increased system power density (5 to 15 MWt/m/sup 3/ versus 0.3 to 0.5 MW/m/sup 3/), considerably reduced physical size of the fusion power core (FPC), and appreciably reduced economic leverage exerted by the FPC and associated physics result. The unique materials requirements anticipated for these compact reactors are outlined against the well documented backdrop provided by similar needs for the mainline approaches. Surprisingly, no single materials need that is unique to the compact systems is identified; crucial uncertainties for the compact approaches must also be addressed by the mainline approaches, particularly for in-vacuum components (FWs, limiters, divertors, etc.).

  2. Carbon-14 production in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, R.D.; Burger, L.L.

    1976-09-01

    Calculations based on existing composition data were performed to estimate the order of magnitude and the final location of /sup 14/C in fusion reactors. These calculations indicate that approximately 8 Ci/day, formed principally by /sup 14/N activation, will be produced in the UWMAK-II reference reactor (5,000 MWth). If Nb-1 percent Zr is used as the structural material instead of stainless steel 316 this quantity will be more than doubled. No information is available on the form of the /sup 14/C produced, but reduced forms such as carbides, hydrocarbons and perhaps CO may be produced. Most of the /sup 14/C may remain fixed in structural and other reactor materials until the material is reclaimed. Activation of air in the plasma chamber would be an immediate concern.

  3. Fusion reactors-high temperature electrolysis (HTE)

    Energy Technology Data Exchange (ETDEWEB)

    Fillo, J.A. (ed.)

    1978-01-01

    Results of a study to identify and develop a reference design for synfuel production based on fusion reactors are given. The most promising option for hydrogen production was high-temperature electrolysis (HTE). The main findings of this study are: 1. HTE has the highest potential efficiency for production of synfuels from fusion; a fusion to hydrogen energy efficiency of about 70% appears possible with 1800/sup 0/C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 1400/sup 0/C HTE units and 40% power cycle efficiency. 2. Relative to thermochemical or direct decomposition methods HTE technology is in a more advanced state of development, 3. Thermochemical or direct decomposition methods must have lower unit process or capital costs if they are to be more attractive than HTE. 4. While design efforts are required, HTE units offer the potential to be quickly run in reverse as fuel cells to produce electricity for restart of Tokamaks and/or provide spinning reserve for a grid system. 5. Because of the short timescale of the study, no detailed economic evaluation could be carried out.A comparison of costs could be made by employing certain assumptions. For example, if the fusion reactor-electrolyzer capital installation is $400/(KW(T) ($1000/KW(E) equivalent), the H/sub 2/ energy production cost for a high efficiency (about 70 %) fusion-HTE system is on the same order of magnitude as a coal based SNG plant based on 1976 dollars. 6. The present reference design indicates that a 2000 MW(th) fusion reactor could produce as much at 364 x 10/sup 6/ scf/day of hydrogen which is equivalent in heating value to 20,000 barrels/day of gasoline. This would fuel about 500,000 autos based on average driving patterns. 7. A factor of three reduction in coal feed (tons/day) could be achieved for syngas production if hydrogen from a fusion-HTE system were used to gasify coal, as compared to a conventional syngas plant using coal-derived hydrogen.

  4. Investigation of materials for fusion power reactors

    Science.gov (United States)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  5. Laser fusion power reactor system (LFPRS)

    Energy Technology Data Exchange (ETDEWEB)

    Kovacik, W. P.

    1977-12-19

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements. (MOW)

  6. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    Science.gov (United States)

    1985-02-01

    Public Affairs Office and is releasaole to the National Technical Information Services (NTIS). At NTIS, it will be available to the general public...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept...reactor core at the top and discharged at the Dotton while the reactor is in operation. The discharged fuel can then b inspected to see if it can De used

  7. The TITAN reversed-field-pinch fusion reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This paper on titan plasma engineering contains papers on the following topics: reversed-field pinch as a fusion reactor; parametric systems studies; magnetics; burning-plasma simulations; plasma transient operations; current drive; and physics issues for compact RFP reactors.

  8. Modular stellarator reactor: a fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.

    1983-07-01

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  9. 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.

  10. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  11. Fusion reactor blanket/shield design study

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Clemmer, R.G.; Harkness, S.D.

    1979-07-01

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented.

  12. A fusion reactor for space applications

    Energy Technology Data Exchange (ETDEWEB)

    Kammash, T.; Galbraith, D.L.

    1987-07-01

    A novel approach to fusion power that combines the favorable aspects of magnetic and inertial confinements has recently been proposed in the ''magnetically insulated inertial confinement fusion'' (MICF) reactor. In contrast to conventional inertial confinement schemes, this approach relies on generating the needed plasma inside of a spherical shell by zapping the inside surface of a hollow pellet with an intense laser beam. Physical confinement is provided by the metallic shell that surrounds the deuterium-tritium fuel-coated inner surface, while very strong, plasma-generated magnetic fields provide the desired thermal insulation of the plasma from the surrounding surface. Because of these unique properties, the inertial confinement time can be increased by about two orders of magnitude relative to that of conventional inertial confinement schemes, with the result that truly impressive energy multiplication factors can result. Carbon dioxide lasers of hundreds of kilojoules may be readily employed for such reactors, and, since they are relatively efficient and can be chemically driven, these systems lend themselves nicely to such space applications as space-based power sources or rocket propulsion.

  13. Neutronic predesign tool for fusion power reactors system assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, J.-C., E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Li Puma, A. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martínez Arroyo, J. [ETSEIB, Internship in CEA (Spain)

    2013-10-15

    SYCOMORE, a fusion reactor system code based on a modular approach, is under development at CEA. In this framework, this paper describes a methodology developed to build the neutronic module of SYCOMORE. This neutronic module aims to evaluate main neutronic parameters characterising a fusion reactor (tokamak): tritium breeding ratio, multiplication factor, nuclear heating as a function of the reactor main geometrical parameters (major radius, elongation, etc.), of the radial build, Li enrichment, blanket and shield thickness, etc. It is based on calculations carried out with APOLLO2 and TRIPOLI-4 CEA transport code on simplified 1D and 2D neutronic models. These models are validated versus a more detailed 3D Monte-Carlo model (using TRIPOLI-4). To ease the integration of this neutronic module in SYCOMORE and provide results instantly, a surrogate model that replicates the 1D and 2D neutronic model results was used. Among the different surrogate models types (polynomial interpolation, responses functions, interpolating by Kriging, artificial neural network, etc.) the neural networks were selected for their efficiency and flexibility. The methodology described in this paper to build SYCOMORE neutronic module is devoted to HCLL blanket, but it could be applied to any breeder blanket concept provided that appropriate validation could be carried out.

  14. The need and prospects for improved fusion reactors

    Science.gov (United States)

    Krakowski, R. A.; Miller, R. L.; Hagenson, R. L.

    1986-09-01

    Conceptual fusion reactor studies over the past 10-15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100-200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed.

  15. Fusion reactor handling operations with cable-driven parallel robots

    Energy Technology Data Exchange (ETDEWEB)

    Izard, Jean-Baptiste, E-mail: jeanbaptiste.izard@tecnalia.com; Michelin, Micael; Baradat, Cédric

    2015-10-15

    Highlights: • CDPR allow 6DOF positioning of loads using cable as links without payload swag. • Conceptual design of a CDPR for carrying and positioning tokamak sectors is given. • A CDPR for threading stellarator coils (6D trajectory following) is provided. • Both designs are capable of fullfilling the required precision without tooling. - Abstract: Cable-driven parallel robots (CDPR) are in their concept cranes with inclined cables which allow control of all the degrees of freedom of its payload, and therefore stability of all the degrees of freedom, including rotations. The workspace of a CDPR is only limited by the length of the cables, and the payload capacity related to the mass of the whole robot is very important. Besides, the control being based on kinematic models, the behavior of a CDPR is really that of a robot capable of automated trajectories or remote handling. The present paper gives a presentation of two use case studies based on some of the assembly phases and remote handling actions as designed for the recent fusion machines. Based on the use cases already in place in fusion reactor baselines, the opportunity of using CDPR for assembly of structural elements and coils is discussed. Finally, prospects for remote handling equipment from the reactor in hot cells are envisioned based on current CDPR research.

  16. Concept for LEU Burst Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kimpland, Robert Herbert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-07

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  17. Assessment of tritium breeding requirements for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, J.

    1983-12-01

    This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios.

  18. Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

    Science.gov (United States)

    Fillo, J. A.

    This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

  19. Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

    Science.gov (United States)

    Fillo, J. A.

    This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

  20. Burning high-level TRU waste in fusion fission reactors

    National Research Council Canada - National Science Library

    Shen, Yaosong

    2016-01-01

    .... A new method of burning high-level transuranic (TRU) waste combined with Thorium–Uranium (Th–U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper...

  1. Project Icarus: Nuclear Fusion Propulsion Concept Comparison

    Science.gov (United States)

    Stanic, M.

    Project Icarus will use nuclear fusion as the primary propulsion, since achieving breakeven is imminent within the next decade. Therefore, fusion technology provides confidence in further development and fairly high technological maturity by the time the Icarus mission would be plausible. Currently there are numerous (over 2 dozen) different fusion approaches that are simultaneously being developed around the World and it is difficult to predict which of the concepts is going to be the most successful one. This study tried to estimate current technological maturity and possible technological extrapolation of fusion approaches for which appropriate data could be found. Figures of merit that were assessed include: current technological state, mass and volume estimates, possible gain values, main advantages and disadvantages of the concept and an attempt to extrapolate current technological state for the next decade or two. Analysis suggests that Magnetic Confinement Fusion (MCF) concepts are not likely to deliver sufficient performance due to size, mass, gain and large technological barriers of the concept. However, ICF and PJMIF did show potential for delivering necessary performance, assuming appropriate techno- logical advances. This paper is a submission of the Project Icarus Study Group.

  2. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  3. Advanced fusion MHD power conversion using the CFAR (compact fusion advanced Rankine) cycle concept

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-10-01

    The CFAR (compact fusion advanced Rankine) cycle concept for a tokamak reactor involves the use of a high-temperature Rankine cycle in combination with microwave superheaters and nonequilibrium MHD disk generators to obtain a compact, low-capital-cost power conversion system which fits almost entirely within the reactor vault. The significant savings in the balance-of-plant costs are expected to result in much lower costs of electricity than previous concepts. This paper describes the unique features of the CFAR cycle and a high- temperature blanket designed to take advantage of it as well as the predicted performance of the MHD disk generators using mercury seeded with cesium. 40 refs., 8 figs., 3 tabs.

  4. Experimental devices in the osiris reactor to study effects of radiations on fusion reactor materials

    Science.gov (United States)

    Lefevre, F.; Thevenot, G.

    1986-11-01

    Within the framework of the Technology Research Program on controlled fusion initiated by the European Communities, the Services des Piles de Saclay (SPS) of Commissariat à l'Energie Atomique (CEA) have been requested to perform some necessary experiments to study the irradiation behaviour of materials which are possible candidates for controlled fusion reactors. This paper describes the devices, generally adapted from a standard model "The COLIBRI", which allow one to carry out, in the OSIRIS reactor, irradiations on the three great families of fusion reactor materials: - lithium containing materials of breeding blanket for in-situ tritium production, - protection materials, and - structural materials.

  5. Experimental devices in the OSIRIS reactor to study effects of radiations on fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Lefevre, F.; Thevenot, G.

    Within the framework of the Technology Research Program on controlled fusion initiated by the European Communities, the Services des Piles de Saclay (SPS) of Commissariat a l'Energie Atomique (CEA) have been requested to perform some necessary experiments to study the irradiation behaviour of materials which are possible candidates for controlled fusion reactors. This paper describes the devices, generally adapted from a standard model The COLIBRI, which allow one to carry out, in the OSIRIS reactor, irradiations on the three great families of fusion reactor materials: Lithium containing materials of breeding blanket for in-situ tritium production, protection materials, and structural materials.

  6. Concept of DT fuel cycle for a fusion neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Anan' ev, S.; Spitsyn, A.V.; Kuteev, B.V.; Cherkez, D.I. [NRC Kurchatov Institute, Moscow (Russian Federation); Shirnin, P.N.; Kazakovsky, N.T. [FSUE RFNC - VNIIEF, Sarov (Russian Federation)

    2015-03-15

    A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 10{sup 18} neutrons per second is under design in Russia. Such a FNS is of interest for many applications: 1) basic and applied research (neutron scattering, etc); 2) testing the structural materials for fusion reactors; 3) control of sub-critical nuclear systems and 4) nuclear waste processing (including transmutation of minor actinides). This paper describes the fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ∼0.5 and ∼0.3 m, magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1-2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems. (authors)

  7. Concept design on RH maintenance of CFETR Tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Wu, Songtao; Wan, Yuanxi; Li, Jiangang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Ye, Minyou [University of Science and Technology of China, Hefei (China); Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2014-10-15

    Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.

  8. 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...

  9. Cost assessment of a generic magnetic fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.; Dory, R.A.; Cohn, S.M.; Delene, J.G.; Parsly, L.F.; Ashby, D.E.T.F.; Reiersen, W.T.

    1986-03-01

    A generic reactor model is used to examine the economic viability of generating electricity by magnetic fusion. The simple model uses components that are representative of those used in previous reactor studies of deuterium-tritium-burning tokamaks, stellarators, bumpy tori, reversed-field pinches (RFPs), and tandem mirrors. Conservative costing assumptions are made. The generic reactor is not a tokamak; rather, it is intended to emphasize what is common to all magnetic fusion rectors. The reactor uses a superconducting toroidal coil set to produce the dominant magnetic field. To this extent, it is not as good an approximation to systems such as the RFP in which the main field is produced by a plasma current. The main output of the study is the cost of electricity as a function of the weight and size of the fusion core - blanket, shield, structure, and coils. The model shows that a 1200-MW(e) power plant with a fusion core weight of about 10,000 tonnes should be competitive in the future with fission and fossil plants. Studies of the sensitivity of the model to variations in the assumptions show that this result is not sensitively dependent on any given assumption. Of particular importance is the result that a fusion reactor of this scale may be realized with only moderate advances in physics and technology capabilities.

  10. Radioactive waste management and disposal scenario for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tabara, Takashi; Yamano, Naoki [Sumitomo Atomic Energy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao

    1997-10-01

    The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a light water reactor (LWR) have been evaluated and compared. At first, the amount and the radioactive level of the radwaste generated in five fusion reactors ware evaluated by an activation calculation code. Next, a possible radwaste disposal scenario applicable to fusion radwaste in Japan is considered and the disposal cost evaluated under certain assumptions. The exposure doses are evaluated for the skyshine of gamma-rays during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical LWR was estimated based on a literature survey and the disposal cost was evaluated using the same assumptions as for the fusion reactors. It is found that the relative cost of disposal is strongly dependent on the cost for interim storage of medium level waste of fusion reactors and the cost of high level waste for the LWR. (author)

  11. Current fusion power plant design concepts

    Energy Technology Data Exchange (ETDEWEB)

    Gore, B.F.; Murphy, E.S.

    1976-09-01

    Nine current U.S. designs for fusion power plants are described in this document. Summary tabulations include a tenth concept, for which the design document was unavailable during preparation of the descriptions. The information contained in the descriptions was used to define an envelope of fusion power plant characteristics which formed the basis for definition of reference first commercial fusion power plant design. A brief prose summary of primary plant features introduces each of the descriptions contained in the body of this document. In addition, summary tables are presented. These tables summarize in side-by-side fashion, plant parameters, processes, combinations of materials used, requirements for construction materials, requirements for replacement materials during operation, and production of wastes.

  12. Proposal for a novel type of small scale aneutronic fusion reactor

    Science.gov (United States)

    Gruenwald, J.

    2017-02-01

    The aim of this work is to propose a novel scheme for a small scale aneutronic fusion reactor. This new reactor type makes use of the advantages of combining laser driven plasma acceleration and electrostatic confinement fusion. An intense laser beam is used to create a lithium-proton plasma with high density, which is then collimated and focused into the centre of the fusion reaction chamber. The basic concept presented here is based on the 7Li-proton fusion reaction. However, the physical and technological fundamentals may generally as well be applied to 11B-proton fusion. The former fusion reaction path offers higher energy yields while the latter has larger fusion cross sections. Within this paper a technological realisation of such a fusion device, which allows a steady state operation with highly energetic, well collimated ion beam, is presented. It will be demonstrated that the energetic break even can be reached with this device by using a combination of already existing technologies.

  13. Vanadium-base alloys for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Loomis, B.A.; Diercks, D.R.

    1984-10-01

    Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined.

  14. Cold fusion reactors and new modern physics

    Directory of Open Access Journals (Sweden)

    Huang Zhenqiang Huang Yuxiang

    2013-10-01

    Full Text Available The author of the "modern physics classical particle quantization orbital motion model general solution", referred to as the “new modern physics” a book. “The nuclear force constraint inertial guidance cold nuclear fusion collides” patent of invention referred to as the “cold nuclear fusion reactor” detailed technical data. Now provide to you, hope you help spread and the mainstream of modern physics of academic and fusion engineering academic communication. We work together to promote the cause of science and technology progress of mankind to contribute

  15. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  16. Developments and Tendencies in Fission Reactor Concepts

    Science.gov (United States)

    Adamov, E. O.; Fuji-Ie, Y.

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

  17. Tokamak fusion reactors with less than full tritium breeding

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. An evaluation of fusion gain in the compact helical fusion reactor FFHR-c1

    Science.gov (United States)

    Miyazawa, J.; Goto, T.; Sakamoto, R.; Sagara, A.; the FFHR Design Group

    2014-01-01

    A new procedure to predict achievable fusion gain in a sub-ignition fusion reactor is proposed. This procedure uses the direct profile extrapolation (DPE) method based on the gyro-Bohm model. The DPE method has been developed to predict the radial profiles in a fusion reactor sustained without auxiliary heating (i.e., in the self-ignition state) from the experimental data. To evaluate the fusion gain in a fusion reactor sustained with auxiliary heating (i.e., in the sub-ignition state), the DPE method is modified to include the influence of the auxiliary heating. The beta scale factor from experiment to reactor is assumed to be 1. Under this assumption, it becomes reasonable to apply the magnetohydrodynamic (MHD) equilibrium (which is calculated to reproduce the experimental data) to the reactor. At the same time, the MHD stability of the reactor plasma is also guaranteed to a certain extent since that beta was already proven in the experiment. The fusion gain in the helical type nuclear test machine FFHR-c1 has been evaluated using this modified DPE method. FFHR-c1 is basically a large duplication of the Large Helical Device (LHD) with a scale factor of 10/3, which corresponds to the major radius of the helical coils of 13.0 m and the plasma volume of ∼1000 m3. Two options with different magnetic field strengths are considered. The fusion gain in FFHR-c1 extrapolated from a set of radial profile data obtained in LHD ranges from 1 to 7, depending on the profiles used together with the assumptions of the magnetic field strength and the alpha heating efficiency.

  19. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    CERN Document Server

    Lee, Seung Kyu; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spectra with high radiation from a gamma-ray background. The signals from neutrons and gamma-rays are discriminated by the digital charge pulse shape discrimination (PSD) method, which uses total to partial charge ratio analysis. The signals are digitized by a flash analog-to-digital convertor (FADC). To evaluate the performance of the fabricated stilbene neutron diagnostic system, the efficiency of 10 mm soft-iron magnetic shielding and the detection efficiency of fast neutrons were tested experimentally using a 252Cf neutr...

  20. Fusion Ash Separation in the Princeton Field-Reversed Configuration Reactor

    Science.gov (United States)

    Abbate, Joseph; Yeh, Meagan; McGreivy, Nick; Cohen, Samuel

    2016-10-01

    The Princeton Field-Reversed Configuration (PFRC) concept relies on low-neutron production by D-3He fusion to enable small, safe nuclear-fusion reactors to be built, an approach requiring rapid and efficient extraction of fusion ash and energy produced by D-3He fusion reactions. The ash exhaust stream would contain energetic (0.1-1 MeV) protons, T, 3He, and 4He ions and nearly 1e5 cooler (ca. 100 eV) D ions. The T extracted from the reactor would be a valuable fusion product in that it decays into 3He, which could be used as fuel. If the T were not extracted it would be troublesome because of neutron production by the D-T reaction. This paper discusses methods to separate the various species in a PFRC reactor's exhaust stream. First, we discuss the use of curved magnetic fields to separate the energetic from the cool components. Then we discuss exploiting material properties, specifically reflection, sputtering threshold, and permeability, to allow separation of the hydrogen from the helium isotopes. DOE Contract Number DE-AC02-09CH11466.

  1. 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.

  2. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

  3. New concepts for shaftless recycle reactors

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.; Berty, I.J.

    1987-01-01

    Berty Reaction Engineers, Ltd. (BREL) is developing two new laboratory recycle reactors, the ROTOBERTY and the TURBOBERTY. These new reactors are basically improved versions of the original Berty reactor. To understand why the reactors have the features that they do, it is first necessary to briefly review laboratory reactors in general and specifically the original Berty reactor.

  4. Computer simulation of tritium releases in inertial fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perlado, J.M.; Velarde, M. [Universidad Politecnica de Madrid, Instituto de Fusion Nuclear, DENIM (Spain)

    2000-07-01

    Accidental releases of tritium from Inertial Fusion reactors are presented. A well-established computer code, MACCS2, is used with realistic models. Release fractions of 1 - 10 - 50 - 100 % of inventories are considered, with height of emissions 10, 30, 60 m, and duration of 10 min. and 2 hours. Only early emergency phase is considered with mitigative actions and shielding factors. It is concluded that except in 100 % releases for some reactors and heights the effective doses to workers and general population does not exceed the regulatory limits. Differences with very conservative results can attain 2 orders of magnitude. (authors)

  5. Concept of magnet systems for LHD-type reactor

    Science.gov (United States)

    Imagawa, S.; Takahata, K.; Tamura, H.; Yanagi, N.; Mito, T.; Obana, T.; Sagara, A.

    2009-07-01

    Heliotron reactors have attractive features for fusion power plants such as having no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered to be the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, a major radius of plasma of 14-17 m with a central toroidal field of 6-4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120-140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress are comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than the 1000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is more than 150 m, that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with a small extension of the ITER technology.

  6. 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.

  7. Hydrogen isotopes transport parameters in fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Serra, E. [Politecnico di Torino (Italy). Dipartimento di Energetica; Benamati, G. [ENEA Fusion Division, CR Brasimone, 40032 Camungnano, Bologna (Italy); Ogorodnikova, O.V. [Moscow State Engineering Physics Institute, Moscow 115409 (Russian Federation)

    1998-06-01

    This work presents a review of hydrogen isotopes-materials interactions in various materials of interest for fusion reactors. The relevant parameters cover mainly diffusivity, solubility, trap concentration and energy difference between trap and solution sites. The list of materials includes the martensitic steels (MANET, Batman and F82H-mod.), beryllium, aluminium, beryllium oxide, aluminium oxide, copper, tungsten and molybdenum. Some experimental work on the parameters that describe the surface effects is also mentioned. (orig.) 62 refs.

  8. Towards the detection of magnetohydrodynamics instabilities in a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sozzi, Carlo, E-mail: sozzi@ifp.cnr.it; Alessi, E., E-mail: sozzi@ifp.cnr.it; Figini, L., E-mail: sozzi@ifp.cnr.it; Galperti, G., E-mail: sozzi@ifp.cnr.it; Lazzaro, E., E-mail: sozzi@ifp.cnr.it; Marchetto, C., E-mail: sozzi@ifp.cnr.it; Nowak, S. [Istituto di Fisica del Plasma, CNR, EURATOM-ENEA Association, Milano (Italy); Mosconi, M. [Dipartimento di Energia, Politecnico di Milano, Milano (Italy)

    2014-08-21

    Various active control strategies of the Neoclassical tearing modes are being studied in present tokamaks using established detection techniques which exploit the measurements of the fluctuations of the magnetic field and of the electron temperature. The extrapolation of such techniques to the fusion reactor scale is made problematic by the neutron fluence and by the physics conditions related to the high plasma temperature and density which degrade the spatial resolution of such measurements.

  9. Ion cyclotron and lower hybrid arrays applicable to current drive in fusion reactors

    Science.gov (United States)

    Bosia, G.; Helou, W.; Goniche, M.; Hillaret, J.; Ragona, R.

    2014-02-01

    This paper presents concepts for Ion Cyclotron and Lower Hybrid Current Drive arrays applicable to fusion reactors and based on periodically loaded line power division. It is shown that, in large arrays, such as the ones proposed for fusion reactor applications, these schemes can offer, in principle, a number of practical advantages, compared with currently adopted ones, such as in-blanket operation at significantly reduced power density, lay out suitable for water cooling, single ended or balanced power feed, simple and load independent impedance matching In addition, a remote and accurate real time measurement of the complex impedance of all array elements as well as detection, location, and measurement of the complex admittance of a single arc occurring anywhere in the structure is possible.

  10. Physical and mechanical characteristics and chemical compatibility of aluminum nitride insulator coatings for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Rink, D.L. [Argonne National Lab., IL (United States). Energy Technology Div.

    1996-04-01

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of various concepts, including liquid metal, molten salt, water, and helium. Based on the requirements for an electrically insulating coating on the first-wall structural material to minimize the MHD pressure drop during the flow of liquid metal in a magnetic field, AlN was selected as a candidate coating material for the Li self-cooled blanket concept. This report discusses the results from an ongoing study of physical and mechanical characteristics and chemical compatibility of AlN electrical insulator coatings in a liquid Li environment. Details are presented on the AlN coating fabrication methods, and experimental data are reported for microstructures, chemistry of coatings, pretreatment of substrate, heat treatment of coatings, hardness data for coatings, coating/lithium interactions, and electrical resistance before and after exposure to lithium. Thermodynamic calculations are presented to establish regions of stability for AlN coatings in an Li environment as a function of O concentration and temperature, which can aid in-situ development of AlN coatings in Li.

  11. ASME Material Challenges for Advanced Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

  12. Individual dose due to radioactivity accidental release from fusion reactor.

    Science.gov (United States)

    Nie, Baojie; Ni, Muyi; Wei, Shiping

    2017-04-05

    As an important index shaping the design of fusion safety system, evaluation of public radiation consequences have risen as a hot topic on the way to develop fusion energy. In this work, the comprehensive public early dose was evaluated due to unit gram tritium (HT/HTO), activated dust, activated corrosion products (ACPs) and activated gases accidental release from ITER like fusion reactor. Meanwhile, considering that we cannot completely eliminate the occurrence likelihood of multi-failure of vacuum vessel and tokamak building, we conservatively evaluated the public radiation consequences and environment restoration after the worst hypothetical accident preliminarily. The comparison results show early dose of different unit radioactivity release under different conditions. After further performing the radiation consequences, we find it possible that the hypothetical accident for ITER like fusion reactor would result in a level 6 accident according to INES, not appear level 7 like Chernobyl or Fukushima accidents. And from the point of environment restoration, we need at least 69 years for case 1 (1kg HTO and 1000kg dust release) and 34-52years for case 2 (1kg HTO and 10kg-100kg dust release) to wait the contaminated zone drop below the general public safety limit (1mSv per year) before it is suitable for human habitation.

  13. Material options for a commercial fusion reactor first wall

    Energy Technology Data Exchange (ETDEWEB)

    Dabiri, A.E.

    1986-05-01

    A study has been conducted to evaluate the potential of various materials for use as first walls in high-power-density commercial fusion reactors. Operating limits for each material were obtained based on a number of criteria, including maximum allowable structural temperatures, critical heat flux, ultimate tensile strength, and design-allowable stress. The results with water as a coolant indicate that a modified alloy similar to HT-9 may be a suitable candidate for low- and medium-power-density reactor first walls with neutron loads of up to 6 MW/m/sup 2/. A vanadium or copper alloy must be used for high-power-density reactors. The neutron wall load limit for vanadium alloys is about 14 MW/sup 2/, provided a suitable coating material is chosen. The extremely limited data base for radiation effects hinders any quantitative assessment of the limits for copper alloys.

  14. Integrated simulation and modeling capability for alternate magnetic fusion concepts

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, B. I.; Hooper, E.B.; Jarboe, T. R.; LoDestro, L. L.; Pearlstein, L. D.; Prager, S. C.; Sarff, J. S.

    1998-11-03

    This document summarizes a strategic study addressing the development of a comprehensive modeling and simulation capability for magnetic fusion experiments with particular emphasis on devices that are alternatives to the mainline tokamak device. A code development project in this area supports two defined strategic thrust areas in the Magnetic Fusion Energy Program: (1) comprehensive simulation and modeling of magnetic fusion experiments and (2) development, operation, and modeling of magnetic fusion alternate- concept experiment

  15. Effect of Fuelling Depth on the Fusion Performance and Particle Confinement of a Fusion Reactor

    Science.gov (United States)

    Wang, Shijia; Wang, Shaojie

    2016-12-01

    The fusion performance and particle confinement of an international thermonuclear experimental reactor (ITER)-like fusion device have been modeled by numerically solving the energy transport equation and the particle transport equation. The effect of fuelling depth has been investigated. The plasma is primarily heated by the fusion produced alpha particles and the loss process of particles and energy in the scrape-off layer has been taken into account. To study the effect of fuelling depth on fusion performance, the ITERH-98P(y,2) scaling law has been used to evaluate the transport coefficients. It is shown that the particle confinement and fusion performance are significantly dependent on the fuelling depth. Deviation of 10% of the minor radius on fuelling depth can make the particle confinement change by ∼ 61% and the fusion performance change by ∼ 108%. The enhancement of fusion performance is due to the better particle confinement induced by deeper particle fuelling. supported by National Natural Science Foundation of China (Nos. 11175178 and 11375196) and the National Magnetic Confinement Fusion Science Program of China (No. 2014GB113000)

  16. New reactor concepts; Nieuwe rectorconcepten - nouveaux reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A

    1998-11-01

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost.

  17. Repair welding of fusion reactor components. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Chin, B.A.; Wang, C.A.

    1997-09-30

    The exposure of metallic materials, such as structural components of the first wall and blanket of a fusion reactor, to neutron irradiation will induce changes in both the material composition and microstructure. Along with these changes can come a corresponding deterioration in mechanical properties resulting in premature failure. It is, therefore, essential to expect that the repair and replacement of the degraded components will be necessary. Such repairs may require the joining of irradiated materials through the use of fusion welding processes. The present ITER (International Thermonuclear Experimental Reactor) conceptual design is anticipated to have about 5 km of longitudinal welds and ten thousand pipe butt welds in the blanket structure. A recent study by Buende et al. predict that a failure is most likely to occur in a weld. The study is based on data from other large structures, particularly nuclear reactors. The data used also appear to be consistent with the operating experience of the Fast Flux Test Facility (FFTF). This reactor has a fuel pin area comparable with the area of the ITER first wall and has experienced one unanticipated fuel pin failure after two years of operation. The repair of irradiated structures using fusion welding will be difficult due to the entrapped helium. Due to its extremely low solubility in metals, helium will diffuse and agglomerate to form helium bubbles after being trapped at point defects, dislocations, and grain boundaries. Welding of neutron-irradiated type 304 stainless steels has been reported with varying degree of heat-affected zone cracking (HAZ). The objectives of this study were to determine the threshold helium concentrations required to cause HAZ cracking and to investigate techniques that might be used to eliminate the HAZ cracking in welding of helium-containing materials.

  18. Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

    Science.gov (United States)

    Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

    2017-01-01

    The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

  19. Synfuels from fusion: producing hydrogen with the Tandem Mirror Reactor and thermochemical cycles

    Energy Technology Data Exchange (ETDEWEB)

    Werner, R.W.; Ribe, F.L.

    1981-01-21

    This volume contains the following sections: (1) the Tandem Mirror fusion driver, (2) the Cauldron blanket module, (3) the flowing microsphere, (4) coupling the reactor to the process, (5) the thermochemical cycles, and (6) chemical reactors and process units. (MOW)

  20. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    Science.gov (United States)

    Was, Gary S.

    2007-08-01

    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems.

  1. 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.

  2. Advanced Fusion Reactors for Space Propulsion and Power Systems

    Science.gov (United States)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  3. 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.

  4. Development of fusion blanket technology for the DEMO reactor.

    Science.gov (United States)

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. Copyright © 2011 Elsevier Ltd. All rights reserved.

  5. Final optics for laser-driven inertial fusion reactors

    Science.gov (United States)

    Woodworth, J. G.; Chase, L. L.; Guinan, M. W.; Krupke, W. F.; Sooy, W. R.

    1991-10-01

    If Inertial Confinement Fusion (ICF) power plus utilizing laser drivers are to be considered for electrical power generation, a method for delivering the driver energy into the reactor must be developed. This driver-reactor interface will necessarily employ 'final optics,' which must survive in the face of fast neutrons, x rays, hot vapors and condensates, and high speed droplets. The most difficult to protect against is fast neutron damage since no optically transmissive shielding material for 14 MeV neutrons is available. Multilayer dielectric mirrors are judged to be unsuitable because radiation induced chemical change, diffusion, and thickness changes will destroy their reflectivity within a few months of plant operation. Recently, grazing incidence metal mirrors were proposed, but optical damage issues are unresolved for this approach. In this study, we considered the use of refractive optics. A baseline design consists of two wedges of fused silica, which put a dogleg into the beam and thus remove optics further upstream from direct sight of the reactor. If the closest optic were located 40 m from the center of a 3 GW sub t reactor it would be subject to an average 14 MeV neutron flux of approx. 5 x 10(exp 12) n/sq cm with a peak flux of approx. 6 x 10(exp 18) n/sq cm. A major question to be answered is: 'what duration of reactor operation can this optic withstand'. To answer this question we have reviewed the literature bearing on radiation induced optical damage in fused silica and assessed its implications for reactor operation with the baseline final optics scheme. It appears possible to continuously anneal the neutron damage in the silica by keeping the wedge at a modestly elevated temperature.

  6. High conductivity Be-Cu alloys for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lilley, E.A. [NGK Metals Corp., Reading, PA (United States); Adachi, Takao; Ishibashi, Yoshiki [NGK Insulators, Ltd., Aichi-ken (Japan)

    1995-09-01

    The optimum material has not yet been identified. This will result in heat from plasma to the first wall and divertor. That is, because of cracks and melting by thermal power and shock. Today, it is considered to be some kinds of copper, alloys, however, for using, it must have high conductivity. And it is also needed another property, for example, high strength and so on. We have developed some new beryllium copper alloys with high conductivity, high strength, and high endurance. Therefore, we are introducing these new alloys as suitable materials for the heat sink in fusion reactors.

  7. 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.

  8. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  9. Thermionic plasma injection for the Lockheed Martin T4 Compact Fusion Reactor experiment

    Science.gov (United States)

    Heinrich, Jonathon

    2015-11-01

    Lockheed Martin's Compact Fusion Reactor (CFR) concept relies on diamagnetic confinement in a magnetically encapsulated linear ring cusp geometry. Plasma injection into cusp field configurations requires careful deliberation. Previous work has shown that axial injection via a plasma gun is capable of achieving high-beta conditions in cusp configurations. We present a pulsed, high power thermionic plasma source and the associated magnetic field topology for plasma injection into the caulked-cusp magnetic field. The resulting plasma fueling and cross-field diffusion is discussed.

  10. Demountable Toroidal Field Magnets for Use in a Compact Modular Fusion Reactor

    Science.gov (United States)

    Mangiarotti, F. J.; Goh, J.; Takayasu, M.; Bromberg, L.; Minervini, J. V.; Whyte, D.

    2014-05-01

    A concept of demountable toroidal field magnets for a compact fusion reactor is discussed. The magnets generate a magnetic field of 9.2 T on axis, in a 3.3 m major radius tokamak. Subcooled YBCO conductors have a critical current density adequate to provide this large magnetic field, while operating at 20 K reduces thermodynamic cooling cost of the resistive electrical joints. Demountable magnets allow for vertical replacement and maintenance of internal components, potentially reducing cost and time of maintenance when compared to traditional sector maintenance. Preliminary measurements of contact resistance of a demountable YBCO electrical joint between are presented.

  11. The TITAN Reversed-Field Pinch fusion reactor study: Scoping phase report

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    The TITAN research program is a multi-institutional effort to determine the potential of the Reversed-Field Pinch (RFP) magnetic fusion concept as a compact, high-power-density, and ''attractive'' fusion energy system from economic (cost of electricity, COE), environmental, and operational viewpoints. In particular, a high neutron wall loading design (18 MW/m/sup 2/) has been chosen as the reference case in order to quantify the issue of engineering practicality, to determine the physics requirements and plasma operating mode, to assess significant benefits of compact systems, and to illuminate the main drawbacks. The program has been divided into two phases, each roughly one year in length: the Scoping Phase and the Design Phase. During the scoping phase, the TITAN design team has defined the parameter space for a high mass power density (MPD) RFP reactor, and explored a variety of approaches to the design of major subsystems. Two major design approaches consistent with high MPD and low COE, the lithium-vanadium blanket design and aqueous loop-in-pool design, have been selected for more detailed engineering evaluation in the design phase. The program has retained a balance in its approach to investigating high MPD systems. On the one hand, parametric investigations of both subsystems and overall system performance are carried out. On the other hand, more detailed analysis and engineering design and integration are performed, appropriate to determining the technical feasibility of the high MPD approach to RFP fusion reactors. This report describes the work of the scoping phase activities of the TITAN program. A synopsis of the principal technical findings and a brief description of the TITAN multiple-design approach is given. The individual chapters on Plasma Physics and Engineering, Parameter Systems Studies, Divertor, Reactor Engineering, and Fusion Power Core Engineering have been cataloged separately.

  12. A Fusion Reactor Design with a Liquid First Wall and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Nygren, R E; Rognlien, T D; Rensink, M E; Smolentsev, S S; Youssef, M E; Sawan, M Z; Merrill, B J; Eberle, C; Fogarty, P J; Nelson, B E; Sze, D K; Majeski, R

    2003-11-13

    Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840MW of fusion power of which 767MW is in the form of energetic particles (alpha power) and 3073MW is in the form of neutrons. The alpha plus auxiliary power total 909MW of which 430MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.

  13. The properties and weldability of materials for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Chin, B.A.; Kee, C.K.; Wilcox, R.C. [Auburn Univ., AL (United States). Dept. of Mechanical Engineering; Zinkle, S.J. [Oak Ridge National Lab., TN (United States)

    1991-11-15

    Low-activation austenitic stainless steels have been suggested for applications within fusion reactors. The use of these nickel-free steels will help to reduce the radioactive waste management problem after service. one requirement for such steels is the ability to obtain sound welds for fabrication purposes. Thus, two austenitic Fe-Cr-Mn alloys were studied to characterize the welded microstructure and mechanical properties. The two steels investigated were a Russian steel (Fe-11.6Cr19.3Mn-0.181C) and an US steel (Fe-12.lCr-19.4Mn-0.24C). Welding was performed using a gas tungsten arc welding (GTAW) process. Microscopic examinations of the structure of both steels were conducted. The as-received Russian steel was found to be in the annealed state. Only the fusion zone and the base metal were observed in the welded Russian steel. No visible heat affected zone was observed. Examination revealed that the as-received US steel was in the cold rolled condition. After welding, a fusion zone and a heat affected zone along with the base metal region were found.

  14. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  15. Material Science Activities for Fusion Reactors in Kazakhstan

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibayeva, I.; Kenzhin, E.; Kulsartov, T. [Institute of Atomic Energy NNC RK, Kurchatov (Kazakhstan); Shestakov, V. [Kazakhstan State University, Almaty (Kazakhstan); Chikhray, Y. [Kazakh National University, Kourmangazy 15, app.lO, 480100 Almaty (Kazakhstan); Azizov, E. [TRINITI, Troitsk (Russian Federation); Filatov, O. [Effremov Institute, Saint Petersburg (Russian Federation); Chernov, V.M. [Bochvar Institute of Inorganic Materials, P.O. Box 369, 123060 Moscow (Russian Federation)

    2007-07-01

    Full text of publication follows: Paper contains results of fusion material testing national program and results of activities on creation of material testing spherical tokamak. Hydrogen isotope behavior (diffusion, permeation, and accumulation) in the components of the first wall and divertor was studied taking into account temperature, pressure, and reactor irradiation. There were carried out out-of-pile and in-pile (reactors IVG-IM, WWRK, RA) studies of beryllium of various grades (TV-56, TShG-56, DV-56, TGP-56, TIP-56), graphites (RG-T, MPG-8, FP 479, R 4340), molybdenum, tungsten, steels (Cr18Ni10Ti, Cr16Ni15, MANET, F82H), alloys V-(4-6)Cr-( 4-5)Ti, Cu+1%Cr+0.1%Zr, and double Be/Cu and triple Be/Cu/steel structures. Tritium permeability from eutectic Pb+17%Li through steels Cr18Ni10Ti, Cr16Ni15, MANET, and F82H were studied taking into account protective coating effects. The tritium production rate was experimentally assessed during in-pile and post-reactor experiments. There were carried out radiation tests of ceramic Li{sub 2}TiO{sub 3} (96% enrichment by Li-6) with in-situ registration of released tritium and following post-irradiation material tests of irradiated samples. Verification of computer codes for simulation of accidents related to LOCA in ITER reactor was carried out. Codes' verification was carried out for a mockup of first wall in a form of three-layer cylinder of beryllium, bronze (Cu-Cr-Zr) and stainless steel. At present Kazakhstan Tokamak for Material testing (tokamak KTM) is created in National Nuclear Center of Republic of Kazakhstan in cooperation with Russian Federation organizations (start-up is scheduled on 2008). Tokamak KTM allows for expansion and specification of the studies and tests of materials, protection options of first wall, receiving divertor tiles and divertor components, methods for load reduction at divertor, and various options of heat/power removal, fast evacuation of divertor volume and development of the

  16. Fusion reactor design-III. Report on the third IAEA technical committee meeting and workshop, Tokyo, Japan, 5-16 October 1981

    Energy Technology Data Exchange (ETDEWEB)

    Iso, Y. (Japan Atomic Energy Research Inst., Tokyo); Stacey, W.M. Jr. (Georgia Inst. of Tech., Atlanta (USA)); Kulcinski, G.L. (Wisconsin Univ., Madison (USA)); Krakowski, R.A. (Los Alamos National Lab., NM (USA)); Carlson, G.A. (Lawrence Livermore National Lab., CA (USA)); Yamanaka, C. (Osaka Univ., Suita, Japan); Casini, G. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Igata, N. (Tokyo Univ. (Japan))

    1982-05-01

    A brief summary is given of the plenary sessions of the Third IAEA Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology. The large tokamak experiments now under construction have brought fusion research to the threshold of fusion reactor power. A number of major in-depth reactor design studies are reported such as the 650-MW INTOR near-term experimental fusion reactor based upon essentially current technology. Also reported are studies of commercial reactor designs. Results are given of the discussions of seven workshops that were based on the papers presented. These groups evaluated the current status and identified key problem areas for each in the following areas: near-term tokamaks, long-term tokamaks, toroidal systems (alternative concepts), open systems, inertial confinement systems (advanced fuels, hybrids, etc.), and fusion reactor materials. The importance of beginning now to prepare for the next major step in fusion reactor development was emphasized and the benefits of international co-operation were evident in the consensus reached in the INTOR results and the strong influence it has had on the direction of the leading national design efforts.

  17. Design of a fusion reactor for eutectic alloys Pb- Li; Diseno de un reactor de fusion para aleaciones autecticas Pb-Li

    Energy Technology Data Exchange (ETDEWEB)

    Quinones, J.; Barrena Perez, M. I.; Gomez de Salazar, J. M.; Serrano, L.; Duran, S.; Conde, E.; Barrado, A. I.; Fernandez, M.; Sedano, L.; Soria Munoz, A.

    2010-07-01

    Given the interest that have the Pb-Li eutectic alloys in the field of production of tritium, designed to optimize energy through nuclear fusion processes, in this paper, we present the design, construction and commissioning of a fusion reactor of Pb-Li alloys eutectic and the optimal process conditions, for these alloys.

  18. The hybrid reactor project based on the straight field line mirror concept

    Science.gov (United States)

    Ågren, O.; Noack, K.; Moiseenko, V. E.; Hagnestâl, A.; Källne, J.; Anglart, H.

    2012-06-01

    The straight field line mirror (SFLM) concept is aiming towards a steady-state compact fusion neutron source. Besides the possibility for steady state operation for a year or more, the geometry is chosen to avoid high loads on materials and plasma facing components. A comparatively small fusion hybrid device with "semi-poor" plasma confinement (with a low fusion Q factor) may be developed for industrial transmutation and energy production from spent nuclear fuel. This opportunity arises from a large fission to fusion energy multiplication ratio, Qr = Pfis/Pfus>>1. The upper bound on Qr is primarily determined by geometry and reactor safety. For the SFLM, the upper bound is Qr≈150, corresponding to a neutron multiplicity of keff=0.97. Power production in a mirror hybrid is predicted for a substantially lower electron temperature than the requirement Te≈10 keV for a fusion reactor. Power production in the SFLM seems possible with Q≈0.15, which is 10 times lower than typically anticipated for hybrids (and 100 times smaller than required for a fusion reactor). This relaxes plasma confinement demands, and broadens the range for use of plasmas with supra-thermal ions in hybrid reactors. The SFLM concept is based on a mirror machine stabilized by qudrupolar magnetic fields and large expander tanks beyond the confinement region. The purpose of the expander tanks is to distribute axial plasma loss flow over a sufficiently large area so that the receiving plates can withstand the heat. Plasma stability is not relying on a plasma flow into the expander regions. With a suppressed plasma flow into the expander tanks, a possibility arise for higher electron temperature. A brief presentation will be given on basic theory for the SFLM with plasma stability and electron temperature issues, RF heating computations with sloshing ion formation, neutron transport computations with reactor safety margins and material load estimates, magnetic coil designs as well as a discussion on

  19. Laser Fusion - A New Thermonuclear Concept

    Science.gov (United States)

    Cooper, Ralph S.

    1975-01-01

    Describes thermonuclear processes induced by interaction of a laser beam with the surface of a fuel pellet. An expanding plasma is formed which results in compression of the element. Laser and reactor technology are discussed. Pictures and diagrams are included. (GH)

  20. Radiation Hydrodynamic Parameter Study of Inertial Fusion Energy Reactor Chambers

    Science.gov (United States)

    Sacks, Ryan; Moses, Gregory

    2014-10-01

    Inertial fusion energy reactors present great promise for the future as they are capable of providing baseline power with no carbon footprint. Simulation work regarding the chamber response and first wall insult is performed with the 1-D radiation hydrodynamics code BUCKY. Simulation with differing chamber parameters are implemented to study the effect of gas fill, gas mixtures and chamber radii. Xenon and argon gases are of particular interest as shielding for the first wall due to their high opacity values and ready availability. Mixing of the two gases is an attempt to engineer a gas cocktail to provide the maximum amount of shielding with the least amount of cost. A parameter study of different chamber radii shows a consistent relationship with that of first wall temperature (~1/r2) and overpressure (~1/r3). This work is performed under collaboration with Lawrence Livermore National Laboratory.

  1. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    Directory of Open Access Journals (Sweden)

    Singh Vishvanath P.

    2016-01-01

    Full Text Available Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fitting at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neutron removal cross-section. It is shown that ZrH2 and VH2 are very good shielding materials for gamma rays and fast neutrons due to their suitable combination of low- and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors.

  2. High temperature indentation tests on fusion reactor candidate materials

    Energy Technology Data Exchange (ETDEWEB)

    Montanari, R. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy)]. E-mail: roberto.montanari@uniroma2.it; Filacchioni, G. [ENEA CR Casaccia, Via Anguillarese 301, I-00060 S.M. di Galeria, Rome (Italy); Iacovone, B. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Plini, P. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Riccardi, B. [Associazione EURATOM-ENEA sulla Fusione, P.O. Box 65, I-00044 Frascati, Rome (Italy)

    2007-08-01

    Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure-penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from -180 to +200 {sup o}C. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 {sup o}C. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress-relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods.

  3. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  4. Radioactivity effects of Pb-17Li in fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rocco, P. (Commission of the European Communities, Joint Research Centre, Ispra (Italy)); Zucchetti, M. (Dipt. di Energetica, Politecnico Turin (Italy))

    1991-12-01

    Research on the eutectic Pb-17Li is part of the blanket studies carried out in Europe for fusion power reactors. The use of this breeder makes easier some safety problems as compared to the case of lithium as a consequence of the lower chemical reactivity of Pb-17Li. On the other hand, it increases the radioactivity problems due to the neutron activation of lead and impurities. This paper presents both short-term (accidents) and long-term (waste disposal and recycling) aspects of the Pb-17Li activation products. They include the production, mobilization, release and environmental impact. Concerning accidents, a particular attention is given to Po-210 and Hg-203. Questions related to waste management are also revised. The most attractive solution seems that of recycling the spent Pb-17Li. This will be possible about 20 y after removal from service. As an alternative to recycling, the breeder disposal as radioactive waste is discussed. (orig.).

  5. Simulation of the ash exhaust in a fusion engineering reactor

    Energy Technology Data Exchange (ETDEWEB)

    Itoh, Kimitaka; Ueda, Noriaki; Itoh, Sanae; Sugihara, Masayoshi.

    1988-09-01

    Numerical analysis of the plasmas in the divertor and scrape-off layer (SOL) of a fusion engineering reactor (FER) is made by using the two-dimensional time-dependent simulation code. The equation for plasmas in the SOL and divertor regions is solved for the given particle and heat sources from the main plasma, GAMMA/sub out/ and Q/sub T/. For the given size of the pumping duct and the pumping speed, the necessary value of GAMMA/sub out/ for the ash exhaust is discussed. Sufficient ash exhaust is possible under the condition which is consistent with the scaling law of the particle and energy confinement of the main plasma.

  6. Fast track concept in the European fusion programme

    Energy Technology Data Exchange (ETDEWEB)

    Bolt, Harald [Max-Planck-Institute for Plasma Physics, Garching (Germany)

    2002-10-01

    Recently an expert meeting regarding a possible acceleration of the fusion programme with a view to energy production ('fast track') was held on the initiative of the EU Research Council. In the course of the discussions about the fast track programme, it has turned out that successful extraction of reactor grade heat and tritium from the blanket modules is essential in the ITER operation. In parallel with ITER, material development using a high intensity neutron source is essential to establish a database for licensing. The operation of a reactor combining DEMO and PROTO generations into a single step could be around 2030. (author)

  7. Flibe use in fusion reactors -- An initial safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, L.C.; Longhurst, G.R.

    1999-03-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF{sub 2}) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  8. Flibe Use in Fusion Reactors - An Initial Safety Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee Charles; Longhurst, Glen Reed

    1999-04-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF2) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  9. Power conversion systems based on Brayton cycles for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Linares, J.I., E-mail: linares@upcomillas.es [Rafael Marino Chair on New Energy Technologies. Comillas Pontifical University, Alberto Aguilera, 25-28015 Madrid (Spain); Herranz, L.E. [Unit of Nuclear Safety Research. CIEMAT, Madrid (Spain); Moratilla, B.Y.; Serrano, I.P. [Rafael Marino Chair on New Energy Technologies. Comillas Pontifical University, Alberto Aguilera, 25-28015 Madrid (Spain)

    2011-10-15

    This paper investigates Brayton power cycles for fusion reactors. Two working fluids have been explored: helium in classical configurations and CO{sub 2} in recompression layouts (Feher cycle). Typical recuperator arrangements in both cycles have been strongly constrained by low temperature of some of the energy thermal sources from the reactor. This limitation has been overcome in two ways: with a combined architecture and with dual cycles. Combined architecture couples the Brayton cycle with a Rankine one capable of taking advantage of the thermal energy content of the working fluid after exiting the turbine stage (iso-butane and steam fitted best the conditions of the He and CO{sub 2} cycles, respectively). Dual cycles set a specific Rankine cycle to exploit the lowest quality thermal energy source, allowing usual recuperator arrangements in the Brayton cycle. The results of the analyses indicate that dual cycles could reach thermal efficiencies around 42.8% when using helium, whereas thermal performance might be even better (46.7%), if a combined CO{sub 2}-H{sub 2}O cycle was set.

  10. Metabolic and environmental aspects of fusion reactor activation products: niobium

    Energy Technology Data Exchange (ETDEWEB)

    Easterly, C.E.; Shank, K.E.

    1977-11-01

    A summary of the metabolic and environmental aspects of niobium is presented. The toxicological symptoms from exposure to niobium are given, along with lethal concentration values for acute and chronic exposures. Existing human data are presented; animal uptake and retention data are analyzed for various routes of administration. Recommended metabolic values are also presented along with comments concerning their use and appropriateness. The natural distribution of niobium is given for freshwater, seawater, and the biosphere. Concentration factors and retention of /sup 95/Nb in the environment are discussed with reference to: plant retention via leaf absorption; plant retention via root uptake; uptake in terrestrial animals from plants; uptake in freshwater organisms; uptake in marine organisms; and movement in soil. Conclusions are drawn regarding needs for future work in these areas. This review was undertaken because niobium is expected to be a key metal in the development of commercial fusion reactors. It is recognized that niobium will likely not be used in the first generation reactors as a structural material but will appear as an alloy in such materials as superconducting wire.

  11. Elevator mode convection in liquid metal blankets for fusion reactors

    Science.gov (United States)

    Zikanov, Oleg; Liu, Li

    2015-11-01

    The work is motivated by the design of liquid-metal blankets for nuclear fusion reactors. Mixed convection in a downward flow in a vertical duct with strong contant-rate heating of one wall (the Grashof number up to 1012) and strong transverse magnetic field (the Hartmann number up to 104) is considered. It is found that in an infinitely long duct the flow is dominated by exponentially growing elevator modes having the form of a combination of ascending and descending jets. An analytical solution approximating the growth rate of the modes is derived. Analogous flows in finite-length pipes and ducts are analyzed using the high-resolution numerical simulations. The results of the recent experiments are reproduced and explained. It is found that the flow evolves in cycles consisting of periods of exponential growth and breakdowns of the jets. The resulting high-amplitude fluctuations of temperature is a feature potentially dangerous for operation of a reactor blanket. Financial support was provided by the US NSF (Grant CBET 1232851).

  12. Development of heat sink concept for near-term fusion power plant divertor

    Science.gov (United States)

    Rimza, Sandeep; Khirwadkar, Samir; Velusamy, Karupanna

    2017-04-01

    Development of an efficient divertor concept is an important task to meet in the scenario of the future fusion power plant. The divertor, which is a vital part of the reactor has to discharge the considerable fraction of the total fusion thermal power (∼15%). Therefore, it has to survive very high thermal fluxes (∼10 MW/m2). In the present paper, an efficient divertor heat exchanger cooled by helium is proposed for the fusion tokamak. The Plasma facing surface of divertor made-up of several modules to overcome the stresses caused by high heat flux. The thermal hydraulic performance of one such module is numerically investigated in the present work. The result shows that the proposed design is capable of handling target heat flux values of 10 MW/m2. The computational model has been validated against high-heat flux experiments and a satisfactory agreement is noticed between the present simulation and the reported results.

  13. High temperature superconductor cable concepts for fusion magnets

    CERN Document Server

    AUTHOR|(CDS)2078397

    2013-01-01

    Three concepts of high temperature superconductor cables carrying kA currents (RACC, CORC and TSTC) are investigated, optimized and evaluated in the scope of their applicability as conductor in fusion magnets. The magnetic field and temperature dependence of the cables is measured; the thermal expansion and conductivity of structure, insulation and filling materials are investigated. High temperature superconductor winding packs for fusion magnets are calculated and compared with corresponding low temperature superconductor cases.

  14. Status and Prospects of the Fast Ignition Inertial Fusion Concept

    Energy Technology Data Exchange (ETDEWEB)

    Key, M H

    2006-11-15

    Fast ignition is an alternate concept in inertial confinement fusion, which has the potential for easier ignition and greater energy multiplication. If realized it could improve the prospects for inertial fusion energy. It poses stimulating challenges in science and technology and the research is approaching a key stage in which the feasibility of fast ignition will be determined. This review covers the concepts, the state of the science and technology, the near term prospects and the challenges and risks involved in demonstrating high gain fast ignition.

  15. Inertial Fusion Power Plant Concept of Operations and Maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Anklam, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Knutson, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunne, A. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kasper, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sheehan, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Lang, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Roberts, V. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mau, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-01-15

    Parsons and LLNL scientists and engineers performed design and engineering work for power plant pre-conceptual designs based on the anticipated laser fusion demonstrations at the National Ignition Facility (NIF). Work included identifying concepts of operations and maintenance (O&M) and associated requirements relevant to fusion power plant systems analysis. A laser fusion power plant would incorporate a large process and power conversion facility with a laser system and fusion engine serving as the heat source, based in part on some of the systems and technologies advanced at NIF. Process operations would be similar in scope to those used in chemical, oil refinery, and nuclear waste processing facilities, while power conversion operations would be similar to those used in commercial thermal power plants. While some aspects of the tritium fuel cycle can be based on existing technologies, many aspects of a laser fusion power plant presents several important and unique O&M requirements that demand new solutions. For example, onsite recovery of tritium; unique remote material handling systems for use in areas with high radiation, radioactive materials, or high temperatures; a five-year fusion engine target chamber replacement cycle with other annual and multi-year cycles anticipated for major maintenance of other systems, structures, and components (SSC); and unique SSC for fusion target waste recycling streams. This paper describes fusion power plant O&M concepts and requirements, how O&M requirements could be met in design, and how basic organizational and planning issues can be addressed for a safe, reliable, economic, and feasible fusion power plant.

  16. Inertial fusion power plant concept of operations and maintenance

    Science.gov (United States)

    Knutson, Brad; Dunne, Mike; Kasper, Jack; Sheehan, Timothy; Lang, Dwight; Anklam, Tom; Roberts, Valerie; Mau, Derek

    2015-02-01

    Parsons and LLNL scientists and engineers performed design and engineering work for power plant pre-conceptual designs based on the anticipated laser fusion demonstrations at the National Ignition Facility (NIF). Work included identifying concepts of operations and maintenance (O&M) and associated requirements relevant to fusion power plant systems analysis. A laser fusion power plant would incorporate a large process and power conversion facility with a laser system and fusion engine serving as the heat source, based in part on some of the systems and technologies advanced at NIF. Process operations would be similar in scope to those used in chemical, oil refinery, and nuclear waste processing facilities, while power conversion operations would be similar to those used in commercial thermal power plants. While some aspects of the tritium fuel cycle can be based on existing technologies, many aspects of a laser fusion power plant presents several important and unique O&M requirements that demand new solutions. For example, onsite recovery of tritium; unique remote material handling systems for use in areas with high radiation, radioactive materials, or high temperatures; a five-year fusion engine target chamber replacement cycle with other annual and multi-year cycles anticipated for major maintenance of other systems, structures, and components (SSC); and unique SSC for fusion target waste recycling streams. This paper describes fusion power plant O&M concepts and requirements, how O&M requirements could be met in design, and how basic organizational and planning issues can be addressed for a safe, reliable, economic, and feasible fusion power plant.

  17. Overview of Indian activities on fusion reactor materials

    Science.gov (United States)

    Banerjee, Srikumar

    2014-12-01

    This paper on overview of Indian activities on fusion reactor materials describes in brief the efforts India has made to develop materials for the first wall of a tokamak, its blanket and superconducting magnet coils. Through a systematic and scientific approach, India has developed and commercially produced reduced activation ferritic/martensitic (RAFM) steel that is comparable to Eurofer 97. Powder of low activation ferritic/martensitic oxide dispersion strengthened steel with characteristics desired for its application in the first wall of a tokamak has been produced on the laboratory scale. V-4Cr-4Ti alloy was also prepared in the laboratory, and kinetics of hydrogen absorption in this was investigated. Cu-1 wt%Cr-0.1 wt%Zr - an alloy meant for use as heat transfer elements for hypervapotrons and heat sink for the first wall - was developed and characterized in detail for its aging behavior. The role of addition of a small quantity of Zr in its improved fatigue performance was delineated, and its diffusion bonding with both W and stainless steel was achieved using Ni as an interlayer. The alloy was produced in large quantities and used for manufacturing both the heat transfer elements and components for the International Thermonuclear Experimental Reactor (ITER). India has proposed to install and test a lead-lithium cooled ceramic breeder test blanket module (LLCB-TBM) at ITER. To meet this objective, efforts have been made to produce and characterize Li2TiO3 pebbles, and also improve the thermal conductivity of packed beds of these pebbles. Liquid metal loops have been set up and corrosion behavior of RAFM steel in flowing Pb-Li eutectic has been studied in the presence as well as absence of magnetic fields. To prevent permeation of tritium and reduce the magneto-hydro-dynamic drag, processes have been developed for coating alumina on RAFM steel. Apart from these activities, different approaches being attempted to make the U-shaped first wall of the TBM box

  18. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor

    Science.gov (United States)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed

    2009-11-01

    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.

  19. Prospects of steady state magnetic diagnostic of fusion reactors based on metallic Hall sensors

    Science.gov (United States)

    Ďuran, I.; Sentkerestiová, J.; Kovařík, K.; Viererbl, L.

    2012-06-01

    Employment of sensors based on Hall effect (Hall sensors) is one of the candidate approaches to detection of almost steady state magnetic fields in future fusion reactors based on magnetic confinement (tokamaks, stellarators etc.), and also in possible fusion-fission hybrid systems having these fusion reactors as a neutron source and driver. This contribution reviews the initial considerations concerning application of metallic Hall sensors in fusion reactor harsh environment that include high neutron loads (>1018 cm-2) and elevated temperatures (>200°C). In particular, the candidate sensing materials, candidate technologies for sensors production, initial analysis of activation and transmutation of sensors under reactor relevant neutron loads and the tests of the the first samples of copper Hall sensors are presented.

  20. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

    Energy Technology Data Exchange (ETDEWEB)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. (eds.)

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  1. Advanced Nuclear Reactor Concepts for China

    Energy Technology Data Exchange (ETDEWEB)

    Knoche, D.; Sassen, F.; Tietsch, W. [Westinghouse Electric Germany, Postfach 10 05 63, 68140 Mannheim (Germany); Yujie, Dong; Li, Cao [INET, Tsinghua University, 100084 Beijing (China)

    2008-07-01

    China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

  2. Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

    Energy Technology Data Exchange (ETDEWEB)

    Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Jaworski, M.A.; Kaita, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Hirooka, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Gray, T.K. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2017-04-15

    Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety concerns (Federici et al., 2001) . It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance (Ono et al., 2013, 2014) . The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium (LL) divertor (RLLD) concept (Ono et al., 2013) and its variant, the active liquid lithium divertor concept (ARLLD) (Ono et al., 2014) , taking advantage of the enhanced non-coronal Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/s of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤450 °C than the first wall ∼600–700 °C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ∼1 l/s (l/s) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust/impurities are removed by relatively simple filter and cold/hot trap systems. Using a

  3. Concept of Operations for Data Fusion Visualization

    Energy Technology Data Exchange (ETDEWEB)

    T.R. McJunkin; R.L. Boring; M.A. McQueen; L.P. Shunn; J.L. Wright; D.I. Gertman; O. Linda; K. McCarty; M. Manic

    2011-09-01

    Situational awareness in the operations and supervision of a industrial system means that decision making entity, whether machine or human, have the important data presented in a timely manner. An optimal presentation of information such that the operator has the best opportunity accurately interpret and react to anomalies due to system degradation, failures or adversaries. Anticipated problems are a matter for system design; however, the paper will focus on concepts for situational awareness enhancement for a human operator when the unanticipated or unaddressed event types occur. Methodology for human machine interface development and refinement strategy is described for a synthetic fuels plant model. A novel concept for adaptively highlighting the most interesting information in the system and a plan for testing the methodology is described.

  4. Plasma engineering studies for Tennessee Tokamak (TENTOK) fusion power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, K.E.; Lacatski, J.T.; Miller, J.B.; Bryan, W.E.; King, P.W.; Santoro, R.T.; Uckan, N.A.; Shannon, T.E.

    1984-02-01

    This paper summarizes the results of the plasma engineering and systems analysis studies for the Tennessee Tokamak (TENTOK) fusion power reactor. TENTOK is a 3000-MW(t) central station power plant that uses deuterium-tritium fuel in a D-shaped tokamak plasma configuration with a double-null poloidal divertor. The major parameters are R/sub 0/ = 6.4 m, a = 1.6 m, sigma (elongation) = 1.65, (n) = 1.5 x 10/sup 20/ m/sup -3/, (T) = 15 keV, (..beta..) = 6%, B/sub T/ (on-axis) = 5.6 T, I/sub p/ = 8.5 MA, and wall loading = 3 MW/m/sup 2/. Detailed analyses are performed in the areas of (1) transport simulation using the one-and-one-half-dimensional (1-1/2-D) WHIST transport code, (2) equilibrium/poloidal field coil systems, (3) neutral beam and radiofrequency (rf) heating, and (4) pellet fueling. In addition, impurity control systems, diagnostics and controls, and possible microwave plasma preheating and steady-state current drive options are also considered. Some of the major features of TENTOK include rf heating in the ion cyclotron range of frequencies, superconducting equilibrium field coils outside the superconducting toroidal field coils, a double-null poloidal divertor for impurity control and alpha ash removal, and rf-assisted plasma preheating and current startup.

  5. Radiation damage of graphite in fission and fusion reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B. (GA Technologies, Inc., San Diego, CA (USA)); Kelly, B.T. (Springfields Nuclear Power Development Labs. (UK))

    1984-05-01

    Increasing the crystalline perfection of artificial graphites is suggested as one method of reducing the crystallite damage. The life expectance for the isotropic conventional graphites will in each case depend on the reactor component for which it will be used and on its design considerations. Based on neutron damage and related dimensional changes it is estimated graphite will be tenable to about 3x10/sup 22/ n/cm/sup 2/ (EDN) at 400/sup 0/C, 0.6x10/sup 22/ n/cm/sup 2/ (EDN) at 1000/sup 0/C and 1.4x10/sup 22/ n/cm/sup 2/ (EDN) at 1400/sup 0/C. There are no data above 1400/sup 0/C on which to speculate. A dose of 2x10/sup 22/ n/cm/sup 2/ may be accumulated in times ranging from as short as a few months in the first wall region of high power density designs to the fusion plant lifetime (30 years) in the neutron reflector region behind the blanket.

  6. Mechanical design aspects of a tandem mirror fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Neef, W.S. Jr.

    1977-04-25

    Two ''plugs'' of dense plasma at either end of a central solenoid cell form the basis of a new mirror fusion power plant concept. A central cell blanket design is presented. Modules on crawler tracks serviced by remote welding and handling machines of very simple design are important features resulting from linear axisymmetric geometry. Three blanket designs are considered and the best one presented in some detail. It has lithium as the breeder material, helium cooled. ''Plug'' magnet field strengths must be high. A novel magnet is presented to satisfy the physics of the end plugs. Beam sources at 1,200 KV present special problems. Methods of voltage standoff, arc damage control, and neutralization are discussed. New secondary containment ideas are presented to allow removable roof sections of balanced design.

  7. An integrated approach to assessing the fracture safe margins of fusion reactor structures

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R. [Univ. of California, Santa Barbara, CA (United States)

    1996-10-01

    Design and operation of fusion reactor structures will require an appropriate data base closely coupled to a reliable failure analysis method to safely manage irradiation embrittlement. However, ongoing irradiation programs will not provide the information on embrittlement necessary to accomplish these objectives. A new engineering approach is proposed based on the concept of a master toughness-temperature curve indexed on an absolute temperature scale using shifts to account for variables such as size scales, crack geometry and loading rates as well as embrittlement. While providing a simple practical engineering expedient, the proposed method can also be greatly enhanced by fundamental mechanism based models of fracture and embrittlement. Indeed, such understanding is required for the effective use of small specimen test methods, which is a integral element in developing the necessary data base.

  8. Innovative fusion reactor design analysis: Annual performance report, May 15, 1988--January 31, 1989

    Energy Technology Data Exchange (ETDEWEB)

    Klein, A.C.

    1989-01-31

    This report discusses the following topics on fusion reactor component design: FLiBe intermediate heat exchanger design analysis; FLiBe properties; design methodology; FLiBe system steam generator freezeup; FLiBe reactor systems studies; tritium breeding ratio control; analysis of original objectives; and budget analysis. 15 refs., 13 figs., 3 tabs. (LSP)

  9. Core Plasma Characteristics of a Spherical Tokamak D-3He Fusion Reactor

    Institute of Scientific and Technical Information of China (English)

    Shi Bingren

    2005-01-01

    The magnetic fusion reactor using the advanced D-3He fuels has the advantage of much less-neutron productions so that the consequent damages to the first wall are less serious. If the establishment of this kind of reactor becomes realistic, the exploration of 3He on the moon will be largely motivated. Based on recent progresses in the spherical torus (ST) research, we have physically designed a D-3He fusion reactor using the extrapolated results from the ST experiments and also the present-day tokamak scaling. It is found that the reactor size significantly depends on the wall reflection coefficient of the synchrotron radiation and of the impurity contaminations.The secondary reaction between D-D that promptly leads to the D-T reaction producing 14 MeV neutrons is also estimated. Comparison of this D-3He ST reactor with the D-T reactor is made.

  10. Concept of a charged fusion product diagnostic for NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Boeglin, W. U.; Valenzuela Perez, R. [Department of Physics, Florida International University, 11200 SW 8th Street, Miami, Florida 33199 (United States); Darrow, D. S. [Princeton Plasma Physics Laboratory, James Forrestal Campus, P.O. Box 451, Princeton, New Jersey 08543 (United States)

    2010-10-15

    The concept of a new diagnostic for NSTX to determine the time dependent charged fusion product emission profile using an array of semiconductor detectors is presented. The expected time resolution of 1-2 ms should make it possible to study the effect of magnetohydrodynamics and other plasma activities (toroidal Alfven eigenmodes (TAE), neoclassical tearing modes (NTM), edge localized modes (ELM), etc.) on the radial transport of neutral beam ions. First simulation results of deuterium-deuterium (DD) fusion proton yields for different detector arrangements and methods for inverting the simulated data to obtain the emission profile are discussed.

  11. Advanced Burner Reactor 1000MWth Reference Concept

    Energy Technology Data Exchange (ETDEWEB)

    Cahalan, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fanning, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kellogg, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, L. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Momozaki, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Reed, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Salev, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Seidensticker, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Tang, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Tzanos, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Wei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Chikazawa, Y. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2007-09-30

    The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence, to validate the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat.

  12. Beyond ITER: neutral beams for a demonstration fusion reactor (DEMO) (invited).

    Science.gov (United States)

    McAdams, R

    2014-02-01

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

  13. Beyond ITER: Neutral beams for a demonstration fusion reactor (DEMO) (invited)

    Energy Technology Data Exchange (ETDEWEB)

    McAdams, R., E-mail: roy.mcadams@ccfe.ac.uk [EURATOM/CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

    2014-02-15

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

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

    Science.gov (United States)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion

  15. Fusion reactor materials semiannual progress report for the period ending March 31, 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    This is the fourteenth in a series of semiannual technical progress reports on fusion reactor materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the US Depart of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. Separate abstracts were prepared for each individual section.

  16. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

  17. Operation of Fusion Reactors in One Atmosphere of Air Instead of Vacuum Systems

    Science.gov (United States)

    Roth, J. Reece

    2009-07-01

    Engineering design studies of both magnetic and inertial fusion power plants have assumed that the plasma will undergo fusion reactions in a vacuum environment. Operation under vacuum requires an expensive additional major system for the reactor-a vacuum vessel with vacuum pumping, and raises the possibility of sudden unplanned outages if the vacuum containment is breached. It would be desirable in many respects if fusion reactors could be made to operate at one atmosphere with air surrounding the plasma, thus eliminating the requirement of a pressure vessel and vacuum pumping. This would have obvious economic, reliability, and engineering advantages for currently envisaged power plant reactors; it would make possible forms of reactor control not possible under vacuum conditions (i.e. adiabatic compression of the fusion plasma by increasing the pressure of surrounding gas); it would allow reactors used as aircraft engines to operate as turbojets or ramjets in the atmosphere, and it would allow reactors used as fusion rockets to take off from the surface of the earth instead of low earth orbit.

  18. Fusion reactor materials. Semiannual progress report for period ending September 30, 1993

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F.; Burn, G.L.; Knee`, S.S.; Dowker, C.L. [comps.

    1994-02-01

    This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  19. Fusion transmutation of waste: design and analysis of the in-zinerator concept.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, S. M.; Cipiti, Benjamin B.; Olson, Craig Lee; Guild-Bingham, Avery (Texas A& M University, College Station, TX); Venneri, Francesco (General Atomics, San Diego, CA); Meier, Wayne (LLNL, Livermore, CA); Alajo, A.B. (Texas A& M University, College Station, TX); Johnson, T. R. (Argonne Mational Laboratory, Argonne, IL); El-Guebaly, L. A. (University of Wisconsin, Madison, WI); Youssef, M. E. (University of California, Los Angeles, CA); Young, Michael F.; Drennen, Thomas E. (Hobart & William Smith College, Geneva, NY); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Morrow, Charles W.; Turgeon, Matthew C.; Wilson, Paul (University of Wisconsin, Madison, WI); Phruksarojanakun, Phiphat (University of Wisconsin, Madison, WI); Grady, Ryan (University of Wisconsin, Madison, WI); Keith, Rodney L.; Smith, James Dean; Cook, Jason T.; Sviatoslavsky, Igor N. (University of Wisconsin, Madison, WI); Willit, J. L. (Argonne Mational Laboratory, Argonne, IL); Cleary, Virginia D.; Kamery, William (Hobart & William Smith College, Geneva, NY); Mehlhorn, Thomas Alan; Rochau, Gary Eugene

    2006-11-01

    Due to increasing concerns over the buildup of long-lived transuranic isotopes in spent nuclear fuel waste, attention has been given in recent years to technologies that can burn up these species. The separation and transmutation of transuranics is part of a solution to decreasing the volume and heat load of nuclear waste significantly to increase the repository capacity. A fusion neutron source can be used for transmutation as an alternative to fast reactor systems. Sandia National Laboratories is investigating the use of a Z-Pinch fusion driver for this application. This report summarizes the initial design and engineering issues of this ''In-Zinerator'' concept. Relatively modest fusion requirements on the order of 20 MW can be used to drive a sub-critical, actinide-bearing, fluid blanket. The fluid fuel eliminates the need for expensive fuel fabrication and allows for continuous refueling and removal of fission products. This reactor has the capability of burning up 1,280 kg of actinides per year while at the same time producing 3,000 MWth. The report discusses the baseline design, engineering issues, modeling results, safety issues, and fuel cycle impact.

  20. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  1. Proceedings of the Office of Fusion Energy/DOE workshop on ceramic matrix composites for structural applications in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.H. (Pacific Northwest Lab., Richland, WA (USA)); Lucas, G.E. (California Univ., Santa Barbara, CA (USA))

    1990-11-01

    A workshop to assess the potential application of ceramic matrix composites (CMCs) for structural applications in fusion reactors was held on May 21--22, 1990, at University of California, Santa Barbara. Participants included individuals familiar with materials and design requirements in fusion reactors, ceramic composite processing and properties and radiation effects. The primary focus was to list the feasibility issues that might limit the application of these materials in fusion reactors. Clear advantages for the use of CMCs are high-temperature operation, which would allow a high-efficiency Rankine cycle, and low activation. Limitations to their use are material costs, fabrication complexity and costs, lack of familiarity with these materials in design, and the lack of data on radiation stability at relevant temperatures and fluences. Fusion-relevant feasibility issues identified at this workshop include: hermetic and vacuum properties related to effects of matrix porosity and matrix microcracking; chemical compatibility with coolant, tritium, and breeder and multiplier materials, radiation effects on compatibility; radiation stability and integrity; and ability to join CMCs in the shop and at the reactor site, radiation stability and integrity of joints. A summary of ongoing CMC radiation programs is also given. It was suggested that a true feasibility assessment of CMCs for fusion structural applications could not be completed without evaluation of a material tailored'' to fusion conditions or at least to radiation stability. It was suggested that a follow-up workshop be held to design a tailored composite after the results of CMC radiation studies are available and the critical feasibility issues are addressed.

  2. Effect of particle pinch on the fusion performance and profile features of an international thermonuclear experimental reactor-like fusion reactor

    Science.gov (United States)

    Wang, Shijia; Wang, Shaojie

    2015-04-01

    The evolution of the plasma temperature and density in an international thermonuclear experimental reactor (ITER)-like fusion device has been studied by numerically solving the energy transport equation coupled with the particle transport equation. The effect of particle pinch, which depends on the magnetic curvature and the safety factor, has been taken into account. The plasma is primarily heated by the alpha particles which are produced by the deuterium-tritium fusion reactions. A semi-empirical method, which adopts the ITERH-98P(y,2) scaling law, has been used to evaluate the transport coefficients. The fusion performances (the fusion energy gain factor, Q) similar to the ITER inductive scenario and non-inductive scenario (with reversed magnetic shear) are obtained. It is shown that the particle pinch has significant effects on the fusion performance and profiles of a fusion reactor. When the volume-averaged density is fixed, particle pinch can lower the pedestal density by ˜30 % , with the Q value and the central pressure almost unchanged. When the particle source or the pedestal density is fixed, the particle pinch can significantly enhance the Q value by 60 % , with the central pressure also significantly raised.

  3. A novel concept for CRIEC-driven subcritical research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nieto, M.; Miley, G.H. [Illinois Univ., Fusion Studies Lab., Dept. of Nuclear, Plasma, and Radiological Engineering, Urbana, IL (United States)

    2001-07-01

    A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

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

    Science.gov (United States)

    Kumar, B. Ramesh; Gangradey, R.

    2012-11-01

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

  5. 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.

  6. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  7. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  8. An Inertial-Fusion Z-Pinch Power Plant Concept

    Energy Technology Data Exchange (ETDEWEB)

    DERZON,MARK S.; ROCHAU,GARY E.; DEGROOT,J.; OLSON,CRAIG L.; PETERSON,P.; PETERSON,R.R.; SLUTZ,STEPHEN A.; ZAMORA,ANTONIO J.

    2000-12-15

    With the promising new results of fast z-pinch technology developed at Sandia National Laboratories, we are investigating using z-pinch driven high-yield Inertial Confinement Fusion (ICF) as a fusion power plant energy source. These investigations have led to a novel fusion system concept based on an attempt to separate many of the difficult fusion engineering issues and a strict reliance on existing technology, or a reasonable extrapolation of existing technology, wherever possible. In this paper, we describe the main components of such a system with a focus on the fusion chamber dynamics. The concept works with all of the electrically-coupled ICF proposed fusion designs. It is proposed that a z-pinch driven ICF power system can be feasibly operated at high yields (1 to 30 GJ) with a relatively low pulse rate (0.01-0.1 Hz). To deliver the required current from the rep-rated pulse power driver to the z-pinch diode, a Recyclable Transmission Line (RTL) and the integrated target hardware are fabricated, vacuum pumped, and aligned prior to loading for each power pulse. In this z-pinch driven system, no laser or ion beams propagate in the chamber such that the portion of the chamber outside the RTL does not need to be under vacuum. Additionally, by utilizing a graded-density solid lithium or fluorine/lithium/beryllium eutectic (FLiBe) blanket between the source and the first-wall the system can breed its own fuel absorb a large majority of the fusion energy released from each capsule and shield the first-wall from a damaging neutron flux. This neutron shielding significantly reduces the neutron energy fluence at the first-wall such that radiation damage should be minimal and will not limit the first-wall lifetime. Assuming a 4 m radius, 8 m tall cylindrical chamber design with an 80 cm thick spherical FLiBe blanket, our calculations suggest that a 20 cm thick 6061-T6 Al chamber wall will reach the equivalent uranium ore radioactivity level within 100 years after a 30

  9. An Inertial-Fusion Z-Pinch Power Plant Concept

    Energy Technology Data Exchange (ETDEWEB)

    DERZON,MARK S.; ROCHAU,GARY E.; DEGROOT,J.; OLSON,CRAIG L.; PETERSON,P.; PETERSON,R.R.; SLUTZ,STEPHEN A.; ZAMORA,ANTONIO J.

    2000-12-15

    With the promising new results of fast z-pinch technology developed at Sandia National Laboratories, we are investigating using z-pinch driven high-yield Inertial Confinement Fusion (ICF) as a fusion power plant energy source. These investigations have led to a novel fusion system concept based on an attempt to separate many of the difficult fusion engineering issues and a strict reliance on existing technology, or a reasonable extrapolation of existing technology, wherever possible. In this paper, we describe the main components of such a system with a focus on the fusion chamber dynamics. The concept works with all of the electrically-coupled ICF proposed fusion designs. It is proposed that a z-pinch driven ICF power system can be feasibly operated at high yields (1 to 30 GJ) with a relatively low pulse rate (0.01-0.1 Hz). To deliver the required current from the rep-rated pulse power driver to the z-pinch diode, a Recyclable Transmission Line (RTL) and the integrated target hardware are fabricated, vacuum pumped, and aligned prior to loading for each power pulse. In this z-pinch driven system, no laser or ion beams propagate in the chamber such that the portion of the chamber outside the RTL does not need to be under vacuum. Additionally, by utilizing a graded-density solid lithium or fluorine/lithium/beryllium eutectic (FLiBe) blanket between the source and the first-wall the system can breed its own fuel absorb a large majority of the fusion energy released from each capsule and shield the first-wall from a damaging neutron flux. This neutron shielding significantly reduces the neutron energy fluence at the first-wall such that radiation damage should be minimal and will not limit the first-wall lifetime. Assuming a 4 m radius, 8 m tall cylindrical chamber design with an 80 cm thick spherical FLiBe blanket, our calculations suggest that a 20 cm thick 6061-T6 Al chamber wall will reach the equivalent uranium ore radioactivity level within 100 years after a 30

  10. Z-Pinch Magneto-Inertial Fusion Propulsion Engine Design Concept

    Science.gov (United States)

    Miernik, Janie H.; Statham, Geoffrey; Adams, Robert B.; Polsgrove, Tara; Fincher, Sharon; Fabisinski, Leo; Maples, C. Dauphne; Percy, Thomas K.; Cortez, Ross J.; Cassibry, Jason

    2011-01-01

    Fusion-based nuclear propulsion has the potential to enable fast interplanetary transportation. Due to the great distances between the planets of our solar system and the harmful radiation environment of interplanetary space, high specific impulse (Isp) propulsion in vehicles with high payload mass fractions must be developed to provide practical and safe vehicles for human spaceflight missions. Magneto-Inertial Fusion (MIF) is an approach which has been shown to potentially lead to a low cost, small fusion reactor/engine assembly (1). The Z-Pinch dense plasma focus method is an MIF concept in which a column of gas is compressed to thermonuclear conditions by an estimated axial current of approximately 100 MA. Recent advancements in experiments and the theoretical understanding of this concept suggest favorable scaling of fusion power output yield as I(sup 4) (2). The magnetic field resulting from the large current compresses the plasma to fusion conditions, and this is repeated over short timescales (10(exp -6) sec). This plasma formation is widely used in the field of Nuclear Weapons Effects (NWE) testing in the defense industry, as well as in fusion energy research. There is a wealth of literature characterizing Z-Pinch physics and existing models (3-5). In order to be useful in engineering analysis, a simplified Z-Pinch fusion thermodynamic model was developed to determine the quantity of plasma, plasma temperature, rate of expansion, energy production, etc. to calculate the parameters that characterize a propulsion system. The amount of nuclear fuel per pulse, mixture ratio of the D-T and nozzle liner propellant, and assumptions about the efficiency of the engine, enabled the sizing of the propulsion system and resulted in an estimate of the thrust and Isp of a Z-Pinch fusion propulsion system for the concept vehicle. MIF requires a magnetic nozzle to contain and direct the nuclear pulses, as well as a robust structure and radiation shielding. The structure

  11. The TITAN reversed-field-pinch fusion reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report discusses research on the titan-1 fusion power core. The major topics covered are: titan-1 fusion-power-core engineering; titan-1 divertor engineering; titan-1 tritium systems; titan-1 safety design and radioactive-waste disposal; and titan-1 maintenance procedures.

  12. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    Science.gov (United States)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-12-01

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  13. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  14. Reference mirror hybrid fusion-fission reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Lee, J.D.; Neef, W.S.

    1977-06-08

    The status of the reference mirror hybrid reactor design being performed by LLL and General Atomic is summarized. The reactor parameters have been chosen to minimize the cost of producing fissile fuel for consumption in fission power reactors. The design draws on the experience developed at LLL in previous hybrid reactor conceptual designs and on GA expertise in gas-cooling technology and fission reactor mechanical design. As in the past, we have emphasized the use of existing technology where possible and a minimum extrapolation of technology otherwise. We consider our projections for the plasma physics parameters to be conservative, in that they are well-founded on the experiments in 2XIIB and the interpretation of these experiments.

  15. Models and analyses for inertial-confinement fusion-reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Bohachevsky, I.O.

    1981-05-01

    This report describes models and analyses devised at Los Alamos National Laboratory to determine the technical characteristics of different inertial confinement fusion (ICF) reactor elements required for component integration into a functional unit. We emphasize the generic properties of the different elements rather than specific designs. The topics discussed are general ICF reactor design considerations; reactor cavity phenomena, including the restoration of interpulse ambient conditions; first-wall temperature increases and material losses; reactor neutronics and hydrodynamic blanket response to neutron energy deposition; and analyses of loads and stresses in the reactor vessel walls, including remarks about the generation and propagation of very short wavelength stress waves. A discussion of analytic approaches useful in integrations and optimizations of ICF reactor systems concludes the report.

  16. Analysis and evaluation of the hydrogen risk in a thermonuclear fusion reactor; Analyse et evaluation du risque hydrogene dans un reacteur de fusion thermonucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Chaudron, V. [Societe Helion, 13 - Aix en Provence (France); Arnould, F. [Technicatome DI SEPS, 13 - Aix en Provence (France); Latge, C. [CEA Cadarache, Dept. d' Etudes des Dechets DED, 13 - Saint Paul lez Durance (France); Laurent, A. [Ecole Nationale Superieure des Industries Chimiques, ENSIC, Lab. des Sciences du Genie Chimique, CNRS INPL, 54 - Villers les Nancy (France)

    2001-07-01

    After a recall of the principle of controlled thermonuclear fusion, the ITER reactor project is briefly described. The integrity of the reactor must be preserved in the case of a potential explosion of the hydrogen generated inside the reactor, in order to avoid any dispersion radioactive, chemical or toxic materials in the environment. The fundamental principles of safety developed to fulfill these objectives, in particular the defense-in-depth concept, are presented. The main potential source of hydrogen production is the oxidation of beryllium, which is used as protection material in the first wall of the torus, and the accidental presence of water, as reported in several scenarios. The confinement strategy is then described with the qualification of the role of the different barriers. Finally, the hydrogen explosion risk is analyzed and evaluated with respect to the sources, to the reference envelope scenarios and to the location of hydrogen inside the ITER reactor. It appears, at the engineering stage, that the vacuum toric vessel, the discharge reservoir and the exchanger compartments are the most worrying parts. (J.S.)

  17. Tabular equation of state of lithium for laser-fusion reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Young, D.A.; Ross, M.; Rogers, F.J.

    1979-01-19

    A tabular lithium equation of state was formulated from three separate equation-of-state models to carry out hydrodynamic simulations of a lithium-waterfall laser-fusion reactor. The models we used are: ACTEX for the ionized fluid, soft-sphere for the liquid and vapor, and pseudopotential for the hot, dense liquid. The models are smoothly joined over the range of density and temperature conditions appropriate for a laser-fusion reactor. We also fitted the models into two forms suitable for hydrodynamic calculations.

  18. Fusion reactor materials: Semiannual progress report for the period ending March 31, 1987

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1987-09-01

    This is the second in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities in the following areas: (1) Alloy Development for Irradiation Performance; (2) Damage Analysis and Fundamental Studies; and (3) Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. Separate analytics were prepared for the reports in this volume.

  19. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    Science.gov (United States)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

  20. Evaluation of the breed/burn fast reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Atefi, B.; Driscoll, M.J.; Lanning, D.D.

    1979-12-01

    A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH/sub 16/) as the moderator (because of the compact assembly and core designs it permitted).

  1. Parametric design study of tandem mirror fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, G.A.

    1977-05-27

    The parametric design study of the tandem mirror reactor (TMR) is described. The results of this study illustrate the variation of reactor characteristics with changes in the independent design parameters, reveal the set of design parameters which minimizes the cost of the reactor, and show the sensitivity of the optimized design to physics and technological uncertainties. The total direct capital cost of an optimized 1000 MWe TMR is estimated to be $1300/kWe. The direct capital cost of a 2000 MWe plant is less than $1000/kWe.

  2. On use of beryllium in fusion reactors: Resources, impurities and necessity of detritiation after irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Kolbasov, B.N., E-mail: b.kolbasov@yandex.ru; Khripunov, V.I., E-mail: Khripunov_VI@nrcki.ru; Biryukov, A.Yu.

    2016-11-01

    Highlights: • Potential needs in Be for fusion power engineering may exceed Be resources. • Be recycling after its operation in a fusion power plant (FPP) seems inevitable. • U impurity in Be seriously impairs environmental properties of fusion power plants. • Upon burial of irradiated Be the main problems are caused by U and {sup 3}H impurities. • Clearance of Be extracted from a FPP is impossible due to U impurity. - Abstract: Worldwide identified resources of beryllium somewhat exceed 80 000 t. Beryllium production in all the countries of the world in 2012 was about 230 t. At the same time, some conceptual designs of fusion power reactors envisage utilization of several hundred tons of this metal. Therefore return of beryllium into the production cycle (recycling) will be necessary. The beryllium ore from some main deposits has uranium content inadmissible for fusion reactors. This fact raises a question on the need to develop and apply an economically acceptable technology for beryllium purification from the uranium. Practically any technological procedure with beryllium used in fusion reactors requires its detritiation. A study of tritium and helium release from irradiated beryllium at different temperatures and rates of temperature increase was performed at Kurchatov Institute.

  3. The TITAN reversed-field-pinch fusion reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report discusses the following topics: overview of titan-2 design; titan-2 fusion-power-core engineering; titan-2 divertor engineering; titan-2 tritium systems; titan-2 safety design and radioactive-waste disposal; and titan-2 maintenance procedures.

  4. Recent contributions to fusion reactor design and technology development

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    The report contains a collection of 16 recent fusion technology papers on the STARFIRE Project, the study of alternate fusion fuel cycles, a maintainability study, magnet safety, neutral beam power supplies and pulsed superconducting magnets and energy transfer. This collection of papers contains contributions for Argonne National Laboratory, McDonnell Douglas Astronautics Company, General Atomic Company, The Ralph M. Parsons Company, the University of Illinois, and the University of Wisconsin. Separate abstracts are presented for each paper. (MOW)

  5. Concept of Staged Approach for International Fusion Materials Irradiation Facility

    CERN Document Server

    Sugimoto, M; Takeuchi, H

    2000-01-01

    The intense neutron source for development of fusion materials planned by international collaboration makes a new step to clarify the technical issues for realizing the 40 MeV, 250 mA deuteron beam facility. The baseline concept employs two identical 125 mA linac modules whose beams are combined at the flowing lithium target. Recent work for reducing the cost loading concerns the staged deployment of the full irradiation capability in three steps. The Japanese activity about the design and development study about IFMIF accelerator in this year is presented and the schedule of next several years is overviewed.

  6. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma (Computational Engineering Analysis, Albuquerque, NM); Al Rashdan, Ahmad (Texas A& M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  7. New reactor concepts. An analysis of the actual research status; Neue Reaktorkonzepte. Eine Analyse des aktuellen Forschungsstands

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias

    2017-04-15

    The report on new reactor concepts covers the following issues: characterization and survey of new reactor concepts; evaluation criteria: safety, resources for fuel supply, waste problems, economy and proliferation; comprehensive relevant aspects: thorium as alternative resource, partitioning and transmutation; actual developments and preliminary experiences for fast breeding reactor (FBR), high-temperature reactor (HTR), molten salt reactor (MSR), small modular reactor (SMR).

  8. A revaluation of helium/dpa ratios for fast reactor and thermal reactor data in fission-fusion correlations

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); Oliver, B.M.

    1996-10-01

    For many years it has been accepted that significant differences exist in the helium/dpa ratios produced in fast reactors and various proposed fusion energy devices. In general, the differences arise from the much larger rate of (n,{alpha}) threshold reactions occurring in fusion devices, reactions which occur for energies {ge} 6 MeV. It now appears, however, that for nickel-containing alloys in fast reactors the difference may not have been as large as was originally anticipated. In stainless steels that have a very long incubation period for swelling, for instance, the average helium concentration over the duration of the transient regime have been demonstrated in an earlier paper to be much larger in the FFTF out-of-core regions than first calculated. The helium/dpa ratios in some experiments conducted near the core edge or just outside of the FFTF core actually increase strongly throughout the irradiation, as {sup 59}Ni slowly forms by transmutation of {sup 58}Ni. This highly exothermic {sup 59}Ni(n,{alpha}) reaction occurs in all fast reactors, but is stronger in the softer spectra of oxide-fueled cores such as FFTF and weaker in the harder spectra of metal-fueled cores such as EBR-II. The formation of {sup 59}Ni also increases strongly in out-of-core unfueled regions where the reactor spectra softens with distance from the core.

  9. Risks of nuclear energy technology safety concepts of light water reactors

    CERN Document Server

    Kessler, Günter; Schlüter, Franz-Hermann

    2014-01-01

    The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on?reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: ? A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Ch

  10. Plasma Heating and Current Drive for Fusion Reactors

    Science.gov (United States)

    Holtkamp, Norbert

    2010-02-01

    ITER (in Latin ``the way'') is designed to demonstrate the scientific and technological feasibility of fusion energy. Fusion is the process by which two light atomic nuclei combine to form a heavier one and thus release energy. In the fusion process two isotopes of hydrogen - deuterium and tritium - fuse together to form a helium atom and a neutron. Thus fusion could provide large scale energy production without greenhouse effects; essentially limitless fuel would be available all over the world. The principal goals of ITER are to generate 500 megawatts of fusion power for periods of 300 to 500 seconds with a fusion power multiplication factor, Q, of at least 10. Q >= 10 (input power 50 MW / output power 500 MW). In a Tokamak the definition of the functionalities and requirements for the Plasma Heating and Current Drive are relevant in the determination of the overall plant efficiency, the operation cost of the plant and the plant availability. This paper summarise these functionalities and requirements in perspective of the systems under construction in ITER. It discusses the further steps necessary to meet those requirements. Approximately one half of the total heating will be provided by two Neutral Beam injection systems at with energy of 1 MeV and a beam power of 16 MW into the plasma. For ITER specific test facility is being build in order to develop and test the Neutral Beam injectors. Remote handling maintenance scheme for the NB systems, critical during the nuclear phase of the project, will be developed. In addition the paper will give an overview over the general status of ITER. )

  11. Fusion Reactor Materials semiannual progress report for the period ending March 31, 1992

    Energy Technology Data Exchange (ETDEWEB)

    1992-07-01

    This is the twelfth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; and Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  12. Fusion reactor materials semiannual progress report for the period ending March 31, 1991

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1991-07-01

    This is the tenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: alloy development for irradiation performance; damage analysis and fundamental studies; special purpose materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of program participants, and to provide a means of communicating the efforts of materials scientists to the test of the fusion community, both nationally and worldwide.

  13. Fusion reactor materials semiannual progress report for the period ending September 30, 1988

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-04-01

    This paper discusses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  14. Fusion reactor materials: Semiannual progress report for period ending September 30, 1987

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1988-03-01

    This is the third in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following technical progress reports: Alloy Development for Irradiation Performances; Damage Analysis and Fundamental Studies; Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  15. Fusion reactor materials semiannual progress report for the period ending September 30, 1989

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This is the seventh in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following technical progress reports: alloy development for irradiation performance, damage analysis and fundamental studies, and special purpose materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  16. 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)

  17. The Feasibility of Pellet Re-Fuelling of a Fusion Reactor

    DEFF Research Database (Denmark)

    Chang, Tinghong; Jørgensen, L. W.; Nielsen, P.

    1980-01-01

    The feasibility of re-fuelling a fusion reactor by injecting pellets of frozen hydrogen isotopes is reviewed. First a general look is taken of the dominant energy fluxes received by the pellet, the re-fuelling rate required and the relation between pellet size, injection speed and frequency...

  18. Fusion reactor materials semiannual progress report for period ending September 30, 1990

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This is the ninth in series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following technical progress reports: Alloy Development of Irradiation Performance; Damage Analysis and Fundamental Studies; and Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  19. Simulation of the target creation through FRC merging for a magneto-inertial fusion concept

    Science.gov (United States)

    Li, Chenguang; Yang, Xianjun

    2017-04-01

    A two-dimensional magnetohydrodynamics model has been used to simulate the target creation process in a magneto-inertial fusion concept named Magnetized Plasma Fusion Reactor (MPFR) [C. Li and X. Yang, Phys. Plasmas 23, 102702 (2016)], where the target plasma created through Field reversed configuration (FRC) merging was compressed by an imploding liner driven by the pulsed-power driver. In the scheme, two initial FRCs (Field reversed configurations) are translated into the region where FRC merging occurs, bringing out the target plasma ready for compression. The simulations cover the three stages of the target creation process: formation, translation, and merging. The factors affecting the achieved target are analyzed numerically. The magnetic field gradient produced by the conical coils is found to determine how fast the FRC is accelerated to peak velocity and the collision merging occurs. Moreover, it is demonstrated that FRC merging can be realized by real coils with gaps showing nearly identical performance, and the optimized target by FRC merging shows larger internal energy and retained flux, which is more suitable for the MPFR concept.

  20. Pellet bed reactor concept for nuclear electric propulsion

    Science.gov (United States)

    El-Genk, Mohamed S.; Morley, Nicholas J.; Juhasz, Albert

    1993-01-01

    For Nuclear Electric Propulsion (NEP) applications, gas cooled nuclear reactors with dynamic energy conversion systems offer high specific power and low total mass. This paper describes the Pellet Bed Reactor (PeBR) concept for potential NEP missions to Mars. The helium cooled, 75-80 MWt PeBR, consists of a single annular fuel region filled with a randomly packed bed of spherical fuel pellets, is designed for multiple starts, and offers unique safety and operation features. Each fuel pellet, about 8-10 mm in diameter, is composed of hundreds of TRISO type fuel microspheres embedded in a graphite matrix for a full retention of fission products. To eliminate the likelihood of a single-point failure, the annular core of the PeBR is divided into three 120° sectors. Each sector is self contained and separate and capable of operating and being cooled on its own and in cooperation with either one or two other sectors. Each sector is coupled to a separate, 5 MWe Closed Brayton Cycle (CBC) energy conversion unit and is subcritical for safe handling and launching. In the event of a failure of the cooling system of a core sector, the reactor power level may be reduced, allowing adjacent sectors to convect the heat away using their own cooling system, thus maintaining reactor operation. Also, due to the absence of an internal core structure in the PeBR core, fueling of the reactor can easily be performed either at the launch facility or in orbit, and refueling can be accomplished in orbit as needed to extend the power system lifetime

  1. Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

    Institute of Scientific and Technical Information of China (English)

    WANG Xin-Hua; GUO Hai-Ping; MOU Yun-Feng; ZHENG Pu; LIU Rong; YANG Xiao-Fei; YANG Jian

    2013-01-01

    A fusion-fission hybrid conceptual reactor is established.It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium.The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D+ beam of the Cockcroft-Walton neutron generator in direct current mode.The measured TPR distribution is compared with the calculated results obtained by the threedimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file.The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S(α,β) thermal scattering model,so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors.

  2. Recent development and application of a new safety analysis code for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: Brad.Merrill@inl.gov; Humrickhouse, Paul W.; Shimada, Masashi

    2016-11-01

    Highlights: • This paper presents recent code development activities for the MELCOR for fusion and Tritium Migration Analysis Program computer codes at the Idaho National Engineering Laboratory. • The capabilities of these computer codes are being merged into a single safety analysis tool for fusion reactor accidents. • The result of benchmarking these codes against previous code versions is presented by the authors of this paper. • This new capability is applied to study the tritium inventory and permeation rate for a water cold tungsten divertor that has neutron damage at 0.3 dpa. - Abstract: This paper describes the recent progress made in the development of two codes for fusion reactor safety assessments at the Idaho National Laboratory (INL): MELCOR for fusion and the Tritium Migration Analysis Program (TMAP). During the ITER engineering design activity (EDA), the INL Fusion Safety Program (FSP) modified the MELCOR 1.8.2 code for fusion applications to perform ITER thermal hydraulic safety analyses. Because MELCOR has undergone many improvements at SNL-NM since version 1.8.2 was released, the INL FSP recently imported these same fusion modifications into the MELCOR 1.8.6 code, along with the multiple fluids modifications of MELCOR 1.8.5 for fusion used in US advanced fusion reactor design studies. TMAP has also been under development for several decades at the INL by the FSP. TMAP treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by fluid flow within a given facility. Recently, TMAP was updated to consider multiple trap site types to allow the simulation of experimental data from neutron irradiated tungsten. The natural development path for both of these codes is to merge their capabilities into one computer code to provide a more comprehensive safety tool for analyzing accidents in fusion reactors. In this paper we detail recent developments in this

  3. Ion source development for a photoneutralization based NBI system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simonin, A.; Esch, H. P. L. de; Garibaldi, P.; Grand, C.; Bechu, S.; Bès, A.; Lacoste, A. [CEA-Cadarache, IRFM, F-13108 St. Paul-lez-Durance (France); LPSC, Grenoble-Alpes University, F-38026 Grenoble France (France)

    2015-04-08

    The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D° beam at 1MeV, including high wall-plug efficiency (η > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D{sup −} beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R and D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.

  4. Conceptual design of the Fast-Liner Reactor (FLR) for fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Moses, R.W.; Krakowski, R.A.; Miller, R.L.

    1979-02-01

    The generation of fusion power from the Fast-Liner Reactor (FLR) concept envisages the implosion of a thin (3-mm) metallic cylinder (0.2-m radius by 0.2-m length) onto a preinjected plasma. This plasma would be heated to thermonuclear temperatures by adiabatic compression, pressure confinement would be provided by the liner inertia, and thermal insulation of the wall-confined plasma would be established by an embedded azimuthal magnetic field. A 2- to 3-mu s burn would follow the approx. 10/sup 4/ m/s radial implosion and would result in a thermonuclear yield equal to 10 to 15 times the energy initially invested into the liner kinetic energy. For implosions occurring once every 10 s a gross thermal power of 430 MWt would be generated. The results of a comprehensive systems study of both physics and technology (economics) optima are presented. Despite unresolved problems associated with both the physics and technology of the FLR, a conceptual power plant design is presented.

  5. Ion source development for a photoneutralization based NBI system for fusion reactors

    Science.gov (United States)

    Simonin, A.; de Esch, H. P. L.; Garibaldi, P.; Grand, C.; Bechu, S.; Bès, A.; Lacoste, A.

    2015-04-01

    The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D° beam at 1MeV, including high wall-plug efficiency (η > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D- beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R&D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.

  6. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  7. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  8. Fast reactor core concepts to improve transmutation efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Fujimura, Koji; Kawashima, Katsuyuki [Hitachi Research Laboratory, Hitachi, Ltd., 7-1-1, Omika-cho, Hitachi-shi, Ibaraki, 319-1221 Japan (Japan); Itooka, Satoshi [Hitachi-GE Nuclear Energy, Ltd., 3-1-1, Saiwai-cho, Hitachi-shi, Ibaraki, 317-0073 Japan (Japan)

    2015-12-31

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  9. Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

    2015-10-15

    Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

  10. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    Science.gov (United States)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  11. Tritium issues to be solved for establishment of a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tanabe, Tetsuo, E-mail: tanabe@nucl.kyushu-u.ac.jp [Interdisciplinary Graduate School of Engineering Science, Kyushu University (Japan)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Overview of Japanese tritium researches. Black-Right-Pointing-Pointer Introduction of necessary works for establishment of tritium safety. Black-Right-Pointing-Pointer Tritium researches for burning plasma. - Abstract: In order to establish a D-T fusion reactor as an energy source, economical conversion of fusion energy to electricity and/or heat, attaining enough margins in tritium breeding, and insuring tritium safety must be simultaneously achieved. Scientists and researchers working on Tritium in Japan are now tackling with T related problems. Their research subjects can be categorized into two, i.e. researches on 'Science and technology' to establish safe and economic Tritium fuel cycle for fusion reactors and 'Tritium safety'. Many researchers from various universities, and institutes such as NIFS, JAEA and IEA (Inst. Environmental Science) in Japan are involved in various research programs. In this report, after brief introduction on Tritium related researches in Japan, important T issues to be solved for establishment of a fusion reactor will be summarized considering the handling of large amount of tritium, i.e. fuelling, D-T burning, T inventory, exhausting, refinement, confinement, permeation, leakage, contamination, regulation and tritium accountancy.

  12. Negative ion source development for a photoneutralization based neutral beam system for future fusion reactors

    Science.gov (United States)

    Simonin, A.; Agnello, R.; Bechu, S.; Bernard, J. M.; Blondel, C.; Boeuf, J. P.; Bresteau, D.; Cartry, G.; Chaibi, W.; Drag, C.; Duval, B. P.; de Esch, H. P. L.; Fubiani, G.; Furno, I.; Grand, C.; Guittienne, Ph; Howling, A.; Jacquier, R.; Marini, C.; Morgal, I.

    2016-12-01

    In parallel to the developments dedicated to the ITER neutral beam (NB) system, CEA-IRFM with laboratories in France and Switzerland are studying the feasibility of a new generation of NB system able to provide heating and current drive for the future DEMOnstration fusion reactor. For the steady-state scenario, the NB system will have to provide a high NB power level with a high wall-plug efficiency (η ˜ 60%). Neutralization of the energetic negative ions by photodetachment (so called photoneutralization), if feasible, appears to be the ideal solution to meet these performances, in the sense that it could offer a high beam neutralization rate (>80%) and a wall-plug efficiency higher than 60%. The main challenge of this new injector concept is the achievement of a very high power photon flux which could be provided by 3 MW Fabry-Perot optical cavities implanted along the 1 MeV D- beam in the neutralizer stage. The beamline topology is tall and narrow to provide laminar ion beam sheets, which will be entirely illuminated by the intra-cavity photon beams propagating along the vertical axis. The paper describes the present R&D (experiments and modelling) addressing the development of a new ion source concept (Cybele source) which is based on a magnetized plasma column. Parametric studies of the source are performed using Langmuir probes in order to characterize and compare the plasma parameters in the source column with different plasma generators, such as filamented cathodes, radio-frequency driver and a helicon antenna specifically developed at SPC-EPFL satisfying the requirements for the Cybele (axial magnetic field of 10 mT, source operating pressure: 0.3 Pa in hydrogen or deuterium). The paper compares the performances of the three plasma generators. It is shown that the helicon plasma generator is a very promising candidate to provide an intense and uniform negative ion beam sheet.

  13. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain); Sedano, L. A. [Asociacion Euratom-Ciematpara Fusion, Av. Complutense 22, 28040 Madrid (Spain); Perlado, J. M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain)

    2008-07-15

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T{sub 2} permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H{sub 2} tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations

  14. Fusion

    Science.gov (United States)

    Herman, Robin

    1990-10-01

    The book abounds with fascinating anecdotes about fusion's rocky path: the spurious claim by Argentine dictator Juan Peron in 1951 that his country had built a working fusion reactor, the rush by the United States to drop secrecy and publicize its fusion work as a propaganda offensive after the Russian success with Sputnik; the fortune Penthouse magazine publisher Bob Guccione sank into an unconventional fusion device, the skepticism that met an assertion by two University of Utah chemists in 1989 that they had created "cold fusion" in a bottle. Aimed at a general audience, the book describes the scientific basis of controlled fusion--the fusing of atomic nuclei, under conditions hotter than the sun, to release energy. Using personal recollections of scientists involved, it traces the history of this little-known international race that began during the Cold War in secret laboratories in the United States, Great Britain and the Soviet Union, and evolved into an astonishingly open collaboration between East and West.

  15. Effects of a liquid lithium curtain as the first wall in a fusion reactor plasma

    Institute of Scientific and Technical Information of China (English)

    Li Cheng-Yue; J.P. Allain; Deng Bai-Quan

    2007-01-01

    This paper explores the effect of a liquid lithium curtain on fusion reactor plasma, such curtain is utilized as the first wall for the engineering outline design of the Fusion Experimental Breeder (FEB-E). The relationships between the surface temperature of a liquid lithium curtain and the effective plasma charge, fuel dilution and fusion power production have been derived. Results indicate that under normal operation, the evaporation of liquid lithium does not seriously affect the effective plasma charge, but effects on fuel dilution and fusion power are more sensitive. As an example, it has investigated the relationships between the liquid lithium curtain flow velocity and the rise of surface temperature based on operation scenario Ⅱ of the FEB-E design with reversed shear configuration and high power density. Results show that even if the liquid lithium curtain flow velocity is as low as 0.5 m/s, the effects of evaporation from the liquid lithium curtain on plasma are negligible. In the present design, the sputtering of liquid lithium curtain and the particle removal effects of the divertor are not yet considered in detail. Further studies are in progress, and in this work implication of lithium erosion and divertor physics on fusion reactor operation are discussed.

  16. PX–An Innovative Safety Concept for an Unmanned Reactor

    Directory of Open Access Journals (Sweden)

    Sung-Jae Yi

    2016-02-01

    Full Text Available An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

  17. Overview of the Lockheed Martin Compact Fusion Reactor (CFR) T4B Experiment

    Science.gov (United States)

    McGuire, Thomas

    2016-10-01

    The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. The CFR uses a diamagnetic, high beta, magnetically encapsulated, linear ring cusp plasma confinement scheme. The goal of the T4B experiment is to demonstrate a suitable plasma target for heating experiments and to characterize the behavior of plasma sources in the CFR configuration. The design of the T4B experiment will be presented, including discussion of predicted behavior, plasma sources, heating mechanisms, diagnostics suite and relevant numerical modeling. ©2016 Lockheed Martin Corporation. All Rights Reserved.

  18. Fusion reactor theory and conceptual design. (Latest citations from the INSPEC: Information Services for the Physics and Engineering Communities database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The bibliography contains citations concerning theoretical and conceptual aspects of fusion reactor physics and designs. A variety of fusion reactors is discussed, including Tokamak, experimental, commercial, tandem mirror, and superconducting magnetic. Topics also include fusion reactor materials, Tokamak devices, blanket design, divertors, fusion plasma production, superconducting magnets, and cryogenic systems. (Contains a minimum of 159 citations and includes a subject term index and title list.)

  19. Disassembly of the fusion-1 capsule after irradiation in the BOR-60 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H. [Argonne National Lab., IL (United States); Kazakov, V.A.; Chakin, V.P. [and others

    1997-04-01

    A U.S./Russia (RF) collaborative irradiation experiment, Fusion-1, was completed in June 1996 after reaching a peak exposure of {approx}17 dpa in the BOR-60 fast reactor at the Research Institute of Atomic Reactors (RIAR) in Russia. The specimens were vanadium alloys, mainly of recent heats from both countries. In this reporting period, the capsule was disassembled at the RIAR hot cells and all test specimens were successfully retrieved. For the disassembly, an innovative method of using a heated diffusion oil to melt and separate the lithium bond from the test specimens was adopted. This method proved highly successful.

  20. Membrane pumping technology for helium and hydrogen isotope separation in the fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pistunovich, V.I. [Kurchatov Inst., Moscow (Russian Federation). NFI RRC; Pigarov, A.Yu. [Kurchatov Inst., Moscow (Russian Federation). NFI RRC; Busnyuk, A.O. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Livshits, A.I. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Notkin, M.E. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Samartsev, A.A. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Borisenko, K.L. [Efremov Institute, St. Petersburg (Russian Federation); Darmogray, V.V. [Efremov Institute, St. Petersburg (Russian Federation); Ershov, B.D. [Efremov Institute, St. Petersburg (Russian Federation); Filippova, L.V. [Efremov Institute, St. Petersburg (Russian Federation); Mudugin, B.G. [Efremov Institute, St. Petersburg (Russian Federation); Odintsov, V.N. [Efremov Institute, St. Petersburg (Russian Federation); Saksagansky, G.L. [Efremov Institute, St. Petersburg (Russian Federation); Serebrennikov, D.V. [Efremov Institute, St. Petersburg (Russian Federation)

    1995-03-01

    A gas pumping system for ITER, improved by implementation of superpermeable membranes for selective hydrogen isotope exhaust, is considered. A study of the pumping capability of a niobium membrane for a hydrogen-helium mixture has been performed.Monte Carlo simulations of gas behaviour for the experimental facility and fusion reactor have been done.The scheme of the ITER pumping system with the membranes and membrane pumping technology was considered. The conceptual study the membrane pump for the ITER was done. This work gives good prospects for the membrane pumping use in ITER to reduce the total inventory of tritium necessary for reactor operation. (orig.).

  1. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  2. REACTOR - a Concept for establishing a System-of-Systems

    Science.gov (United States)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

    2014-05-01

    well suited to establish brokers, which mediate metadata and semantic information about the resources of all involved systems. This concept has been developed within the project Collaborative, Complex, and Critical Decision-Support in Evolving Crises (TRIDEC) on the basis of semantic registries describing all facets of events and services utilisable for crisis management systems. The implementation utilises an operative infrastructure including an Enterprise Service Bus (ESB), adapters to proprietary sensor systems, a workflow engine, and a broker-based MOM. It also applies current technologies like actor-based frameworks for highly concurrent, distributed, and fault tolerant event-driven applications. Therefore REACTOR implementations are well suited to be hosted in a cloud that provides Infrastructure as a Service (IaaS). To provide low entry barriers for legacy and future systems, REACTOR adapts the principles of Design by Contract (DbC) as well as standardised and common information models like the Sensor Web Enablement (SWE) or the JavaScript Object Notation for geographic features (GeoJSON). REACTOR has been applied exemplarily within two different scenarios, Natural Crisis Management and Industrial Subsurface Development.

  3. Vibration of fusion reactor components with magnetic damping

    Energy Technology Data Exchange (ETDEWEB)

    D’Amico, Gabriele; Portone, Alfredo [Fusion for Energy – Torres Diagonal Litoral B3 – c/Josep Plá n.2, Barcelona (Spain); Rubinacci, Guglielmo [Department of Electrical Eng. and Information Technologies, Università di Napoli Federico II, Via Claudio, 21, 80125 Napoli (Italy); Testoni, Pietro, E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy – Torres Diagonal Litoral B3 – c/Josep Plá n.2, Barcelona (Spain)

    2016-11-01

    The aim of this paper is to assess the importance of the magnetic damping in the dynamic response of the main plasma facing components of fusion machines, under the strong Lorentz forces due to Vertical Displacement Events. The additional eddy currents due to the vibration of the conducting structures give rise to volume loads acting as damping forces, a kind of viscous damping, being these additional loads proportional to the vibration speed. This effect could play an important role when assessing, for instance, the inertial loads associated to VV movements in case of VDEs. In this paper, we present the results of a novel numerical formulation, in which the field equations are solved by adopting a very effective fully 3D integral formulation, not limited to the analysis of thin shell structures, as already successfully done in several approaches previously published.

  4. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. [eds.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  5. Code development incorporating environmental, safety and economic aspects of fusion reactors; Annual progress report

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T.K.; Greenspan, E.; Holdren, J.P.

    1993-12-31

    This document is a proposal to continue the authors work on the Environmental, Safety and Economic (ESE) aspects of fusion reactors under DOE contract DE-FR03-89ER52514. The grant objectives continue those from the previous grant: (1) completion of first-generation Environmental, Safety and Economic (ESE) computer modules suitable as integral components of tokamak systems codes. (2) continuation of work on special topics, in support of the above and in response to OFE requests. The proposal also highlights progress on the contract in the twelve months since April, 1992. This has included work with the ARIES and ITER design teams, work on tritium management, studies on materials activation, and calculation of radioactive inventories in fusion reactors.

  6. Status of R&D Activities on Materials for Fusion Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baluc, N.; Abe, K.; Boutard, J. L.; Chernov, V. M.; Diegele, E.; Jitsukawa, S.; Kimura, Akihiko; Klueh, R. L.; Kohyama, Akira; Kurtz, Richard J.; Lasser, R.; Matsui, H.; Moslang, A.; Muroga, T.; Odette, George R.; Tran, M. Q.; van der Schaaf, B.; Wu, Y.; Yu, J.; Zinkle, Steven J.

    2007-09-19

    Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, P.R. China, Russia and the USA. They relate to development of new high temperature, radiation resistant materials, development of coatings that shall act as erosion, corrosion, permeation or electrical/MHD barriers, characterization of the whole candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

  7. New Revelation of Lightning Ball Observation and Proposal for a Nuclear Reactor Fusion Experiment

    CERN Document Server

    Tar, Domokos

    2009-01-01

    In this paper, the author brings further details regarding his Lightning Ball observation that were not mentioned in the first one (Ref.1-2). Additionally, he goes more into detail as the three forces that are necessary to allow the residual crescent form the hydrodynamic vortex ring to shrink into a sphere.Further topics are the similarities and analogies between the Lightning Ball formation's theory and the presently undertaken Tokamak-Stellarator-Spheromak fusion reactor experiments. A new theory and its experimental realisation are proposed as to make the shrinking of the hot plasma of reactors into a ball possible by means of the so called long range electromagnetic forces. In this way,the fusion ignition temperature could possibly atteined.

  8. Results from systematic modeling of neutron damage in inertial fusion energy reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perlado, J.M. E-mail: mperlado@denim.upm.es; Dominguez, E.; Malerba, L.; Marian, J.; Lodi, D.; Salvador, M.; Alonso, E.; Caturla, Ma.J.; Diaz de la Rubia, T

    2002-01-01

    Radiation damage is an important issue in the lifetime of the structural materials in an Inertial Fusion Energy (IFE) Reactor. The effect will strongly depend on the class of chamber protection at the IFE Reactor design. This paper gives results from DENIM, and collaboration with LLNL, on the necessary magnitudes for the final evaluation of neutron damage. The determination of the neutron intensities and energy spectra emerging from the target, the energy spectra of the Primary Knock-on Atoms (PKA) resulting from the neutron interactions, the modeling at microscopic scale of the pulsed irradiation in metals are reported, in addition to reference to the work on the time dependence of neutron flux in IFE protected chamber. Results are also presented on the damage accumulation in SiC, relevant both for magnetic (MFE) and inertial fusion.

  9. Assessment of radiological releases to the environment from a fusion reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Shank, K.E.; Oakes, T.W.; Easterly, C.E.

    1978-05-01

    This report summarizes the expected tritium and activation-product inventories and presents an assessment of the potential radiological releases from a fusion reactor power plant, hypothetically located at the Oak Ridge National Laboratory. Routine tritium releases and the resulting dose assessment are discussed. Uncertainties associated with the conversion of tritium gas to tritium oxide and the global tritium cycling are evaluated. The difficulties of estimating releases of activated materials and the subsequent dose commitment are reviewed.

  10. Resistive sensor and electromagnetic actuator for feedback stabilization of liquid metal walls in fusion reactors

    CERN Document Server

    Mirhoseini, S H M

    2016-01-01

    Liquid metal walls in fusion reactors will be subject to instabilities, turbulence, induced currents, error fields and temperature gradients that will make them locally bulge, thus entering in contact with the plasma, or deplete, hence exposing the underlying solid substrate. To prevent this, research has begun to actively stabilize static or flowing liquid metal layers by locally applying forces in feedback with thickness measurements. Here we present resistive sensors of liquid metal thickness and demonstrate jxB actuators, to locally control it.

  11. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  12. High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu

    1996-10-01

    A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)

  13. Development of whole energy absorption spectrometer for decay heat measurement on fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    To measure decay heat on fusion reactor materials irradiated by D-T neutrons, a Whole Energy Absorption Spectrometer (WEAS) consisting of a pair of large BGO (bismuth-germanate) scintillators was developed. Feasibility of decay heat measurement with WEAS for various materials and for a wide range of half-lives (seconds - years) was demonstrated by experiments at FNS. Features of WEAS, such as high sensitivity, radioactivity identification, and reasonably low experimental uncertainty of {approx} 10 %, were found. (author)

  14. Direct extrapolation of radial profile data to a self-ignited fusion reactor based on the gyro-Bohm model

    Energy Technology Data Exchange (ETDEWEB)

    Miyazawa, J., E-mail: miyazawa@LHD.nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Goto, T.; Morisaki, T.; Goto, M.; Sakamoto, R.; Motojima, G.; Peterson, B.J.; Suzuki, C.; Ida, K.; Yamada, H.; Sagara, A. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer The DPE method predicts temperature and density profiles in a fusion reactor. Black-Right-Pointing-Pointer This method is based on the gyro-Bohm type parameter dependence. Black-Right-Pointing-Pointer The size of fusion reactor is determined to fulfill the power balance. Black-Right-Pointing-Pointer The reactor size is proportional to a factor and -4/3 power of the magnetic field. Black-Right-Pointing-Pointer This factor can be a measure of plasma performance like the fusion triple product. - Abstract: A new method named direct profile extrapolation (DPE) has been developed to estimate the radial profiles of temperature and density in a fusion reactor. This method directly extrapolates the radial profiles observed in present experiments to the fusion reactor condition assuming gyro-Bohm type parameter dependence. The magnetohydrodynamic equilibrium that fits the experimental profile data is used to determine the plasma volume. Four enhancement factors for the magnetic field strength, the density, the plasma beta, and the energy confinement are assumed. Then, the plasma size is determined so as to fulfill the power balance in the reactor plasma. The plasma performance can be measured by an index, C{sub exp}, introduced in the DPE method. The minimum magnetic stored energy of the fusion reactor to achieve self-ignition is shown to be proportional to the cube of C{sub exp} and inversely proportional to the square of magnetic field strength. Using this method, the design window of a self-ignited fusion reactor that can be extrapolated from recent experimental results in the Large Helical Device (LHD) is considered. Also discussed is how large an enhancement is needed for the LHD experiment to ensure the helical reactor design of FFHR2m2.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Surface modifications of fusion reactor relevant materials on exposure to fusion grade plasma in plasma focus device

    Energy Technology Data Exchange (ETDEWEB)

    Niranjan, Ram, E-mail: niranjan@barc.gov.in [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Rout, R.K.; Srivastava, R. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravarthy, Y. [Laser and Plasma Technology Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Mishra, P. [Materials Processing Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kaushik, T.C.; Gupta, Satish C. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2015-11-15

    Graphical abstract: - Highlights: • Exposure of materials (W, Ni, SS, Mo and Cu) to fusion plasma in a plasma focus device. • The erosion and the formations of blisters, pores, craters, micro-cracks after irradiation. • The structural phase transformation in the SS sample after irradiation. • The surface layer alloying of the samples with the plasma focus anode material. - Abstract: An 11.5 kJ plasma focus (PF) device was used here to irradiate materials with fusion grade plasma. The surface modifications of different materials (W, Ni, stainless steel, Mo and Cu) were investigated using various available techniques. The prominent features observed through the scanning electron microscope on the sample surfaces were erosions, cracks, blisters and craters after irradiations. The surface roughness of the samples increased multifold after exposure as measured by the surface profilometer. The X-ray diffraction analysis indicated the changes in the microstructures and the structural phase transformation in surface layers of the samples. We observed change in volumes of austenite and ferrite phases in the stainless steel sample. The energy dispersive X-ray spectroscopic analysis suggested alloying of the surface layer of the samples with elements of the PF anode. We report here the comparative analysis of the surface damages of materials with different physical, thermal and mechanical properties. The investigations will be useful to understand the behavior of the perspective materials for future fusion reactors (either in pure form or in alloy) over the long operations.

  17. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  18. Experimental facilities for investigation of structural material properties for fusion reactor under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G.M.; Strebkov, Yu.S.; Sidorenkov, A.V.; Zyryanov, A.P.; Barsanov, V.I.; Shushlebin, V.V. (Research and Development Inst. of Power Engineering, Moscow (Russia)); Rybin, V.V.; Vinokurov, V.F.; Odintsov, N.B. (Central Scientific and Research Inst. of Structural Materials, St. Petersburg (Russia)); Zykanov, V.A.; Shamardin, V.K.; Kazakov, V.A. (Scientific Research Inst. of Atomic Reactors, Dimitrovgrad (Russia))

    1992-09-01

    The study of sturctural and breeding materials for fusion reactors covers a wide range of investigations including the effect of different operating factors; irradiation is the main factor. This paper presents basic reactor characteristics, the types of investigations on structural and breeding materials carried out at these reactors, and the reactor irradiation conditions. The design of equipment used for parameter control during the irradiations is also discussed. CM-2 and BOR-60 reactors are primarily used to irradiate structural materials for the blanket, first wall and divertor at temperatures of 80 and 350deg C and fluences up to 5x10[sup 22] n/cm[sup 2]. The IVV-2 reactor is used to investigate breeding blanket materials and to study the problems of hydrogen/tritium permeability and recovery from Li-Pb eutectic and through 0.4C-16Cr-11Ni-3Mo-Ti steel. In addition, there are facilities for carrying out irradiation experiments at cryogenic temperatures as well as in different media. (orig.).

  19. The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies

    CERN Document Server

    Unholzer, S; Klein, H; Seidel, K

    2002-01-01

    The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...

  20. Model for collisional fast ion diffusion into Tokamak Fusion Test Reactor loss cone

    Energy Technology Data Exchange (ETDEWEB)

    Chang, C.S. [New York Univ., NY (United States). Courant Inst. of Mathematical Sciences]|[Korea Advanced Inst. of Science and Technology, Seoul (Korea, Republic of); Zweben, S.J.; Schivell, J.; Budny, R.; Scott, S. [Princeton Univ., NJ (United States). Plasma Physics Lab.

    1994-08-01

    An analytic model is developed to estimate the classical pitch angle scattering loss of energetic fusion product ions into prompt loss orbits in a tokamak geometry. The result is applied to alpha particles produced by deutrium-tritium fusion reactions in a plasma condition relevant to Tokamak Fusion Test Reactor (TFTR). A poloidal angular distribution of collisional fast ion loss at the first wall is obtained and the numerical result from the TRANSP code is discussed. The present model includes the effect that the prompt loss boundary moves away from the slowing-down path due to reduction in banana thickness, which enables us to understand, for the first time. the dependence of the collisional loss rate on Z{sub eff}.

  1. Avalanche boron fusion by laser picosecond block ignition with magnetic trapping for clean and economic reactor

    CERN Document Server

    Hora, H; Eliezer, S; Lalousis, N Nissim P; Giuffrida, L; Margarone, D; Picciotto, A; Miley, G H; Moustaizis, S; Martinez-Val, J -M; Barty, C P J; Kirchhoff, G J

    2016-01-01

    After the very long consideration of the ideal energy source by fusion of the protons of light hydrogen with the boron isotope 11 (boron fusion HB11) the very first two independent measurements of very high reaction gains by lasers basically opens a fundamental breakthrough. The non-thermal plasma block ignition with extremely high power laser pulses above petawatt of picosecond duration in combination with up to ten kilotesla magnetic fields for trapping has to be combined to use the measured high gains as proof of an avalanche reaction for an environmentally clean, low cost and lasting energy source as potential option against global warming. The unique HB11 avalanche reaction is are now based on elastic collisions of helium nuclei (alpha particles) limited only to a reactor for controlled fusion energy during a very short time within a very small volume.

  2. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  3. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    Energy Technology Data Exchange (ETDEWEB)

    DeVan, J.H.

    1979-01-01

    Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF--BeF/sub 2/, Pb--Li alloys, and solid ceramic compounds such as Li/sub 2/O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies.

  4. What we miss in order to be able to design and build a commercially viable fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Andreani, R. [ENEA, Centro Ricerche Frascati, RM (Italy). Dipt. Energia

    1999-07-01

    The paper considers in a critical way the different areas in which work is required to provide sufficient information in view of designing a reliable and attractive fusion reactor. [Italian] Il rapporto considera in modo critico le differenti aree nelle quali si richiede ulteriore lavoro per fornire informazioni al fine di progettare un reattore a fusione affidabile ed economicamente competitivo.

  5. Study of thorium-uranium based molten salt blanket in a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhao Jing, E-mail: zhao_jing@mail.tsinghua.edu.cn [INET, Tsinghua University, Beijing 100084 (China); Yang Yongwei; Zhou Zhiwei [INET, Tsinghua University, Beijing 100084 (China)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer A molten salt blanket has been designed for the fusion-fission hybrid reactor. Black-Right-Pointing-Pointer The use of Thorium in the molten salt fuels has been studied. Black-Right-Pointing-Pointer The molten salt was consisted of F-Li-Be and with the thickness of 40 cm. Black-Right-Pointing-Pointer The concentration of {sup 6}Li was chosen to be the natural enrichment ratio. Black-Right-Pointing-Pointer The result shows that TBR is greater than 1, M is about 15-16. - Abstract: Not only solid fuels, but also liquid fuels can be used for the fusion-fission symbiotic reactor blanket. The operational record of the molten salt reactor with F-Li-Be was very successful, so the F-Li-Be blanket was chosen for research. The molten salt has several features which are suited for the fusion-fission applications. The fuel material uranium and thorium were dissolved in the F-Li-Be molten salt. A combined program, COUPLE, was used for neutronics analysis of the molten salt blanket. Several cases have been calculated and compared. Not only the influence of the different fuels have been studied, but also the thickness of the molten salt, and the concentration of the {sup 6}Li in the molten salt. Preliminary studies indicate that when thorium-uranium-plutonium fuels were added into a F-Li-Be molten salt blanket and with a component of 71% LiF-2% BeF{sub 2}-13.5% ThF{sub 4}-8.5% UF{sub 4}-5% PuF{sub 3}, and also with the molten salt thickness of 40 cm and natural concentration of {sup 6}Li, the appropriate blanket energy multiplication factor and TBR can be obtained. The result shows that thorium-uranium molten salt can be used in the blanket of a fusion-fission symbiotic reactor. The research on the molten salt blanket must be valuable for the design of fusion-fission symbiotic reactor.

  6. ORNL fusion power demonstration study: the concept of the cassette blanket

    Energy Technology Data Exchange (ETDEWEB)

    Werner, R. W.

    1977-10-01

    The cassette blanket introduces four major improvements in fusion reactor blanket design. These are: (1) the cassette itself which by design furnishes the key unit for simplification of blanket replacement and maintenance and also isolates the lithium moderator from the plasma by enveloping it in the coolant; (2) the concept of blanket zoning, which uses to advantage the fact that radiation damage to structure decreases exponentially with distance. With the use of cassettes in series, only the front fraction of the blanket, the first cassette, need be changed due to damage over the life of the plant; (3) the rectangular blanket concept, which recognizes that blankets must envelop the plasma but need not conform to plasma shape. With this rectangular geometry, cassettes may be installed or removed by simple linear motion between magnet coils; (4) internal tritium recovery, which uses a favorable temperature gradient and ''MHD-frozen'' lithium to diffuse tritium out of the cassette. Supporting calculations and illustrative cases are provided for these four areas using two coolants: helium and HITEC, a eutectic mixture of inorganic salts (potassium nitrate, sodium nitrate, and sodium nitrite).

  7. Considerations for the development of neutral beam injection for fusion reactors or DEMO

    Science.gov (United States)

    Hemsworth, R. S.; Boilson, D.

    2017-08-01

    Neutral beam injection (NBI) has been the most successful heating scheme applied to fusion devices, the majority of which have been based on the acceleration and neutralization in a gas target of accelerated positive ions. For large fusion devices such as ITER, DEMO and fusion reactors, beam energies of the order of 0.5 MeV per nucleon or higher are required to penetrate deeply into the fusing plasma, and thus to heat the plasma in the most important region, i.e. near the poloidal axis of the device, and to drive current in the plasma. Because the efficiency of neutralization of positive ions in a gas target becomes unacceptably low at energies above ≈100 keV/nucleon, future injectors will be based on the neutralization of negative ions, either in a gas target, by photons or in a plasma target. So far only two systems based on negative ions have been used on fusion devices, at JT-60U and at LHD, both based on neutralization in a gas target. The injectors for ITER will also use a gas target, but the energy and operating environment are reactor and DEMO relevant. Also the ITER injectors will have to operate for pulse lengths orders of magnitude higher than all previous NBI systems. In this paper the R&D required for an NBI system for a reactor, or DEMO, is considered against the background of the ITER NBI system development, and the main elements of the required R&D are identified.

  8. 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.

  9. High-Yield Lithium-Injection Fusion-Energy (HYLIFE) reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blink, J.A.; Hogam, W.J.; Hovingh, J.; Meier, E.R.; Pitts, J.H. (comps.)

    1985-12-23

    The High-Yield Lithium-Injection Fusion Energy (HYLIFE) concept to convent inertial confinement fusion energy into electric power has undergone intensive research and refinement at LLNL since 1978. This paper reports on the final HYLIFE design, focusing on five major areas: the HYLIFE reaction chamber (which includes neutronics, liquid-metal jet-array hydrocynamics, and structural design), supporting systems, primary steam system and balance of plant, safety and environmental protection, and costs. An annotated bibliography of reports applicable to HYLIFE is also provided. We conclude that HYLIFE is a particularly viable concept for the safe, clean production of electrical energy. The liquid-metal jet array, HYLIFE's key design feature, protects the surrounding structural components from x-rays, fusion fuel-pellet debris, neutron damage and activation, and high temperatures and stresses, allowing the structure to last for the plant's entire 30-year lifetime without being replaced. 127 refs., 18 figs.

  10. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  11. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    Energy Technology Data Exchange (ETDEWEB)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.

  12. Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

    Science.gov (United States)

    Aleksandrova, I. V.; Koresheva, E. R.; Krokhin, O. N.; Osipov, I. E.

    2016-12-01

    In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.

  13. Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

    Energy Technology Data Exchange (ETDEWEB)

    Aleksandrova, I. V.; Koresheva, E. R., E-mail: elena.koresheva@gmail.com; Krokhin, O. N. [Russian Academy of Sciences, Lebedev Physical Institute (Russian Federation); Osipov, I. E. [Power Efficiency Centre, Inter RAO UES (Russian Federation)

    2016-12-15

    In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.

  14. Application of computational fluid dynamics (CFD) codes as design tools for inertial confinement fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, A; MartInez-Val, J M [Instituto de Fusion Nuclear, c/Jose Gutierrez Abascal 2, 28006 - Madrid (Spain); Sordo, F; Lafuente, A [Escuela Tecnica Superior de Ingenieros Industriales-UPM, c/Jose Gutierrez Abascal 2, 28006 - Madrid (Spain); Munoz, J [Fundacion para el Fomento de la Innovacion Industrial, c/Jose Gutierrez Abascal 2, 28006 - Madrid (Spain)], E-mail: abanades@etsii.upm.es

    2008-05-15

    The engineering design of the new innovative fusion reactors constitutes a clear challenge for the need to overcome several new technological edges in every engineering aspect. The great amount of thermal energy delivered into any inertial fusion chamber and the large temperatures and thermal gradients that are envisaged, joined to the even more demanding aspects related to neutron activation, Tritium breeding and the characteristics that are imposed to the coolant that could be used for that purpose, converged into material selection in which liquid metal seems to be one of the most interesting options. The safety assessment of such Fusion reactors should be clearly provided to fulfill the requirements asked by the Regulatory Bodies in a near-term future, when licensing will be a must. Therefore the availability of well proven and validated engineering design tools is a must. In this context, CFD is one of the tools that are potentially needed for thermal-hydraulic design of such complex machines. The state-of-the-art of CFD technologies will be shown, in particular in relation with liquid metals.

  15. Application of computational fluid dynamics (CFD) codes as design tools for inertial confinement fusion reactor

    Science.gov (United States)

    Abánades, A.; Sordo, F.; Lafuente, A.; Muñoz, J.; Martínez-Val, J. M.

    2008-05-01

    The engineering design of the new innovative fusion reactors constitutes a clear challenge for the need to overcome several new technological edges in every engineering aspect. The great amount of thermal energy delivered into any inertial fusion chamber and the large temperatures and thermal gradients that are envisaged, joined to the even more demanding aspects related to neutron activation, Tritium breeding and the characteristics that are imposed to the coolant that could be used for that purpose, converged into material selection in which liquid metal seems to be one of the most interesting options. The safety assessment of such Fusion reactors should be clearly provided to fulfill the requirements asked by the Regulatory Bodies in a near-term future, when licensing will be a must. Therefore the availability of well proven and validated engineering design tools is a must. In this context, CFD is one of the tools that are potentially needed for thermal-hydraulic design of such complex machines. The state-of-the-art of CFD technologies will be shown, in particular in relation with liquid metals.

  16. X-rays from Proton Bremsstrahlung: Evidence from Fusion Reactors and Its Implication in Astrophysics

    CERN Document Server

    Luo, Nie

    2009-01-01

    In a fusion reactor, a proton and a neutron generated in previous reactions may again fuse with each other. Or they can in turn fuse with or be captured by an un-reacted deuteron. The average center-of-mass (COM) energy for such reaction is around 10 keV in a typical fusion reactor, but could be as low as 1 keV. At this low COM energy, the reacting nucleons are in an s-wave state in terms of their relative angular momentum. The single-gamma radiation process is thus strongly suppressed due to conservation laws. Instead the gamma ray released is likely to be accompanied by x-ray photons from a nuclear bremsstrahlung process. The x-ray thus generated has a continuous spectrum and peaks around a few hundred eV to a few keV. The average photon energy and spectrum properties of such a process are calculated with a semiclassical approach. The results give a peak near 1.1 keV for the proton-deuteron fusion and a power-to-the-minus-second law in the spectrum's high-energy limit. An analysis of some prior tokamak disc...

  17. ORNL fusion power demonstration study: arguments for a vacuum building in which to enclose a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Werner, R.W.

    1976-12-01

    Fusion reactors as presently contemplated are excessively complicated, are virtually inaccessible for some repairs, and are subject to frequent loss of function. This dilemma arises in large part because the closed surface that separates the ''hard'' vacuum of the plasma zone from atmospheric pressure is located at the first wall or between blanket and shield. This closed surface is one containing hundreds to thousands of linear meters of welds or mechanical seals which are subject to radiation damage and cyclic fatigue. In situ repair is extremely difficult. This paper examines the arguments favoring the enclosing of the entire reactor in a vacuum building and thus changing the character of this closed surface from one requiring absolute vacuum integrity to one of high pumping impedance. Two differentially pumped vacuum zones are imagined, one clean zone for the plasma and one for the balance of the volume. Both would be at substantially the same pressure. Other advantages for the vacuum enclosure are also cited and discussed.

  18. Mitigation of the hydrogen risk in fusion reactors; Mitigation du risque hydrogene dans les reacteurs de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Maruejouls, C.; Robin, J.C. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs; Arnould, F.; Bachellerie, E. [Technicatome DI SEPS, 13 - Aix en Provence (France); Latge, C. [CEA Cadarache, Dept. d' Etudes des Dechets DED, 13 - Saint Paul lez Durance (France); Laurent, A. [Institut National Polytechnique de Lorraine, Lab. des Sciences du Genie Chimique, 54 - Nancy (France)

    2001-07-01

    The rupture of the first wall and the intrusion of water vapor inside the torus, is one of the major accident that can occur in a thermonuclear fusion reactor. In this situation, water oxidizes the beryllium of the wall and the reaction produces hydrogen with a strong risk of explosion inside the reactor. In order to mitigate this risk, a process based on the reduction of metal oxides (MnO{sub 2}, Ag{sub 2}O) has been developed. The aim of this study is the determination of the kinetics of this reduction reaction. A mixture of both oxides has been deposited on the surface of porous balls for an experiment on fixed beds. The modeling of the phenomenon is based on the equations used in heterogenous catalysis and the experimental determination of the kinetics of the reaction is performed with the CIGNE test-facility. The velocity of the reduction reaction is deduced from the remaining amount of hydrogen in the test-gas (N{sub 2} with 1 to 2% of H{sub 2}) after it has been flowed on the oxides coated balls of the fixed bed. (J.S.)

  19. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  20. Simulation of plasma-surface interactions in a fusion reactor by means of QSPA plasma streams: recent results and prospects

    Science.gov (United States)

    Garkusha, I. E.; Aksenov, N. N.; Byrka, O. V.; Makhlaj, V. A.; Herashchenko, S. S.; Malykhin, S. V.; Petrov, Yu V.; Staltsov, V. V.; Surovitskiy, S. V.; Wirtz, M.; Linke, J.; Sadowski, M. J.; Skladnik-Sadowska, E.

    2016-09-01

    This paper is devoted to plasma-surface interaction issues at high heat-loads which are typical for fusion reactors. For the International Thermonuclear Experimental Reactor (ITER), which is now under construction, the knowledge of erosion processes and the behaviour of various constructional materials under extreme conditions is a very critical issue, which will determine a successful realization of the project. The most important plasma-surface interaction (PSI) effects in 3D geometry have been studied using a QSPA Kh-50 powerful quasi-stationary plasma accelerator. Mechanisms of the droplet and dust generation have been investigated in detail. It was found that the droplets emission from castellated surfaces has a threshold character and a cyclic nature. It begins only after a certain number of the irradiating plasma pulses when molten and shifted material is accumulated at the edges of the castellated structure. This new erosion mechanism, connected with the edge effects, results in an increase in the size of the emitted droplets (as compared with those emitted from a flat surface). This mechanism can even induce the ejection of sub-mm particles. A concept of a new-generation QSPA facility, the current status of this device maintenance, and prospects for further experiments are also presented.

  1. Pacific Northwest Laboratory report on fusion reactor technology, April 1976 - June 1976

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-07-01

    This quarterly report consists of progress summaries of research conducted by the staff of Pacific Northwest Laboratories (PNL). This reporting period includes progress made from April 1, 1976 through June 30, 1976. The summaries are presented in four major categories of: (1) fusion systems engineering, (2) material research and radiation environment simulation, (3) environmental effects of fusion concepts, and (4) manpower development. At the beginning of each section is a brief summary of the reports making up the section. The reports themselves have been kept relatively short and include preliminary results which ultimately are expected to be published elsewhere.

  2. Selected transport studies of a tokamak-based DEMO fusion reactor

    Science.gov (United States)

    Fable, E.; Wenninger, R.; Kemp, R.

    2017-02-01

    As a next-step in the tokamak-based fusion programme, the DEMO fusion reactor is foreseen to produce relevant output electricity, in the order of  ∼500 MW delivered to the network. The scenarios that are being presently investigated consist of a pulsed device, called DEMO1, and a steady-state device, called DEMO2. In this work, which is focused on the pulsed device DEMO1, scenarios are studied from the point of view of core transport, to assess plasma performance and limitations due to core microinstabilities. The role of radiated power, aspect ratio, and height of temperature pedestal are assessed as they impact both core energy and particle transport. Open issues in this framework are also discussed.

  3. Introduction to the special issue on the technical status of materials for a fusion reactor

    Science.gov (United States)

    Stork, D.; Zinkle, S. J.

    2017-09-01

    Materials determine in a fundamental way the performance and environmental attractiveness of a fusion reactor: through the size (power fluxes to the divertor, neutron fluxes to the first wall); economics (replacement lifetime of critical in-vessel components, thermodynamic efficiency through operating temperature etc); plasma performance (erosion by plasma fluxes to the divertor surfaces); robustness against off-normal accidents (safety); and the effects of post-operation radioactivity on waste disposal and maintenance. The major philosophies and methodologies used to formulate programmes for the development of fusion materials are outlined, as the basis for other articles in this special issue, which deal with the fundamental understanding of the issues regarding these materials and their technical status and prospects for development.

  4. Optimisation of confinement in a fusion reactor using a nonlinear turbulence model

    CERN Document Server

    Highcock, E G; Barnes, M; Dorland, W

    2016-01-01

    The confinement of heat in the core of a magnetic fusion reactor is optimised using a multidimensional optimisation algorithm. For the first time in such a study, the loss of heat due to turbulence is modelled at every stage using first-principles nonlinear simulations which accurately capture the turbulent cascade and large-scale zonal flows. The simulations utilise a novel approach, with gyrofluid treatment of the small-scale drift waves and gyrokinetic treatment of the large-scale zonal flows. A simple near-circular equilibrium with standard parameters is chosen as the initial condition. The figure of merit, fusion power per unit volume, is calculated, and then two control parameters, the elongation and triangularity of the outer flux surface, are varied, with the algorithm seeking to optimise the chosen figure of merit. An optimal configuration is discovered at an elongation of 1.5 and a triangularity of 0.03.

  5. Recent advances in physics and technology of ion cyclotron resonance heating in view of future fusion reactors

    Science.gov (United States)

    Ongena, J.; Messiaen, A.; Kazakov, Ye O.; Koch, R.; Ragona, R.; Bobkov, V.; Crombé, K.; Durodié, F.; Goniche, M.; Krivska, A.; Lerche, E.; Louche, F.; Lyssoivan, A.; Vervier, M.; Van Eester, D.; Van Schoor, M.; Wauters, T.; Wright, J.; Wukitch, S.

    2017-05-01

    Ion temperatures of over 100 million degrees need to be reached in future fusion reactors for the deuterium-tritium fusion reaction to work. Ion cyclotron resonance heating (ICRH) is a method that has the capability to directly heat ions to such high temperatures, via a resonant interaction between the plasma ions and radiofrequency waves launched in the plasma. This paper gives an overview of recent developments in this field. In particular a novel and recently developed three-ion heating scenario will be highlighted. It is a flexible scheme with the potential to accelerate heavy ions to high energies in high density plasmas as expected for future fusion reactors. New antenna designs will be needed for next step large future devices like DEMO, to deliver steady-state high power levels, cope with fast variations in coupling due to fast changes in the edge density and to reduce the possibility for impurity production. Such a new design is the traveling wave antenna (TWA) consisting of an array of straps distributed around the circumference of the machine, which is intrinsically resilient to edge density variations and has an optimized power coupling to the plasma. The structure of the paper is as follows: to provide the general reader with a basis for a good understanding of the later sections, an overview is given of wave propagation, coupling and RF power absorption in the ion cyclotron range of frequencies, including a brief summary of the traditionally used heating scenarios. A special highlight is the newly developed three-ion scenario together with its promising applications. A next section discusses recent developments to study edge-wave interaction and reduce impurity influx from ICRH: the dedicated devices IShTAR and Aline, field aligned and three-strap antenna concepts. The principles behind and the use of ICRH as an important option for first wall conditioning in devices with a permanent magnetic field is discussed next. The final section presents ongoing

  6. High-power-density approaches to magnetic fusion energy: Problems and promise of compact reversed-field pinch reactors (CRFPR)

    Science.gov (United States)

    Hagenson, Randy L.; Krakowski, Robert A.; Dreicer, Harry

    1983-03-01

    If the costing assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves overly optimistic, approaches that promise considerably increased system power density and reduced mass utilization will be required. These more compact reactor embodiments generally must operate with reduced shield thickness and resistive magnets. Because of the unique magnetic topology associated with the Reversed-Field Pinch (RFP), the compact reactor embodiment for this approach is particularly attractive from the viewpoint of low-field resistive coils operating with ohmic losses that can be made small relative to the fusion power. The RFP, therefore, is used as one example of a high-power-density (HPD) approach to magnetic fusion energy. A comprehensive system model is described and applied to select a unique, cost-optimized design point that will be used for a subsequent conceptual engineering design of the compact RFP Reactor (CRFPR). This cost-optimized CRFPR design serves as an example of a HPD fusion reactor that would operate with system power densities and mass utilizations that are comparable to fission power plants, these measures of system performance being an order of magnitude more favorable than the conventional approaches to magnetic fusion energy (MFE).

  7. Treatment of irradiation effects in structural design criteria for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S. [Argonne National Lab., IL (United States). Fusion Power Program; Smith, P. [San Diego Joint Work Site, CA (United States). ITER Joint Central Team

    1997-03-01

    The irradiation environment experienced by the in-vessel components of fusion reactors such as the International Thermonuclear Experimental Reactor (ITER) presents structural design challenges not envisioned in the development of existing structural design criteria such as the ASME Code or RCC-MR. From the standpoint of structural design criteria, the most significant issues stem from the irradiation-induced changes in material properties, specifically the reduction of ductility, strain hardening capability, and fracture toughness with neutron irradiation. These effects call into question the basis of the design rules in existing structural design criteria which assume that only code-approved materials with high toughness, ductility and strain hardening capability will be used. The present paper reviews the basis of new rules that address these issues in Draft 5 of the interim ITER structural design criteria (ISDC) which was released recently for trial use by the ITER designers.

  8. Potential for use of high-temperature superconductors in fusion reactors

    Science.gov (United States)

    Hull, John R.

    1992-09-01

    The present rate of development of high-temperature superconductors (HTSs) is sufficiently rapid that there may be opportunities for their use in contemporary fusion devices such as the International Thermonuclear Experimental Reactor (ITER). The most likely application is for delivering power to the superconducting magnets, especially in substituting for the current leads between the temperatures of 4 and 77 K. A second possible application of HTSs is in a liquid-nitrogen-cooled power bus connecting the power supplies to the magnets, thus reducing ohmic heating losses in these relatively long cables. A third potential application of HTSs is in inner high-field windings of the toroidal field coils that would operate at ~ 20 K. While the use of higher temperature magnets offers significant advantages to the reactor system, it is unlikely that tested HTSs for this application will be available within the ITER time frame.

  9. Thermal-hydraulic characteristics in a tokamak vacuum vessel of fusion reactor after transient events occurred

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Kazuyuki; Kunugi, Tomoaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Seki, Yasushi

    1997-12-31

    The thermal-hydraulic characteristics in a vacuum vessel (VV) of fusion reactor under the ingress-of-coolant-event (ICE) or loss-of-vacuum-event (LOVA) condition were carried out to investigate experimentally the thermofluid safety in the International Thermonuclear Experimental Reactor (ITER) under transient events. In the ICE experiments, the pressure rise and wall temperatures in the VV were measured and the performance of a suppression tank was confirmed. In the LOVA experiments, the exchange time inside the VV from the vacuum to be the atmospheric pressure was measured for various breach size and the exchange flow rates through the breaches of the VV under the atmospheric pressure conditions were clarified. (author)

  10. Fusion core start-up, ignition and burn simulations of reversed-field pinch (RFP) reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chu, Yuh-Yi

    1988-01-01

    A transient reactor simulation model is developed to investigate and simulate the start-up, ignition and burn of a reversed-field pinch reactor. The simulation is based upon a spatially averaged plasma balance model with field profiles obtained from MHD quasi-equilibrium analysis. Alpha particle heating is estimated from Fokker-Planck calculations. The instantaneous plasma current is derived from a self-consistent circuit analysis for plasma/coil/eddy current interactions. The simulation code is applied to the TITAN RFP reactor design which features a compact, high-power-density reversed-field pinch fusion system. A contour analysis is performed using the steady-state global plasma balance. The results are presented with contours of constant plasma current. A saddle point is identified in the contour plot which determines the minimum value of plasma current required to achieve ignition. An optimized start-up to ignition and burn path can be obtained by passing through the saddle point. The simulation code is used to study and optimize the start-up scenario. In the simulations of the TITAN RFP reactor, the OH-driven superconducting EF coils are found to deviate from the required equilibrium values as the induced plasma current increases. This results in the modification of superconducting EF coils and the addition of a set of EF trim coils. The design of the EF coil system is performed with the simulation code subject to the optimization of trim-coil power and current. In addition, the trim-coil design is subject to the constraints of vertical-field stability index and maintenance access. A power crowbar is also needed to prevent the superconducting EF coils from generating excessive vertical field. A set of basic results from the simulation of TITAN RFP reactor yield a picture of RFP plasma operation in a reactor. Investigations of eddy current are also presented. 145 refs., 37 figs., 2 tabs.

  11. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  12. Note: Readout of a micromechanical magnetometer for the ITER fusion reactor.

    Science.gov (United States)

    Rimminen, H; Kyynäräinen, J

    2013-05-01

    We present readout instrumentation for a MEMS magnetometer, placed 30 m away from the MEMS element. This is particularly useful when sensing is performed in high-radiation environment, where the semiconductors in the readout cannot survive. High bandwidth transimpedance amplifiers are used to cancel the cable capacitances of several nanofarads. A frequency doubling readout scheme is used for crosstalk elimination. Signal-to-noise ratio in the range of 60 dB was achieved and with sub-percent nonlinearity. The presented instrument is intended for the steady-state magnetic field measurements in the ITER fusion reactor.

  13. Modifications Made to the MELCOR Code for Analyzing Lithium Fires in Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad Johnson

    2000-04-01

    This report documents initial modifications made to the MELCOR code that allows MELCOR to predict the consequences of lithium spill accidents for evolving fusion reactor designs. These modifications include thermodynamic and transport properties for lithium, and physical models for predicting the rate of reaction of and energy production from the lithium-air reaction. A benchmarking study was performed with this new MELCOR capability. Two lithium-air reaction tests conducted at the Hanford Engineering Development Laboratory (HEDL) were selected for this benchmark study. Excellent agreement was achieved between MELCOR predictions and measured data. Recommendations for modeling lithium fires with MELCOR and for future work in this area are included in this report.

  14. Modifications made to the MELCOR Code for Analyzing Lithium Fires in Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    B. J. Merrill

    2000-04-01

    This report documents initial modifications made to the MELCOR code that allows MELCOR to predict the consequences of lithium spill accidents for evolving fusion reactor designs. These modifications include thermodynamic and transport properties for lithium, and physical models for predicting the rate of reaction of and energy production from the lithium-air reaction. A benchmarking study was performed with this new MELCOR capability. Two lithium-air reaction tests conducted at the Hanford Engineering Development Laboratory (HEDL) were selected for this benchmark study. Excellent agreement was achieved between MELCOR predictions and measured data. Recommendations for modeling lithium fires with MELCOR and for future work in this area are included in this report.

  15. 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.

  16. Scientific report. Plasma-wall interaction studies related to fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Temmerman, G. De

    2006-07-01

    This scientific report summarises research done on erosion and deposition mechanisms affecting the optical reflectivity of potential materials for use in the mirrors used in fusion reactors. Work done in Juelich, Germany, at the Federal Institute of Technology in Lausanne, Switzerland, the JET laboratory in England and in Basle is discussed. Various tests made with the mirrors are described. Results obtained are presented in graphical and tabular form and commented on. The influence of various material choices on erosion and deposition mechanisms is discussed.

  17. Polarized SANS study of microstructural evolution in a martensitic steel for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coppola, R. [ENEA-Casaccia, FIS, CP 2400, 00100 Roma (Italy); Glaettli, H. [CEA-SACLAY, SPEC and Laboratoire Leon Brillouin (CEA-CNRS), 91191 Gif-sur-Yvette (France); Valli, M. [ENEA-Bologna, FIS, V. Don Fiammelli 2, 40128 Bologna (Italy)

    2002-07-01

    The results of a polarized SANS study of a martensitic steel (MANET) developed for fusion-reactor technology are presented. The measurements were carried out to investigate Cr-redistribution phenomena in the martensitic matrix, which can play a crucial role in ductile-to-brittle transition changes under irradiation. The nuclear-magnetic interference term and the ratio of nuclear plus magnetic to nuclear SANS cross sections show that such inhomogeneities, which are present immediately after quenching and give rise to Fe-rich precipitates, dissolve even for short tempering times. (orig.)

  18. Neutronics optimization study for D-D fusion reactor blanket/shield

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, T.; Kanda, Y.; Nakashima, H.

    1985-12-01

    Position-dependent optimization calculations have been carried out on a D-D fusion reactor blanket/shield to maximize the energy gain in the blanket and to minimize the atomic displacement rate of the copper stabilizer in the superconducting magnet. The results obtained by using the optimization code SWAN indicate the advantage of D/sub 2/O coolant over H/sub 2/O coolant with respect to increasing the energy gain, and the difference in the optimal shield distributions between D-T and D-D neutron sources. The possibility of improving both the energy gain and radiation shielding characteristics is also discussed.

  19. Fusion reactor materials: Semiannual progress report for period ending September 30, 1986

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1987-09-01

    These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The major areas of concern covered in this report are irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters and activation calculations; materials engineering and design requirements; radiation effects; development of structural alloys; solid breeding materials; ceramics and superconducting magnet materials. There are 61 reports cataloged separately. (LSP)

  20. On the implementation of new technology modules for fusion reactor systems codes

    Energy Technology Data Exchange (ETDEWEB)

    Franza, F., E-mail: fabrizio.franza@kit.edu [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Boccaccini, L.V.; Fisher, U. [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Gade, P.V.; Heller, R. [Institute for Technical Physics, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany)

    2015-10-15

    Highlights: • At KIT a new technology modules for systems code are under development. • A new algorithm for the definition of the main reactor's components is defined. • A new blanket model based on 1D neutronics analysis is described. • A new TF coil stress model based on 3D electromagnetic analysis is described. • The models were successfully benchmarked against more detailed models. - Abstract: In the frame of the pre-conceptual design of the next generation fusion power plant (DEMO), systems codes are being used from nearly 20 years. In such computational tools the main reactor components (e.g. plasma, blanket, magnets, etc.) are integrated in a unique computational algorithm and simulated by means of rather simplified mathematical models (e.g. steady state and zero dimensional models). The systems code tries to identify the main design parameters (e.g. major radius, net electrical power, toroidal field) and to make the reactor's requirements and constraints to be simultaneously accomplished. In fusion applications, requirements and constraints can be either of physics or technology kind. Concerning the latest category, at Karlsruhe Institute of Technology a new modelling activity has been recently launched aiming to develop improved models focusing on the main technology areas, such as neutronics, thermal-hydraulics, electromagnetics, structural mechanics, fuel cycle and vacuum systems. These activities started by developing: (1) a geometry model for the definition of poloidal profiles for the main reactors components, (2) a blanket model based on neutronics analyses and (3) a toroidal field coil model based on electromagnetic analysis, firstly focusing on the stresses calculations. The objective of this paper is therefore to give a short outline of these models.

  1. The concept of isochoric central spark ignition and its fuel gain in inertial fusion

    CERN Document Server

    Kuzmin, A D

    2004-01-01

    One of the best methods in inertial confinement fusion (ICF) is the concept of central spark ignition, consisting of two distinct regions named as hot and cold regions and formed by hydro-dynamical implosion of fuel micro-sphere central spark ignition method in inertial fusion and fuel pellet design condition in fusion power plant has been investigated and fuel gain for isochoric model in this method is calculated. We have shown the effects of different physical parameters of inertial fusion on fuel gain and optimized limit for fuel density and fuel pellet radius has been calculated.

  2. Estimation of the environmental or radiological impact in the event of accidental release of radionuclides in a DCLL fusion reactor; Estimacion del impacto radiologico ambiental en caso de liberacion accidental de radionucleidos en un reactor de fusion DCLL

    Energy Technology Data Exchange (ETDEWEB)

    Palermo, I.; Gomez Ros, J. M.; Sanz, J.; Mota, F.

    2013-07-01

    Tritium production and activation in the LiPb products can pose a radiological risk in the event of accidental release in a fusion reactor. Within the research programme Consolider TECNO{sub F}US (CSD2008-079) fusion technology has developed a design for a reactor with regenerative wrap with dual refrigeration (DCLL). The purpose of this communication is to present estimates of the radiological impact derived from an accidental release of radionuclides from the circuit of LiPb provinients. (Author)

  3. Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor

    Science.gov (United States)

    Ragheb, Magdi M. H.; Maynard, Charles W.

    1981-10-01

    A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components. To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×1015 n/(m2·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 1017 n/(m2·s) for a 1 MW/m2 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.

  4. Process Model of A Fusion Fuel Recovery System for a Direct Drive IFE Power Reactor

    Science.gov (United States)

    Natta, Saswathi; Aristova, Maria; Gentile, Charles

    2008-11-01

    A task has been initiated to develop a detailed representative model for the fuel recovery system (FRS) in the prospective direct drive inertial fusion energy (IFE) reactor. As part of the conceptual design phase of the project, a chemical process model is developed in order to observe the interaction of system components. This process model is developed using FEMLAB Multiphysics software with the corresponding chemical engineering module (CEM). Initially, the reactants, system structure, and processes are defined using known chemical species of the target chamber exhaust. Each step within the Fuel recovery system is modeled compartmentally and then merged to form the closed loop fuel recovery system. The output, which includes physical properties and chemical content of the products, is analyzed after each step of the system to determine the most efficient and productive system parameters. This will serve to attenuate possible bottlenecks in the system. This modeling evaluation is instrumental in optimizing and closing the fusion fuel cycle in a direct drive IFE power reactor. The results of the modeling are presented in this paper.

  5. Study of safety features and accident scenarios in a fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, M., E-mail: nakamura.makoto@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Tobita, K. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Gulden, W. [Fusion for Energy, c/o EFDA Garching and Max-Plank-Institut fuer Plasmaphysik, Garching D-85748 (Germany); Watanabe, K. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Toshiba Corporation, Yokohama, Kanagawa 235-8523 (Japan); Someya, Y. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Tanigawa, H. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Sakamoto, Y. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Araki, T.; Matsumiya, H.; Ishii, K. [Toshiba Corporation, Yokohama, Kanagawa 235-8523 (Japan); Utoh, H. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Takase, H. [IFERC Project Team, Rokkasho, Aomori 039-3212 (Japan); Hayashi, T. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Satou, A.; Yonomoto, T. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Federici, G. [Fusion for Energy, c/o EFDA Garching and Max-Plank-Institut fuer Plasmaphysik, Garching D-85748 (Germany); Okano, K. [IFERC Project Team, Rokkasho, Aomori 039-3212 (Japan)

    2014-10-15

    Highlights: This paper reports progress in the fusion DEMO safety research conducted under the Broader Approach DEMO Design Activities; Hazards of a reference DEMO concept have been assessed; Reference accident event sequences in the reference DEMO in this study have been analyzed based on the master logic diagram (MLD) and the functional failure mode and effect analysis (FFMEA) techniques; Accident events of particular concern in the DEMO have been selected based on the MLD and FFMEA analysis. Abstract: After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li{sub 2}TiO{sub 3}) and beryllium–titanium (Be{sub 12}Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment.

  6. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  7. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  8. Comparison of the Recently proposed Super Marx Generator Approach to Thermonuclear Ignition with the DT Laser Fusion-Fission Hybrid Concept (LIFE) by the Lawrence Livermore National Laboratory.

    Science.gov (United States)

    Winterberg, Friedwardt

    2009-05-01

    The recently proposed Super Marx pure deuterium micro-detonation ignition concept [1] is compared to the Lawrence Livermore National Ignition Facility (NIF) laser DT fusion-fission hybrid concept (LIFE) [2]. A typical example of the LIFE concept is a fusion gain 30, and a fission gain of 10, making up for a total gain of 300, with about 10 times more energy released into fission as compared to fusion. This means a substantial release of fission products, as in fusion-less pure fission reactors. In the Super Marx approach for the ignition of a pure deuterium micro-detonation gains of the same magnitude can in theory be reached. If the theoretical prediction can be supported by more elaborate calculations, the Super Marx approach is likely to make lasers obsolete as a means for the ignition of thermonuclear micro-explosions. [1] ``Ignition of a Deuterium Micro-Detonation with a Gigavolt Super Marx Generator,'' Winterberg, F., Journal of Fusion Energy, Springer, 2008. http://www.springerlink.com/content/r2j046177j331241/fulltext.pdf. [2] ``LIFE: Clean Energy from Nuclear Waste,'' https://lasers.llnl.gov/missions/energy&_slash;for&_slash;the&_slash;future/life/

  9. Analysis of Induced Gamma Activation by D-T Neutrons in Selected Fusion Reactor Relevant Materials with EAF-2010

    Science.gov (United States)

    Klix, Axel; Fischer, Ulrich; Gehre, Daniel

    2016-02-01

    Samples of lanthanum, erbium and titanium which are constituents of structural materials, insulating coatings and tritium breeder for blankets of fusion reactor designs have been irradiated in a fusion peak neutron field. The induced gamma activities were measured and the results were used to check calculations with the European activation system EASY-2010. Good agreement for the prediction of major contributors to the contact dose rate of the materials was found, but for minor contributors the calculation deviated up to 50%.

  10. Comparative study of survived displacement damage defects in iron irradiated in IFMIF and fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simakov, S.P. [Forschungszentrum Karlsruhe, Institute for Reactor Safety, D-76021 Karlsruhe (Germany)], E-mail: simakov@irs.fzk.de; Konobeyev, A.Yu.; Fischer, U.; Heinzel, V. [Forschungszentrum Karlsruhe, Institute for Reactor Safety, D-76021 Karlsruhe (Germany)

    2009-04-30

    The assessment of the primary survived defects rates in iron such as vacancies-interstitials pairs and simplest clusters have been performed for the IFMIF, fusion power plant and research reactor. This was achieved by a modified version of the NJOY code, when processing evaluated nuclear cross section file. The modifications accounted for the reduction of the available damage energy predicted by the standard NRT model by the surviving defects fractions. These fractions were picked-up from the molecular dynamics and binary collisions simulation results available in the literature. The number of primary survived and clustered defects in the {alpha}-iron irradiated in the high flux test module of IFMIF was estimated as 10 and 6 dpa/fpy or several times less than standard NRT estimates at the level of 30 dpa/fpy. The comparison with damages in iron calculated for irradiation in the first wall of fusion power plant gave however the same reduction factors, that supports the qualification of IFMIF as a fusion material irradiation facility.

  11. Anomalous fast ion losses at high β on the tokamak fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2015-03-15

    This paper describes experiments carried out on the Tokamak Fusion Test Reactor (TFTR) [R. J. Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] to investigate the dependence of β-limiting disruption characteristics on toroidal field strength. The hard disruptions found at the β-limit in high field plasmas were not found at low field, even for β's 50% higher than the empirical β-limit of β{sub n} ≈ 2 at high field. Comparisons of experimentally measured β's to TRANSP simulations suggest anomalous loss of up to half of the beam fast ions in the highest β, low field shots. The anomalous transport responsible for the fast ion losses may at the same time broaden the pressure profile. Toroidal Alfvén eigenmodes, fishbone instabilities, and Geodesic Acoustic Modes are investigated as possible causes of the enhanced losses. Here, we present the first observations of high frequency fishbones [F. Zonca et al., Nucl. Fusion 49, 085009 (2009)] on TFTR. The interpretation of Axi-symmetric Beam-driven Modes as Geodesic Acoustic Modes and their possible correlation with transport barrier formation are also presented.

  12. Burn control of an ITER-like fusion reactor using fuzzy logic

    Science.gov (United States)

    Garcia-Amador, A. Sair; Martinell, Julio J.

    2016-10-01

    The fuel burn in a fusion reactor has to be kept at a nearly constant rate in order to have a steady power exhaust. Here, we develop a control system based on a fuzzy logic controller in order that adjusts external parameters to keep the plasma temperature and density at the design values of a reactor of the characteristics of ITER. The control parameters chosen are the D-T refueling rate, the auxiliary heating power and a neutral helium beam. We use a fuzzy controller of the Mamdani type that uses a number of membership functions appropriate to produce a response to parameter deviations that minimizes the response time. The inference rules are determined in a way to provide stabilization to all perturbations of the temperature, density and alpha particle fraction. The dynamical response of the reactor is simulated with a 0D model that uses confinement times provided by the ITER scaling. We show that the system is feedback stabilized for a large range of parameters around the nominal values. The recovery time after a departure from the steady values is of the order of one second. We compare the results with another control system based on neural networks that was developed previously. Funded by projects PAPIIT IN109115 and Conacyt 152905.

  13. Design of a fusion reactor for Pb-Li eutectic alloy; Diseno de un reactor de fusion para aleaciones eutecticas Pb-Li

    Energy Technology Data Exchange (ETDEWEB)

    Quinones, J.; Barrena, M. I.; Gomez de Salazar, J. M.; Serrano, L.; Durana, S.; Conde, E.; Barrado, A. I.; Fernandez, M.; Sedano, L.; Soria, A.

    2011-07-01

    Nowadays, Pb-Li eutectic alloy shows an increase interest due to its new industrial applications in the field of production of tritium. This isotope is critical in the nuclear fusion reactor. In this paper, it is presented the design, construction and commissioning of a melting furnace of Pb-Li eutectic alloy and the optimization procedure necessary in order to produce the alloy. In the study presented, different parameters have been taken into account, such as: environmental conditions (shielding gas), melting temperature (375-800 degree centigrade); the possibility of using slag with different chemical composition (i.e.; LiCl-Kcl eutectic molten salt). The alloys obtained were characterized both from an analytical point of view as microstructural. These characterization results allow to determine, for example: the range of %w Li in the ingots, presence of different phases, possible micro segregations effect, etc. The characterization techniques used were X-ray diffraction, scanning electron microscopy and chemical analysis by ICP-MS. From these studies and the results presented in this paper it is possible to point out that the furnace designed for obtaining ingots of Pb-Li eutectic composition allows to produces very homogeneous ingots. (Author)

  14. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  15. Materials for DEMO and reactor applications—boundary conditions and new concepts

    Science.gov (United States)

    Coenen, J. W.; Antusch, S.; Aumann, M.; Biel, W.; Du, J.; Engels, J.; Heuer, S.; Houben, A.; Hoeschen, T.; Jasper, B.; Koch, F.; Linke, J.; Litnovsky, A.; Mao, Y.; Neu, R.; Pintsuk, G.; Riesch, J.; Rasinski, M.; Reiser, J.; Rieth, M.; Terra, A.; Unterberg, B.; Weber, Th; Wegener, T.; You, J.-H.; Linsmeier, Ch

    2016-02-01

    DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER towards realizing fusion. For the realization of fusion energy especially, materials questions pose a significant challenge already today. Heat, particle and neutron loads are a significant problem to material lifetime when extrapolating to DEMO. For many of the issues faced, advanced materials solutions are under discussion or already under development. In particular, components such as the first wall and the divertor of the reactor can benefit from introducing new approaches such as composites or new alloys into the discussion. Cracking, oxidation as well as fuel management are driving issues when deciding for new materials. Here {{{W}}}{{f}}/{{W}} composites as well as strengthened CuCrZr components together with oxidation resilient tungsten alloys allow the step towards a fusion reactor. In addition, neutron induced effects such as transmutation, embrittlement and after-heat and activation are essential. Therefore, when designing a component an approach taking into account all aspects is required.

  16. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  17. Data fusion: a new concept in non-destructive testing; Fusion de donnees: un nouveau concept en controle non destructif

    Energy Technology Data Exchange (ETDEWEB)

    Georgel, B.; Lavayssiere, B.

    1995-12-31

    Non-destructive testing of some components (made of austenitic steel, or of a complex shape for example) requires quite often the use of several methods such as X-ray, ultrasonics, Eddy Currents. Then, a skilled operator is able to perform the expertise of the specimen. The main goal of this paper is to show that 3D diagnosis may be improved in term of reliability and precision by fusion of several NDT techniques. A data fusion algorithm is more that trying to improve the visualisation or the rendering of NDT data sets. It consists for each volume element, in computing a new value representing the combined information and in formulating a diagnosis on this basis. To achieve such a goal, know-how in modeling of physical phenomena and in applied mathematics is crucial. (authors). 4 refs., 2 figs.

  18. Evaluation of Nb-base alloys for the divertor structure in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Purdy, I.M. [Argonne National Laboratory, Upton, IL (United States)

    1996-04-01

    Niobium-base alloys are candidate materials for the divertor structure in fusion reactors. For this application, an alloy should resist aqueous corrosion, hydrogen embrittlement, and radiation damage and should have high thermal conductivity and low thermal expansion. Results of corrosion and embrittlement screening tests of several binary and ternary Nb alloys in high-temperature water indicated the Mb-1Zr, Nb-5MO-1Zr, and Nb-5V-1Z4 (wt %) showed sufficient promise for further investigation. These alloys, together with pure Nb and Zircaloy-4 have been exposed to high purity water containing a low concentration of dissolved oxygen (<12 ppb) at 170, 230, and 300{degrees}C for up to {approx}3200 h. Weight-change data, microstructural observations, and qualitative mechanical-property evaluation reveal that Nb-5V-1Zr is the most promising alloy at higher temperatures. Below {approx}200{degrees}C, the alloys exhibit similiar corrosion behavior.

  19. Tritium permeation characterization of materials for fusion and generation IV very high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Thomson, S.; Pilatzke, K.; McCrimmon, K.; Castillo, I.; Suppiah, S. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, ON (Canada)

    2015-03-15

    The objective of this work is to establish the tritium-permeation properties of structural alloys considered for Fusion systems and very high temperature reactors (VHTR). A description of the work performed to set up an apparatus to measure permeation rates of hydrogen and tritium in 304L stainless steel is presented. Following successful commissioning with hydrogen, the test apparatus was commissioned with tritium. Commissioning tests with tritium suggest the need for a reduction step that is capable of removing the oxide layer from the test sample surfaces before accurate tritium-permeation data can be obtained. Work is also on-going to clearly establish the temperature profile of the sample to correctly estimate the tritium-permeability data.

  20. Magnet design with 100-kA HTS STARS conductors for the helical fusion reactor

    Science.gov (United States)

    Yanagi, N.; Terazaki, Y.; Ito, S.; Tamura, H.; Hamaguchi, S.; Mito, T.; Hashizume, H.; Sagara, A.

    2016-12-01

    The high-temperature superconducting (HTS) option is employed for the conceptual design of the LHD-type helical fusion reactor FFHR-d1. The 100-kA-class STARS (Stacked Tapes Assembled in Rigid Structure) conductor is used for the magnet system including the continuously wound helical coils. Protection of the magnet system in case of a quench is a crucial issue and the hot-spot temperature during an emergency discharge is estimated based on the zero-dimensional and one-dimensional analyses. The number of division of the coil winding package is examined to limit the voltage generation. For cooling the HTS magnet, helium gas flow is considered and its feasibility is examined by simple analysis as a first step.

  1. Fusion Reactor and Break-Even Experiment Based on Stabilized Liner Compression of Plasma

    Science.gov (United States)

    Turchi, Peter; Frese, Sherry; Frese, Michael

    2016-10-01

    An optimum regime, known as magnetized-target or magneto-inertial fusion (MTF/MIF), requires magnetic fields at megagauss levels, which are attainable by use of dynamic conductors called liners. The stabilized liner compressor (SLC) provides the basis for controlled implosion and re-capture of the liner for reversible energy exchange between liner kinetic energy and the internal energy of a magnetized-plasma target. This exchange requires rotational stabilization of Rayleigh-Taylor modes on the inner surface of the liner and pneumatically driven free-pistons that eliminate such modes at the outer surface. We discuss the implications of the SLC approach for the power reactor, a breakeven experiment, and intermediate experiments to develop the plasma target. Features include the importance of pneumatic drive and the liner-blanket for economic feasibility of MTF/MIF. Supported by ARPA-E ALPHA Program.

  2. Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1997-11-01

    A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  3. Polarised SANS study of microstructural evolution under neutron irradiation in a martensitic steel for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coppola, R.; Dewhurst, C.D.; Lindau, R.; May, R.P.; Moeslang, A.; Valli, M

    2004-03-01

    This work presents the results of polarised small-angle neutron scattering (SANS) measurements of modified martensitic steel DIN1.4914, originally developed for application in future fusion reactors (MANET steel). SANS measurements were made using the D22 instrument at the ILL Grenoble using an ad hoc polarised beam set-up. The investigated MANET samples were neutron irradiated and subsequently post-irradiation tempered to reproduce as much as possible the expected service conditions. The results, based on the analysis of the nuclear-magnetic interference, are discussed taking into account both the occurrence of Cr redistribution phenomena with correlated changes in the composition of the precipitate phases, and the growth of non-magnetic defects (He-bubbles or microvoids)

  4. The role of risk management in the design of diagnostics for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ingesson, L. C. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Collaboration: F4E Diagnostic Project Team

    2014-08-21

    A project-oriented approach is beneficial for the selection and design of viable diagnostics for fusion reactors because of the associated complex physical and organizational environment. The project-oriented approach includes rigorous risk management. The nature and impact of risks related to technical, organizational and commercial aspects in relation to the development of ITER diagnostics under EU responsibility are analyzed. The majority of risks are related to organizational aspects and technical feasibility issues. The experience with ITER is extrapolated to DEMO and beyond. It should not be taken for granted that technical solutions will be found, while a risk analysis of various diagnostic techniques with quantitative assessments undertaken early in the design of DEMO would be beneficial.

  5. Application of Kelvin Probe to Studies of Fusion Reactor Materials under Irradiation

    Institute of Scientific and Technical Information of China (English)

    Luo Guangnan; K. Yamaguchi; T. Terai; M. Yamawaki

    2005-01-01

    Recently, the work function (WF) changes in metallic and ceramic materials to be potentially used in future fusion reactors have been examined by means of Kelvin probe (KP),under He ion irradiation in high energy (MeV) and / or low energy (500 eV) ranges. The results of polycrystalline Ni samples indicate that the 1 MeV beam only induces decrease in the WF within the experimental fluence range; whereas the irradiation of 500 eV beam results in decrease in the WF firstly, then increase till saturation. A dual layer surface model is employed to explain the observed phenomena, together with computer simulation results by SRIM code. Charges buildup on the surface of lithium ceramics has been found to greatly influence the probe output, which can be explained qualitatively using a model concerning an induction electric field due to external field and free charges on the ceramic surface.

  6. Experimental determination of creep properties of Beryllium irradiated to relevant fusion power reactor doses

    Science.gov (United States)

    Scibetta, M.; Pellettieri, A.; Sannen, L.

    2007-08-01

    A dead weight machine has been developed to measure creep in irradiated beryllium relevant to fusion power reactors. Due to the external compressive load, the material will creep and the specimen will shrink. However, the specimen also swells due to the combined effect of internal pressure in helium bubbles and creep. One of the major challenges is to unmask swelling and derive intrinsic creep properties. This has been achieved through appropriate pre-annealing experiments. Creep has been measured on irradiated and unirradiated specimens. The temperature and stress dependence is characterized and modeled using the product of an Arrhenius' law for the temperature dependence and a power law for the stress dependence. Irradiation increases the sensitivity to creep but the irradiation effects can be rationalized by taking into account the irradiation-induced porosity. Experimental evidence supports dislocation climb by vacancy absorption to be the most plausible intrinsic creep mechanism.

  7. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  8. Comparison of the Recently proposed Super Marx Generator Approach to Thermonuclear Ignition with the DT Laser Fusion-Fission Hybrid Concept by the Lawrence Livermore National Laboratory

    CERN Document Server

    Winterberg, Friedwardt

    2009-01-01

    The recently proposed Super Marx generator pure deuterium micro-detonation ignition concept is compared to the Lawrence Livermore National Ignition Facility (NIF) Laser DT fusion-fission hybrid concept (LiFE) [1]. In a Super Marx generator a large number of ordinary Marx generators charge up a much larger second stage ultra-high voltage Marx generator, from which for the ignition of a pure deuterium micro-explosion an intense GeV ion beam can be extracted. A typical example of the LiFE concept is a fusion gain of 30, and a fission gain of 10, making up for a total gain of 300, with about 10 times more energy released into fission as compared to fusion. This means a substantial release of fission products, as in fusion-less pure fission reactors. In the Super Marx approach for the ignition of a pure deuterium micro-detonation a gain of the same magnitude can in theory be reached [2]. If feasible, the Super Marx generator deuterium ignition approach would make lasers obsolete as a means for the ignition of ther...

  9. Materials recycle and waste management in fusion power reactors. Progress report for 1982

    Energy Technology Data Exchange (ETDEWEB)

    Vogler, S.; Jung, J.; Steindler, M.J.; Maya, I.; Levine, H.E.; Peterman, D.D.; Strausburg, S.; Schultz, K.R.

    1983-01-01

    Several components of a STARFIRE fusion reactor have been studied. The breeding ratios were calculated as a function of lithium enrichment and neutron multiplier for systems containing either Li/sub 2/O or LiAlO/sub 2/. The lithium requirements for a fusion economy were also estimated for those cases and the current US resources were found to be adequate. However, competition with other lithium demands in the future emphasizes the need for recovering and reusing lithium. The radioactivities induced in the breeder and the impurities responsible for their formation were determined. The residual radioactivities of several low-activation structural materials were compared with the radioactivity from the prime candidate alloy (PCA) a titanium modified Type 316 stainless steel used in STARFIRE. The impurities responsible for the radioactivity levels were identified. From these radioactive impurity levels it was determined that V15Cr5Ti could meet the requirements for shallow land burial as specified by the Nuclear Regulatory Commission (10CFR61), whereas PCA would require a more restrictive disposal mode, i.e. in a geologic medium. The costs for each of these disposal modes were then estimated.

  10. Activation calculation and radiation analysis for China Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zhi, E-mail: zchen@ustc.edu.cn; Qiao, Shiji; Jiang, Shuai; Xu, X. George

    2016-11-01

    Highlights: • Activation calculation was performed using FLUKA for the main components of CFETR. • Radionuclides and radioactive wastes were assessed for CFETR. • The Waste Disposal Ratings (WDR) were assessed for CFETR. - Abstract: The activation calculation and analysis for the China Fusion Engineering Test Reactor (CFETR) will play an important role in its system design, maintenance, inspection and assessment of nuclear waste. Using the multi-particle transport code FLUKA and its associated data library, we calculated the radioactivity, specific activity, waste disposal rating from activation products, nuclides in the tritium breeding blanket, shielding layer, vacuum vessel and toroidal field coil (TFC) of CFETR. This paper presents the calculation results including neutron flux, activation products and waste disposal rating after one-year full operation of the CFETR. The findings show that, under the assumption of one-year operation at the 200 MW fusion power, the total radioactivity inventory will be 1.05 × 10{sup 19} Bq at shutdown and 1.03 × 10{sup 17} Bq after ten years. The primary residual nuclide is found to be {sup 55}Fe in ten years after the shutdown. The waste disposal rating (WDR) values are very low (<<1), according to Class C limits, CFETR materials are qualified for shallow land burial. It is shown that CFETR has no serious activation safety issue.

  11. Mitigation of hydrogen hazard in a fusion reactor by reduction of metallic oxides

    Energy Technology Data Exchange (ETDEWEB)

    Arnould, F.; Chaudron, V. [Technicatome, Dir. de l' Ingenierie, SEPS, 13 - Aix-en-Provence (France); Latge, C. [CEA/Cadarache, Dept. d' Etudes des Reacteurs (DER), 13 - Saint-Paul-lez-Durance (France); Laurent, A. [Institut National Polytechnique de Lorraine, Lab. des Sciences du Genie Chimique, 54 - Nancy (France)

    1998-07-01

    Significant quantities of hydrogen could be generated inside a fusion reactor during a loss of coolant accident (LOCA) by chemical reaction between steam released into the torus and plasma-facing components at high temperature. On the basis of functional specifications stemming from ITER accident sequence analyses, it appears that the implementation of an anaerobic hydrogen elimination system inside the torus or its expansion volume is the most advisable mitigation means because it allows removal of the hydrogen at its source of production before it comes into contact with oxygen. After a review of literature data and an experimental evaluation of various types of metallic oxides at laboratory scale, manganese oxide catalyzed by silver compounds was found to be the most efficient hydrogen getter. In order to test this hydrogen mitigation technique in representative fusion plant accident conditions, Technicatome and the CEA designed and built a pilot installation named MIRHABEL (MItigation of the Risk linked to Hydrogen by ABsorption and ELimination). After a short description of MIRHABEL, this paper presents and discusses the test results with respect to ITER accident conditions. (authors)

  12. Decay heat experiment and validation of calculation code systems for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

    1999-10-01

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

  13. Liquid immersion blanket design for use in a compact modular fusion reactor

    Science.gov (United States)

    Sorbom, Brandon; Ball, Justin; Barnard, Harold; Haakonsen, Christian; Hartwig, Zachary; Olynyk, Geoffrey; Sierchio, Jennifer; Whyte, Dennis

    2012-10-01

    Traditional tritium breeding blankets in fusion reactor designs include a large amount of structural material. This results in complex engineering requirements, complicated sector maintenance, and marginal tritium breeding ratios (TBR). We present a conceptual design of a fully liquid blanket. To maximize tritium breeding volume, the vacuum vessel is completely immersed in a continuously recycled FLiBe blanket, with the exception of small support posts. FLiBe has a wide liquid temperature window (459 C to 1430 C), low electrical conductivity to minimize MHD effects, similar thermal/fluid characteristics to water, and is chemically inert. While tritium breeding with FLiBe in traditional blankets is poor, we use MCNP neutronics analysis to show that the immersion blanket design coupled with a beryllium neutron multiplier results in TBR > 1. FLiBe is shown to be a sufficient radiation shield for the toroidal field magnets and can be used as a coolant for the vacuum vessel and divertor, allowing for a simplified single-phase, low-pressure, single-fluid cooling scheme. When coupled with a high-field compact reactor design, the immersion blanket eliminates the need for complex sector maintenance, allows the vacuum vessel to be a replaceable component, and reduces financial cost.

  14. Activation analysis and materials choice in the laser fusion reactor KOYO

    Science.gov (United States)

    Perlado, J. M.; Mima, K.; Nakai, S.; Alonso, E.; Mun˜oz, E.; Sanz, J.

    1996-10-01

    The laser fusion conceptual reactor KOYO, developed by the ILE Osaka, is presented and analyzed from the activation perspective. The reactor is driven by a laser diode pumped solid state laser which dramatically increases the efficiency of the system, and uses liquid LiPb film protection flowing through ceramic SiC porous tubes in the blanket. Neutron fluxes have been computed using 2/3D models and compared with spherical approaches. Two blanket areas with different packing fractions are considered, and we show the availability of a large fraction of the SiC with impurities to be considered as shallow land burial (SLB). We propose a more complete solution for SLB through the use of porous woven graphite (C) fabric tubes. A graphite reflector is included with important effect in the activation of the chamber wall. Ferritic HT-9 is considered as the structural material for the chamber wall, allowing its SLB and different recycling options. Releases of 1 kg of target-emissions-facing SiC tubes and HT-9 materials have also been simulated with optimum performances.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bell, M.G.; Batha, S.; Beer, M.; Bell, R.E.; Belov, A.; Berk, H.; Bernabei, S.; Bitter, M.; Breizman, B.; Bretz, N.L.; Budny, R.; Bush, C.E.; Callen, J.; Cauffman, S.; Chang, C.S.; Chang, Z.; Cheng, C.Z.; Darrow, D.S.; Dendy, R.O.; Dorland, W.; Duong, H.; Efthimion, P.C.; Ernst, D.; Evenson, H.; Fisch, N.J.; Fisher, R.; Fonck, R.J.; Fredrickson, E.D.; Fu, G.Y.; Furth, H.P.; Gorelenkov, N.N.; Goloborodko, V.Y.; Grek, B.; Grisham, L.R.; Hammett, G.W.; Hawryluk, R.J.; Heidbrink, W.; Herrmann, H.W.; Herrmann, M.C.; Hill, K.W.; Hogan, J.; Hooper, B.; Hosea, J.C.; Houlberg, W.A.; Hughes, M.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Kaita, R.; Kaye, S.; Kesner, J.; Kim, J.S.; Kissick, M.; Krasilnikov, A.V.; Kugel, H.; Kumar, A.; Lam, N.T.; Lamarche, P.; LeBlanc, B.; Levinton, F.M.; Ludescher, C.; Machuzak, J.; Majeski, R.P.; Manickam, J.; Mansfield, D.K.; Mauel, M.; Mazzucato, E.; McChesney, J.; McCune, D.C.; McKee, G.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mirnov, S.V.; Mueller, D.; Nagayama, Y.; Navratil, G.A.; Nazikian, R.; Okabayashi, M.; Osakabe, M.; Owens, D.K.; Park, H.K.; Park, W.; Paul, S.F.; Petrov, M.P.; Phillips, C.K.; Phillips, M.; Phillips, P.; Ramsey, A.T.; Rice, B.; Redi, M.H.; Rewoldt, G.; Reznik, S.; Roquemore, A.L.; Rogers, J.; Ruskov, E.; Sabbagh, S.A.; Sasao, M.; Schilling, G.; Schmidt, G.L.; Scott, S.D.; Semenov, I.; Senko, T.; Skinner, C.H.; Stevenson, T.; Strait, E.J.; Stratton, B.C.; Strachan, J.D.; Stodiek, W.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Thompson, M.E.; von Goeler, S.; Von Halle, A.; Walters, R.T.; Wang, S.; White, R.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.L.; Wurden, G.A.; Yamada, M.; Yavorski, V.; Young, K.M.; Zakharov, L.; Zarnstorff, M.C.; Zweben, S.J. [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States)

    1997-05-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas {bold 2}, 2176 (1995)] have explored several novel regimes of improved tokamak confinement in deuterium{endash}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{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through {ital 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{sub a}{approx}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{sub 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{sub 0}{gt}1 and weak magnetic shear in the central region, a toroidal Alfvn 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. {copyright} {ital 1997 American Institute of Physics.}

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

    Energy Technology Data Exchange (ETDEWEB)

    Bell, M.G.; Beer, M. [Princeton Univ., NJ (United States). Princeton Plasma Physics Lab.; Batha, S. [Fusion Physics and Technology, Torrance, CA (United States)] [and others

    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{sub 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{sub a} {approx} 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{sub 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{sub 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.

  17. Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Jaboulay, Jean-Charles, E-mail: Jean-Charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martin, Brunella, E-mail: brunella.martin@gmail.com [Incka, 19-21 Rue du 8 mai 1945, F-94110 Arcueil (France)

    2014-10-15

    SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.

  18. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    Science.gov (United States)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  19. Hydrogen Spectral Line Shape Formation in the SOL of Fusion Reactor Plasmas

    Directory of Open Access Journals (Sweden)

    Valery S. Lisitsa

    2014-05-01

    Full Text Available The problems related to the spectral line-shape formation in the scrape of layer (SOL in fusion reactor plasma for typical observation chords are considered. The SOL plasma is characterized by the relatively low electron density (1012–1013 cm−3 and high temperature (from 10 eV up to 1 keV. The main effects responsible for the line-shape formation in the SOL are Doppler and Zeeman effects. The main problem is a correct modeling of the neutral atom velocity distribution function (VDF. The VDF is determined by a number of atomic processes, namely: molecular dissociation, ionization and charge exchange of neutral atoms on plasma ions, electron excitation accompanied by the charge exchange from atomic excited states, and atom reflection from the wall. All the processes take place step by step during atom motion from the wall to the plasma core. In practice, the largest contribution to the neutral atom radiation emission comes from a thin layer near the wall with typical size 10–20 cm, which is small as compared with the minor radius of modern devices including international test experimental reactor ITER (radius 2 m. The important problem is a strongly non-uniform distribution of plasma parameters (electron and ion densities and temperatures. The distributions vary for different observation chords and ITER operation regimes. In the present report, most attention is paid to the problem of the VDF calculations. The most correct method for solving the problem is an application of the Monte Carlo method for atom motion near the wall. However, the method is sometimes too complicated to be combined with other numerical codes for plasma modeling for various regimes of fusion reactor operation. Thus, it is important to develop simpler methods for neutral atom VDF in space coordinates and velocities. The efficiency of such methods has to be tested via a comparison with the Monte Carlo codes for particular plasma conditions. Here a new simplified method

  20. Preliminary design concept of a subcritical reactor using available resources

    Energy Technology Data Exchange (ETDEWEB)

    Churnetski, E.L. [Oak Ridge Y-12 Plant, TN (United States); Hoyny, V.; Chaudhuri, B.R.; Taprantzis, A.; Yavas, A. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    1993-12-31

    During the Fall 1993 semester, a project was initiated within the Nuclear Engineering Department of the University of Tennessee with the objective of developing a design for a subcritical reactor with maximized multiplication factor using materials currently available. Such a device, if constructed, would serve as a teaching tool for the Department of Nuclear Engineering. Design work was conducted as a large number of computer calculations, with trial pile configurations based on fundamental nuclear engineering principles, in an effort to maximize multiplication factor through fuel element geometry, moderator type, fissile/moderator ratio, and reflector character. The principal objective of the design group for the early phase of this project was to present several possible ``baseline`` reactor designs and identify directions for improvements. For the sake of calculational ease, the cores analyzes to date have been of nearly cubic shape. The SCALE CSAS25 software which runs KENO.Va, a Monte Carlo code, was used for all neutronics calculations. The baseline reactors resulting from work to date are cuboidal in shape and graphite reflected. Two types of fuel element geometries are proposed, a typical triangular pitch rod lattice and an arrangement of discrete fuel slugs placed in a lattice corresponding to body centered cubic packing. The latter arrangement provides slightly higher multiplication factors than the former. Calculations were performed for both graphite and heavy water moderation with heavy water moderation producing considerably higher multiplication factors, as expected. In general, the maximum k{sub eff} for the reactors are in the range of 0.5 to 0.9, well subcritical, except in the cases of the extreme possible values of fuel assay where critical configurations are possible. In these cases, designs with reduced fuel loading are recommended to assure a subcritical multiplication factor.

  1. Materials technology for an advanced space power nuclear reactor concept: Program summary

    Science.gov (United States)

    Gluyas, R. E.; Watson, G. K.

    1975-01-01

    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  2. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  3. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  4. 球形托卡马克聚变嬗变堆中子学设计%NEUTRONICS DESIGN FOR A SPHERICAL TOKAMAK FUSION-TRANSMUTATION REACTOR

    Institute of Scientific and Technical Information of China (English)

    冯开明; 张国书; 郭增基

    2001-01-01

    Based on studies of spherical tokamak fusion reactors,a concept of fusion-transmutation reactor is put forward.A set of plasma parameters suitable for the transmutation blanket is selected.Using the transport and burn-up calculation code BISON3.0 and its associated database,transmutation rate of MA nuclear waste,energy multiplication,and tritium breeder rate in the transmutation blanket are calculated.%基于对球形托卡马克ST聚变堆的研究,提出了ST聚变嬗变堆的设计概念。对堆芯参数作了初步选择,确定了一组适合于嬗变包层的堆芯参数供中子学计算和结构设计参考,给出了旨在以嬗变次锕系元素(MA)核废物为目标的一维中子学计算结果。

  5. Analysis of Induced Gamma Activation by D-T Neutrons in Selected Fusion Reactor Relevant Materials with EAF-2010

    Directory of Open Access Journals (Sweden)

    Klix Axel

    2016-01-01

    Full Text Available Samples of lanthanum, erbium and titanium which are constituents of structural materials, insulating coatings and tritium breeder for blankets of fusion reactor designs have been irradiated in a fusion peak neutron field. The induced gamma activities were measured and the results were used to check calculations with the European activation system EASY-2010. Good agreement for the prediction of major contributors to the contact dose rate of the materials was found, but for minor contributors the calculation deviated up to 50%.

  6. Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited).

    Science.gov (United States)

    Smith, Roger J

    2008-10-01

    A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B(pol) diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T(e), n(e), and B(parallel) along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n(e)B(parallel) product and higher n(e) and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

  7. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1998-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  8. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1996-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  9. Evaluation of permeable and non-permeable tritium in normal condition in a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marta, V; Manuel, P J [Instituto de Fusion Nuclear (DENIM)/ETSII, Universidad Politecnica Madrid (UPM) (Spain); Sedano Luis, A [Ministerio de Educacion y Ciencia, Ciemat (Spain)], E-mail: marta@denim.upm.es

    2008-05-15

    The tritium cycle, technologies of process and control of the tritium in the plant will constitute a fraction of the environmental impact of the first generation of DT fusion reactors. The efforts of conceptual development of the tritium cycle are centered in the Internal Regenerator Cycle. The tritium could be recovered from a flow of He gas, or directly from solid breeder. The limits of transfers to the atmosphere are assumed {approx} 1 gr-T/a ({approx}20 Ci/a) (without species distinction). In the case of ITER, for example, we have global demands of control of 5 orders of magnitude have been demonstrated at experimental level. The transfer limits determine the key parameters in tritium Cycle (HT, HTO, as dominant, and T2, T2O as marginal). Presently, the transfer from the cycle to the environment is assumed through the exchange system of the power plant (primary to secondary). That transport is due to the permeation through HT, T2, or leakage to the coolant in the primary system. It is key the chemical optimization in the primary system, that needs to be reanalyzed in terms of radiological impact both for permeable, HT, T2, and non-permeable HTO, T2O. It is necessary considered the pathway of tritium from the reactor to the atmosphere, these processes are modelled adequately. Results of the assessments were early and chronic doses which have been evaluated for the Most Exposed Individual at particular distance bands from the release point. The impact evaluations will be performed with the computational tools (NORMTRI), besides national regulatory models, internationally accepted computer these code for dosimetric evaluations of tritiated effluents in operational conditions.

  10. Fusion Studies in Japan

    Science.gov (United States)

    Ogawa, Yuichi

    2016-05-01

    A new strategic energy plan decided by the Japanese Cabinet in 2014 strongly supports the steady promotion of nuclear fusion development activities, including the ITER project and the Broader Approach activities from the long-term viewpoint. Atomic Energy Commission (AEC) in Japan formulated the Third Phase Basic Program so as to promote an experimental fusion reactor project. In 2005 AEC has reviewed this Program, and discussed on selection and concentration among many projects of fusion reactor development. In addition to the promotion of ITER project, advanced tokamak research by JT-60SA, helical plasma experiment by LHD, FIREX project in laser fusion research and fusion engineering by IFMIF were highly prioritized. Although the basic concept is quite different between tokamak, helical and laser fusion researches, there exist a lot of common features such as plasma physics on 3-D magnetic geometry, high power heat load on plasma facing component and so on. Therefore, a synergetic scenario on fusion reactor development among various plasma confinement concepts would be important.

  11. Model Fusion Tool - the Open Environmental Modelling Platform Concept

    Science.gov (United States)

    Kessler, H.; Giles, J. R.

    2010-12-01

    The vision of an Open Environmental Modelling Platform - seamlessly linking geoscience data, concepts and models to aid decision making in times of environmental change. Governments and their executive agencies across the world are facing increasing pressure to make decisions about the management of resources in light of population growth and environmental change. In the UK for example, groundwater is becoming a scarce resource for large parts of its most densely populated areas. At the same time river and groundwater flooding resulting from high rainfall events are increasing in scale and frequency and sea level rise is threatening the defences of coastal cities. There is also a need for affordable housing, improved transport infrastructure and waste disposal as well as sources of renewable energy and sustainable food production. These challenges can only be resolved if solutions are based on sound scientific evidence. Although we have knowledge and understanding of many individual processes in the natural sciences it is clear that a single science discipline is unable to answer the questions and their inter-relationships. Modern science increasingly employs computer models to simulate the natural, economic and human system. Management and planning requires scenario modelling, forecasts and ‘predictions’. Although the outputs are often impressive in terms of apparent accuracy and visualisation, they are inherently not suited to simulate the response to feedbacks from other models of the earth system, such as the impact of human actions. Geological Survey Organisations (GSO) are increasingly employing advances in Information Technology to visualise and improve their understanding of geological systems. Instead of 2 dimensional paper maps and reports many GSOs now produce 3 dimensional geological framework models and groundwater flow models as their standard output. Additionally the British Geological Survey have developed standard routines to link geological

  12. Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Richou, Marianne; Magaud, Philippe; Missirlian, Marc [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, IT-00044 Frascati (Italy); Ridolfini, Vincenzo Pericoli [EFDA-CSU Garching, PPPT department, D-85748 Garching bei München (Germany)

    2013-10-15

    In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies. For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed. Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R and D activities are reported. This work has been carried out in the frame of EFDA PPPT Work Programme activities.

  13. A CONCEPT FOR NEXT STEP ADVANCED TOKAMAK FUSION DEVICE

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    A concept is introduced for initiating the design study of a special class of tokamak,which has a magnetic confinement configuration intermediate between contemporary advanced tokamak and the recently established spherical torus (ST,also well known by the name "spherical tokamak").The leading design parameter in the present proposal is a dimensionless geometrical parameter, the machine aspect ratio A=R0/a0=2.0,where the parameters a0 and R0 denote,respectively,the plasma (equatorial) minor radius and the plasma major radius.The aim of this choice is to technologically and experimentally go beyond the aspect ratio frontier (R0/a0≈2.5) of present day tokamaks and enter a broad unexplored domain existing on the (a0,R0) parameter space in current international tokamak database,between the data region already moderately well covered by the advanced conventional tokamaks and the data region planned to be covered by STs.Plasma minor radius a0 has been chosen to be the second basic design parameter, and consequently,the plasma major radius R0 is regarded as a dependent design parameter.In the present concept,a nominal plasma minor radius a0=1.2m is adopted to be the principal design value,and smaller values of a0 can be used for auxiliary design purposes,to establish extensive database linkage with existing tokamaks.Plasma minor radius can also be adjusted by mechanical and/or electromagnetic means to smaller values during experiments,for making suitable data linkages to existing machines with higher aspect ratios and smaller plasma minor radii.The basic design parameters proposed enable the adaptation of several confinement techniques recently developed by STs,and thereby a specially arranged central-bore region inside the envisioned tokamak torus,with retrieved space in the direction of plasma minor radius,will be available for technological adjustments and maneuverings to facilitate implementation of engineering instrumentation and real time high

  14. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy systems

    OpenAIRE

    Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This report contains two parts: (1) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (2) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showin...

  15. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Blink, J.A.

    1985-03-01

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs.

  16. Transmutation of Minor Actinides in a Spherical Torus Tokamak Fusion Reactor

    Institute of Scientific and Technical Information of China (English)

    FENGKaiming; ZHANGGuoshu

    2002-01-01

    Fusion energy will be a long-term energy source. Great efforts have been devoted to fusion research in the past 50 years, and there is still a long way to go. Transmutation of high-level waste (HLW) utilizing D-T fusion neutrons is a good choice for an early application of fusion.

  17. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    Science.gov (United States)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

  18. Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, Joel T.; Blandford, Edward D., E-mail: edb@unm.edu

    2016-07-15

    Highlights: • A novel directional heat exchanger design has been developed. • Hydrodynamic tests have been performed on the proposed design. • Heat transfer performance is inferred by hydrodynamic results. • Results are discussed and future work is suggested. - Abstract: The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

  19. A miniaturized test method for the mechanical characterization of structural materials for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gondi, P. [Rome-2 Univ. (Italy). Mech. Eng. Dept.; Donato, A. [ENEA CRE, Fusion Sector, Frascati, Rome (Italy); Montanari, R. [Rome-2 Univ. (Italy). Mech. Eng. Dept.; Sili, A. [Rome-2 Univ. (Italy). Mech. Eng. Dept.

    1996-10-01

    This work deals with a non-destructive method for mechanical tests which is based on the indentation of materials at a constant rate by means of a cylinder with a small radius and penetrating flat surface. The load versus penetration depth curves obtained using this method have shown correspondences with those of tensile tests and have given indications about the mechanical properties on a reduced scale. In this work penetration tests have been carried out on various kinds of Cr martensitic steels (MANET-2, BATMAN and modified F82H) which are of interest for first wall and structural applications in future fusion reactors. The load versus penetration depth curves have been examined with reference to data obtained in tensile tests and to microhardness measurements. Penetration tests have been performed at various temperature (from -180 to 100 C). Conclusions, which can be drawn for the ductile to brittle transition, are discussed for MANET-2 steel. Preliminary results obtained on BATMAN and modified F82H steels are reported. The characteristics of the indenter imprints have been studied by scanning electron microscopy. (orig.).

  20. Investigating the Neutral-Gas Manometers in the Wendelstein 7-X Experimental Fusion Reactor

    Science.gov (United States)

    Maisano-Brown, Jeannette; Wenzel, Uwe; Sunn-Pederson, Thomas

    2017-01-01

    The neutral-gas manometer is a powerful diagnostic tool used in the Wendelstein 7-X stellarator, a magnetized fusion experiment located in Germany. The Wendelstein, produced at a cost of 1.2 billion euros, and 20 years in the making, had its first experimental results in Winter 2016. Initial findings exceeded expectations but further study is still necessary. The particular instrument we examined was a hot-cathode ionization gauge, critical for attaining a quality in-vessel environment and a stable plasma. However, after the winter operation of Wendelstein, we found that some of the gauges had failed the six-second (maximum) plasma runs. Wendelstein is on track for 30-minute operations within three years, so it has become of utmost importance to scrutinize gauge design claims. We therefore subjected the devices to high magnetic field, input current, and temperature, as well as to long operational periods. Our results confirmed that the manometer cannot survive a 30-minute run. Though our findings did motivate promising recommendations for design improvement and for further experimentation so that the gauge can be ready for upcoming operations in Summer 2017 and eventual installment in ITER, the International Thermonuclear Experimental Reactor, currently under construction. This research was graciously supported by the Max Planck Institute and the MIT-Germany Initiative.

  1. Repetitive tabletop plasma focus to produce a tunable damage factor on materials for fusion reactors

    Science.gov (United States)

    Soto, Leopoldo; Pavez, Cristian; Inestrosa-Izurieta, Maria Jose; Moreno, Jose; Davis, Sergio; Bora, Biswajit; Avaria, Gonzalo; Jain, Jalaj; Altamirano, Luis; Panizo, Miguel; Gonzalez, Raquel; Rivera, Antonio

    2016-10-01

    Future thermonuclear reactors, both magnetic and inertial confinement approaches, need materials capable of withstanding the extreme radiation and heat loads expected from high repetition rate plasma. A damage factor (F = qτ1/2) in the order of 104 (W/cm2) s1/2 is expected. The axial plasma dynamics after the pinch in a tabletop plasma focus of hundred joules, PF-400J, was characterized by means of pulsed optical refractive diagnostics. The energy, interaction time and power flux of the plasma burst interacting with targets was obtained. Results show a high dependence of the damage factor with the distance from the anode top where the sample is located. A tunable damage factor in the range 10- 105(W/cm2) s1/2 can be obtained. At present the PF-400J operating at 0.077 Hz is being used to study the effects of fusion-relevant pulses on material target, including nanostructured materials. A new tabletop device to be operated up to 1Hz including tunable damage factor has been designed and is being constructed, thus thousand cumulative shots on materials could be obtained in few minutes. The scaling of the damage factor for plasma foci operating at different energies is discussed. Supported by CONICYT: PIA ACT-1115, PAI 79130026.

  2. Development of new generation reduced activation ferritic-martensitic steels for advanced fusion reactors

    Science.gov (United States)

    Tan, L.; Snead, L. L.; Katoh, Y.

    2016-09-01

    International development of reduced activation ferritic-martensitic (RAFM) steels has focused on 9 wt percentage Cr, which primarily contain M23C6 (M = Cr-rich) and small amounts of MX (M = Ta/V, X = C/N) precipitates, not adequate to maintain strength and creep resistance above ∼500 °C. To enable applications at higher temperatures for better thermal efficiency of fusion reactors, computational alloy thermodynamics coupled with strength modeling have been employed to explore a new generation RAFM steels. The new alloys are designed to significantly increase the amount of MX nanoprecipitates, which are manufacturable through standard and scalable industrial steelmaking methods. Preliminary experimental results of the developed new alloys demonstrated noticeably increased amount of MX, favoring significantly improved strength, creep resistance, and Charpy impact toughness as compared to current RAFM steels. The strength and creep resistance were comparable or approaching to the lower bound of, but impact toughness was noticeably superior to 9-20Cr oxide dispersion-strengthened ferritic alloys.

  3. Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C. [Forschungszentrum Juelich GmbH, EURATOM Association, Juelich (Germany)

    2010-07-01

    More and more attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO{sub 2}-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible meterials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PEMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. Materials for plasma facing components have to fulfill a number of requirements. First of all the materials have to be plasma compatible, i.e. they should exhibit a low atomic number to avoid radiative losses whenever atoms from the wall material will be ionized in the plasma. In addition, the materials must have a high melting point, a high thermal conductivity, and adequate mechanical properties. To select the most suitable material candidates, a comprehensive data base is required which includes all thermo-physical and mechanical properties. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm{sup -2}, meanwhile the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm{sup -2} for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs

  4. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  5. Report of the second joint Research Committee for Fusion Reactor and Materials. July 12, 2002, Tokyo, Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    Joint research committees in purpose of the discussion on DEMO blanket in view point of the both of reactor technology and materials were held by the Research Committee for Fusion Reactor and Fusion Materials. The joint research committee was held in Tokyo on July 12, 2002. In the committee, the present status of development of solid and liquid breeding blanket, the present status of development of reduced activation structure materials, and IFMIF (International Fusion Materials Irradiation Facility) program were discussed based on the discussions of the development programs of the blanket and materials at the first joint research committee. As a result, it was confirmed that high electric efficiency with 41% would be obtained in the solid breeding blanket system, that neutron radiation data of reduced activation ferritic steel was obtained by HFIR collaboration, and that KEP (key element technology phase) of IFMIF would be finished at the end of 2002 and the data base for the next step, i.e. EVEDA (engineering validation/engineering design activity) was obtained. In addition, the present status of ITER CTA, which was a transient phase for the construction, and the outline of ITER Fast Track, which was an accelerated plan for the performance of the power plants, were reported. This report consists of the summary of the discussion and the viewgraphs which were used at the second joint research committee, and these are very useful for the researchers of the fusion area in Japan. (author)

  6. A passively-safe fusion reactor blanket with helium coolant and steel structure

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, K.M.

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  7. Fusion

    CERN Document Server

    Mahaffey, James A

    2012-01-01

    As energy problems of the world grow, work toward fusion power continues at a greater pace than ever before. The topic of fusion is one that is often met with the most recognition and interest in the nuclear power arena. Written in clear and jargon-free prose, Fusion explores the big bang of creation to the blackout death of worn-out stars. A brief history of fusion research, beginning with the first tentative theories in the early 20th century, is also discussed, as well as the race for fusion power. This brand-new, full-color resource examines the various programs currently being funded or p

  8. Composition Optimization of Lithium-Based Ternary Alloy Blankets for Fusion Reactors

    Science.gov (United States)

    Jolodosky, Alejandra

    The goal of this dissertation is to examine the neutronic properties of a novel type of fusion reactor blanket material in the form of lithium-based ternary alloys. Pure liquid lithium, first proposed as a blanket for fusion reactors, is utilized as both a tritium breeder and a coolant. It has many attractive features such as high heat transfer and low corrosion properties, but most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns including degradation of the concrete containment structure. The work of this thesis began as a collaboration with Lawrence Livermore National Laboratory in an effort to develop a lithium-based ternary alloy that can maintain the beneficial properties of lithium while reducing the reactivity concerns. The first studies down-selected alloys based on the analysis and performance of both neutronic and activation characteristics. First, 3-D Monte Carlo calculations were performed to evaluate two main neutronics performance parameters for the blanket: tritium breeding ratio (TBR), and energy multiplication factor (EMF). Alloys with adequate results based on TBR and EMF calculations were considered for activation analysis. Activation simulations were executed with 50 years of irradiation and 300 years of cooling. It was discovered that bismuth is a poor choice due to achieving the highest decay heat, contact dose rates, and accident doses. In addition, it does not meet the waste disposal ratings (WDR). The straightforward approach to obtain Monte Carlo TBR and EMF results required 231 simulations per alloy and became computationally expensive, time consuming, and inefficient. Consequently, alternate methods were pursued. A collision history-based methodology recently developed for the Monte Carlo code Serpent, calculates perturbation effects on practically

  9. Concept of an inherently-safe high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro [Nuclear Hydrogen and Heat Application Research Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

    2012-06-06

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  10. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  11. A new concept of fusion neutron monitoring for PF-1000 device

    Directory of Open Access Journals (Sweden)

    Jednorog Slawomir

    2017-03-01

    Full Text Available The power output of plasma experiments and fusion reactors is a crucial parameter. It is determined by neutron yields that are proportional and directly related to the fusion yield. The number of emitted neutrons should be known for safety reasons and for neutron budget management. The PF-1000 is the large plasma facility based on the plasma focus phenomenon. PF-1000 is operating in the Institute of Plasma Physics and Laser Microfusion in Warsaw. Neutron yield changes during subsequent pulses, which is immanent part of this type device and so it must be monitored in terms of neutron emission. The reference diagnostic intended for this purpose is the silver activation counter (SAC used for many years. Our previous studies demonstrated the applicability of radio-yttrium for neutron yield measurements during the deuterium campaign on the PF-1000 facility. The obtained results were compared with data from silver activation counter and shown linear dependence but with some protuberances in local scale. Correlation between results for both neutron monitors was maintained. But the yttrium monitor registered the fast energy neutron that reached measurement apparatus directly from the plasma pinch. Based on the preliminary experiences, the yttrium monitor was designed to automatically register neutron-induced yttrium activity. The MCNP geometrical model of PF-1000 and yttrium monitor were both used for calculation of the activation coefficient for yttrium. The yttrium monitor has been established as the permanent diagnostic for monitoring fusion reactions in the PF-1000 device.

  12. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    Science.gov (United States)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  13. Reactor for boron fusion with picosecond ultrahigh power laser pulses and ultrahigh magnetic field trapping

    CERN Document Server

    Miley, G H; Kirchhoff, G

    2015-01-01

    Compared with the deuterium tritium (DT) fusion, the environmentally clean fusion of protons with 11B is extremely difficult. When instead of nanosecond laser pulses for thermal-ablating driven ignition, picosecond pulses are used, a drastic change by nonlinearity results in ultrahigh acceleration of plasma blocks. This radically changes to economic boron fusion by a measured new avalanche ignition.

  14. Condensation of ablated first-wall materials in the cascade inertial confinement fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ladd, A.J.C.

    1985-12-18

    This report concerns problems involved in recondensing first-wall materials vaporized by x rays and pellet debris in the Cascade inertial confinement fusion reactor. It examines three proposed first-wall materials, beryllium oxide (BeO), silicon carbide (SiO), and pyrolytic graphite (C), paying particular attention to the chemical equilibrium and kinetics of the vaporized gases. The major results of this study are as follows. Ceramic materials composed of diatomic molecules, such as BeO and SiC, exist as highly dissociated species after vaporization. The low gas density precludes significant recombination during times of interest (i.e., less than 0.1 s). The dissociated species (Be, O, Si, and C) are, except for carbon, quite volatile and are thermodynamically stable as a vapor under the high temperature and low density found in Cascade. These materials are thus unsuitable as first-wall materials. This difficulty is avoided with pyrolytic graphite. Since the condensation coefficient of monatomic carbon vapor (approx. 0.5) is greater than that of the polyatomic vapor (<0.1), recondensation is assisted by the expected high degree of dissociation. The proposed 10-layer granular carbon bed is sufficient to condense all the carbon vapor before it penetrates to the BeO layer below. The effective condensation coefficient of the porous bed is about 50% greater than that of a smooth wall. An estimate of the mass flux leaving the chamber results in a condensation time for a carbon first wall of about 30 to 50 ms. An experiment to investigate condensation in a Cascade-like chamber is proposed.

  15. Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Uebeyli, Mustafa E-mail: mubeyli@gazi.edu.tr

    2004-12-01

    Significant amounts of nuclear wastes consisting of plutonium, minor actinides and long lived fission products are produced during the operation of commercial nuclear power plants. Therefore, the destruction of these wastes is very important with respect to public health, environment and also the future of nuclear energy. In this study, transmutation of minor actinides (MAs) discharged from LMFBR spent fuel in a high power density fusion reactor has been investigated under a neutron wall load of 10 MW/m{sup 2} for an operation period of 10 years. Also, the effect of MA percentage on the transmutation has been examined. The fuel zone, containing MAs as spheres cladded with W-5Re, has been located behind the first wall to utilize the high neutron flux for transmutation effectively. Helium at 40 atm has been used as an energy carrier. At the end of the operation period, the total burning and transmutation are greater than the total buildups in all investigated cases, and very high burnups (420-470 GWd/tHM) are reached, depending on the MA content. The total transmutation rate values are 906 and 979 kg/GW{sub th} year at startup and decrease to 140 and 178 kg/GW{sub th} year at the end of the operation for fuel with 10% and 20% MA, respectively. Over an operation period of 10 years, the effective half lives decrease from 2.38, 2.21 and 3.08 years to 1.95, 1.80 and 2.59 years for {sup 237}Np, {sup 241}Am and {sup 243}Am, respectively. Total atomic densities decrease exponentially during the operation period. The reductions in the total atomic densities with respect to the initial ones are 79%, 81%, 82%, 83%, 85% and 86% for 10%, 12%, 14%, 16%, 18% and 20% MAs, respectively.

  16. Thermal hydraulic characteristics during ingress of coolant and loss of vacuum events in fusion reactors

    Science.gov (United States)

    Takase, K.; Kunugi, T.; Seki, Y.; Akimoto, H.

    2000-03-01

    The thermal hydraulic characteristics in the vacuum vessel (VV) of a fusion reactor under an ingress of coolant event (ICE) and a loss of vacuum event (LOVA) were investigated quantitatively using preliminary experimental apparatuses. In the ICE experiments, pressure rise characteristics in the VV were clarified for experimental parameters of the wall temperature and water temperature and for cases with and without a blowdown tank. In addition, the functional performance of a blowdown tank with and without a water cooling system was examined and it was confirmed that the blowdown tank with a water cooling system is effective for suppressing the pressure rise during the ICE. In the LOVA experiments, the saturation time in the VV from vacuum to atmosphere was investigated for various breach sizes and it was found that the saturation time is in inverse proportion to the breach size. In addition, the characteristics of exchange flow through breaches were clarified for the different breach positions on the VV. It was proven from the experimental results that the exchange flow became a counter-current flow when the breach was positioned on the top of the VV and a stratified flow when it was formed on the side wall of the VV, and that the exchange flow under the stratified flow condition was smoother than that of counter-current flow. On the basis of these results, the severest breach condition in ITER was changed from the top-break case to the side-break case. To predict with high accuracy the thermal hydraulic characteristics during ICEs and LOVAs under ITER conditions, a large scale test facility will be necessary. The current conceptual design of the combined ICE-LOVA test facility with a scaling factor of 1/1000 in comparison with the ITER volume is presented.

  17. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 92--94). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. [eds.

    1994-11-01

    This is the Final Report for a three-year (FY 92--94) study of the Environmental, Safety, and Economic (ESE) aspects of fusion energy systems, emphasizing development of computerized approaches suitable for incorporation as modules in fusion system design codes. First, as is reported in Section 2, the authors now have operating a simplified but complete environment and safety evaluation code, BESAFE. The first tests of BESAFE as a module of the SUPERCODE, a design optimization systems code at LLNL, are reported in Section 3. Secondly, as reported in Section 4, the authors have maintained a strong effort in developing fast calculational schemes for activation inventory evaluation. In addition to these major accomplishments, considerable progress has been made on research on specific topics as follows. A tritium modeling code TRIDYN was developed in collaboration with the TSTA group at LANL and the Fusion Nuclear Technology group at UCLA. A simplified algorithm has been derived to calculate the transient temperature profiles in the blanket during accidents. The scheme solves iteratively a system of non-linear ordinary differential equations describing about 10 regions of the blanket by preserving energy balance. The authors have studied the physics and engineering aspects of divertor modeling for safety applications. Several modifications in the automation and characterization of environmental and safety indices have been made. They have applied this work to the environmental and safety comparisons of stainless steel with alternative structural materials for fusion reactors. A methodology in decision analysis utilizing influence and decision diagrams has been developed to model fusion reactor design problems. Most of the work during this funding period has been reported in 26 publications including theses, journal publications, conference papers, and technical reports, as listed in Section 11.

  18. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

  19. An autonomous long-term fast reactor system and the principal design limitations of the concept

    Science.gov (United States)

    Tsvetkova, Galina Valeryevna

    The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National

  20. A HUMAN AUTOMATION INTERACTION CONCEPT FOR A SMALL MODULAR REACTOR CONTROL ROOM

    Energy Technology Data Exchange (ETDEWEB)

    Le Blanc, Katya; Spielman, Zach; Hill, Rachael

    2017-06-01

    Many advanced nuclear power plant (NPP) designs incorporate higher degrees of automation than the existing fleet of NPPs. Automation is being introduced or proposed in NPPs through a wide variety of systems and technologies, such as advanced displays, computer-based procedures, advanced alarm systems, and computerized operator support systems. Additionally, many new reactor concepts, both full scale and small modular reactors, are proposing increased automation and reduced staffing as part of their concept of operations. However, research consistently finds that there is a fundamental tradeoff between system performance with increased automation and reduced human performance. There is a need to address the question of how to achieve high performance and efficiency of high levels of automation without degrading human performance. One example of a new NPP concept that will utilize greater degrees of automation is the SMR concept from NuScale Power. The NuScale Power design requires 12 modular units to be operated in one single control room, which leads to a need for higher degrees of automation in the control room. Idaho National Laboratory (INL) researchers and NuScale Power human factors and operations staff are working on a collaborative project to address the human performance challenges of increased automation and to determine the principles that lead to optimal performance in highly automated systems. This paper will describe this concept in detail and will describe an experimental test of the concept. The benefits and challenges of the approach will be discussed.

  1. Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bourham, Mohamed A.; Gilligan, John G.

    1999-08-14

    Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

  2. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  3. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  4. A breed and burn reshuffling scheme for an Astrid-like reactor concept design

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois L, J. L., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2016-09-15

    The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

  5. Automated disposable small scale reactor for high throughput bioprocess development: a proof of concept study.

    Science.gov (United States)

    Bareither, Rachel; Bargh, Neil; Oakeshott, Robert; Watts, Kathryn; Pollard, David

    2013-12-01

    The acceleration of bioprocess development for biologics and vaccines can be enabled by automated high throughput technologies. This will alleviate the significant resource burden from the multi-factorial statistical experimentation required for controlling product quality attributes of complex biologics. Recent technology advances have improved clone evaluation and screening, but have struggled to combine the scale down criteria required for both high cell density cell culture and microbial processes, with sufficient automation and disposable technologies to accelerate process development. This article describes the proof of concept evaluations of an automated disposable small scale reactor for high throughput upstream process development. Characterization studies established the small scale stirred tank disposable 250 mL reactor as similar to those of lab and pilot scale. The reactor generated equivalent process performance for industrial biologics processes for therapeutic protein and monoclonal antibody production using CHO cell culture, Pichia pastoris and E. coli. This included similar growth, cell viability, product titer, and product quality. The technology was shown to be robust across multiple runs and met the requirements for the ability to run high cell density processes (>400 g/L wet cell weight) with exponential feeds and sophisticated event triggered processes. Combining this reactor into an automated array of reactors will ultimately be part of a high throughput process development strategy. This will combine upstream, small scale purification with rapid analytics that will dramatically shorten timelines and costs of developing biological processes.

  6. Two generic concepts for space propulsion based on thermal nuclear fusion

    Science.gov (United States)

    Gabrielli, R. A.; Petkow, D.; Herdrich, G.; Laufer, R.; Röser, H.-P.

    2014-08-01

    In the present work, two different concepts for fusion based space propulsion are compared. While the first concept is based solely on propulsion by hypothetic ejection of fusion products and hence may be called ash drive, the second one uses an additional coolant for thrust enhancement. Since this coolant was initially assumed to be gaseous and since it is doing most of the propulsion work, the name of “working gas drive” has been proposed. Propulsive characteristics for both types are evaluated for four fusion reactant couples (D-T; D-3He; 3He-3He; 11B-p). In working gas drives, only hydrogen is considered as coolant due to its exceptionally good caloric and propulsive properties. The results of comparative studies show that while ash drives excel working gas drives in terms of specific impulse the latter yield considerably more thrust than ash drives. Another major drawback of the ash drives is relatively small thrust efficiencies. The plasma power has to be disposed of nearly entirely as waste heat leading to prohibitive radiator masses.

  7. Reactor concepts for atomic layer deposition on agitated particles: A review

    Energy Technology Data Exchange (ETDEWEB)

    Longrie, Delphine, E-mail: delphine.longrie@asm.com; Deduytsche, Davy; Detavernier, Christophe, E-mail: christophe.detavernier@ugent.be [Department of Solid State Sciences, Ghent University, Krijgslaan 281/S1, B-9000 Gent (Belgium)

    2014-01-15

    The number of possible applications for nanoparticles has strongly increased in the last decade. For many applications, nanoparticles with different surface and bulk properties are necessary. A popular surface modification technique is coating the particle surface with a nanometer thick layer. Atomic layer deposition (ALD) is known as a reliable method for depositing ultrathin and conformal coatings. In this article, agitation or fluidization of the particles is necessary for performing ALD on (nano)particles. The principles of gas fluidization of particles will be outlined, and a classification of the gas fluidization behavior of particles based on their size and density will be given. Following different reactor concepts that have been designed to conformally coat (nano)particles with ALD will be described, and a concise overview will be presented of the work that has been performed with each of them ending with a concept reactor for performing spatial ALD on fluidized particles.

  8. Characterization of scintillator materials for fast-ion loss detectors in nuclear fusion reactors

    Science.gov (United States)

    Jiménez-Ramos, M. C.; García López, J.; García-Muñoz, M.; Rodríguez-Ramos, M.; Carmona Gázquez, M.; Zurro, B.

    2014-08-01

    In fusion plasma reactors, fast ion generated by heating systems and fusion born particles must be well confined. The presence of magnetohydrodynamic (MHD) instabilities can lead to a significant loss of these ions, which may reduce drastically the heating efficiency and may cause damage to plasma facing components in the vacuum vessel. In order to understand the physics underlying the fast ion loss mechanism, scintillator based detectors have been installed in several fusion devices. In this work we present the absolute photon yield and its degradation with ion fluence in terms of the number of photons emitted per incident ion of several scintillators thin coatings: SrGa2S4:Eu2+ (TG-Green), Y3Al5O12:Ce3+ (P46) and Y2O3:Eu3+ (P56) when irradiated with light ions of different masses (deuterium ions, protons and α-particles) at energies between approximately 575 keV and 3 MeV. The photon yield will be discussed in terms of the energy deposited by the particles into the scintillator. For that, the actual composition and thickness of the thin layers were determined by Rutherford Backscattering Spectrometry (RBS). A collimator with 1 mm of diameter, which defines the beam size for the experiments, placed at the entrance of the chamber. An electrically isolated sample holder biased to +300 V to collect the secondary electrons, connected to a digital current integrator (model 439 by Ortec) to measure the incident beam current. A home made device has been used to store the real-time evolution of the beam current in a computer file allowing the correction of the IL yields due to the current fluctuations. The target holder is a rectangle of 150 × 112 mm2 and can be tilted. The X and Y movements are controlled through stepping motors, which permits a fine control of the beam spot positioning as well as the study of several samples without venting the chamber. A silica optical fiber of 1 mm diameter fixed to the vacuum chamber, which collects the light from the scintillators

  9. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  10. Automation of nonlinear calculations in the theory of fusion reactor; Automatisation des calculs non lineaires dans la theorie des reacteurs a fusion

    Energy Technology Data Exchange (ETDEWEB)

    Braffort, P.; Chaigne, M. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    1) Introduction: The difficulties of the formulation of the equations of phenomena occurring during the operation of a fusion reactor are underlined. 2) The possibilities presented by analog computation of the solution of nonlinear differential equations are enumerated. The accuracy and limitations of this method are discussed. 3) The analog solution in the stationary problem of the measurement of the discharge confinement is given and comparison with experimental results. 4) The analog solution of the dynamic problem of the evolution of the discharge current in a simple case is given and it is compared with experimental data. 5) The analog solution of the motion of an isolated ion in the electromagnetic field is given. A spatial field simulator used for this problem (bidimensional problem) is described. 6) The analog solution of the preceding problem for a tridimensional case for particular geometrical configurations using simultaneously 2 field simulators is given. 7) A method of computation derived from Monte Carlo method for the study of dynamic of plasma is described. 8) Conclusion: the essential differences between the analog computation of fission reactors and fusion reactors are analysed. In particular the theory of control of a fusion reactor as described by SCHULTZ is discussed and the results of linearized formulations are compared with those of nonlinear simulation. (author)Fren. [French] 1) Introduction. On souligne les difficultes que presente la mise en equation des phenomenes mis en jeu lors du fonctionnement d'un reacteur a fusion. On selectionne un certain nombre d'equations generalement utilisees et on montre les impossibilites analytiques auxquelles on se heurte alors. 2) On rappelle les possibilites du calcul analogique pour la resolution des systemes differentiels non lineaires et on indique la precision de la methode ainsi que ses limitations. 3) On decrit esolution analogique du probleme statique de la mesure du confinement de la

  11. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  12. Classical physics impossibility of magnetic fusion reactor with neutral beam injection at thermonuclear energies below 200 KeV

    Science.gov (United States)

    Maglich, Bogdan; Hester, Timothy; Vaucher, Alexander

    2016-10-01

    Lawson criterion was specifically derived for inertial fusion and DT gas of stable lifetime without ions and magnetic fields. It was revised with realistic parametrers. To account for the losses of unstable ions against neutralization with lifetime τ, n (t) = nτ [ 1 - exp (- t / - tτ τ) ] -> nτ for τ CT resonance regime below critical energy To, τ 10-5 , and Lawson requirement ntL 1021 i.e. not realistic. Luminosity (reaction rate for σ = 1) is that of two unstable particles each with lifetime τ: L =n2(t)v12 =n2t2v12 . In subcritical regime, L =10-10n2 forn =1014cm-3 , v 109 cms-1 = L =1027 . Which is negligible and implies a negative power flow reactor. But above T0 , atTD = 725 KeV , τ = 20 s was observed implying L =1039 i.e. massive fusion energy production.

  13. The Pulsed Fission-Fusion (PUFF) Concept for Deep Space Exploration and Terrestrial Power Generation

    Science.gov (United States)

    Adams, Robert; Cassibry, Jason; Schillo, Kevin

    2017-01-01

    This team is exploring a modified Z-pinch geometry as a propulsion system, imploding a liner of liquid lithium onto a pellet containing both fission and fusion fuel. The plasma resulting from the fission and fusion burn expands against a magnetic nozzle, for propulsion, or a magnetic confinement system, for terrestrial power generation. There is considerable synergy in the concept; the lithium acts as a temporary virtual cathode, and adds reaction mass for propulsion. Further, the lithium acts as a radiation shield against generated neutrons and gamma rays. Finally, the density profile of the column can be tailored using the lithium sheath. Recent theoretical and experimental developments (e.g. tailored density profile in the fuel injection, shear stabilization, and magnetic shear stabilization) have had great success in mitigating instabilities that have plagued previous fusion efforts. This paper will review the work in evaluating the pellet sizes and z-pinch conditions for optimal PuFF propulsion. Trades of pellet size and composition with z-pinch power levels and conditions for the tamper and lithium implosion are evaluated. Current models, both theoretical and computational, show that a z-pinch can ignite a small (1 cm radius) fission-fusion target with significant yield. Comparison is made between pure fission and boosted fission targets. Performance is shown for crewed spacecraft for high speed Mars round trip missions and near interstellar robotic missions. The PuFF concept also offers a solution for terrestrial power production. PuFF can, with recycling of the effluent, achieve near 100% burnup of fission fuel, providing a very attractive power source with minimal waste. The small size of PuFF relative to today's plants enables a more distributed power network and less exposure to natural or man-made disruptions.

  14. Accuracy and convergence of coupled finite-volume/Monte Carlo codes for plasma edge simulations of nuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ghoos, K., E-mail: kristel.ghoos@kuleuven.be [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium); Dekeyser, W. [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium); Samaey, G. [KU Leuven, Department of Computer Science, Celestijnenlaan 200A, 3001 Leuven (Belgium); Börner, P. [Institute of Energy and Climate Research (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); Baelmans, M. [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium)

    2016-10-01

    The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracy by making use of averaging in the Random Noise coupling technique.

  15. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States); Tsai, H.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  16. Progress in studies of Li/sub 17/Pb/sub 83/ as liquid breeder for fusion reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.

    1983-09-01

    A review of the experimental and conceptual design work in progress at JRC-Ispra to investigate the feasibility of the eutectic Li/sub 17/Pb/sub 83/ as a liquid breeder for experimental power reactors is presented. Results of recent measurements to implement the data base of this material are given in the following areas: physical parameters, hydrogen solubility and recovery, chemical reactivity with air and water, compatibility with steel. The studies carried out on blanket concepts for the INTOR (International Tokamak Reactor)/NET (Next European Torus) projects are outlined and discussed.

  17. Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ryskamp, John M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. This work can be viewed as an update to a project from the 1990’s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the “Cylindrical” variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a

  18. Chemical thermodynamics of fusion reactor breeding materials and their interaction with tritium

    Energy Technology Data Exchange (ETDEWEB)

    Ihle, H.R.; Wu, C.H. (Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.))

    1985-02-01

    Liquid lithium, lithium alloys (solid and liquid) and ceramic lithium compounds are candidate breeding materials for (D, T) fusion reactors. Besides their tritium breeding capability, which results from neutron capture, their thermochemical properties and their interaction with tritium are of particular interest. A good knowledge of the physical and chemical properties of liquid lithium exists; and the systems Li-LiH, Li-LiD and Li-LiT have been studied in great detail. For dilute solutions of D/sub 2/ in liquid lithium, Sieverts' law was found to be valid down to an atom fraction of xsub(D)=10/sup -6/; in the vapor, lithium polymers up to Li/sub 4/ and lithium deuterides are found. In the system liquid Li-Pb, the solubility of D/sub 2/ was measured as a function of temperature and alloy composition, and correlated with the activities of the constituent metals. The solubility of D/sub 2/ was found to obey Sieverts' law at low concentrations, and is many orders of magnitude smaller than that in liquid lithium. This holds also for solid 'Li/sub 7/ Pb/sub 2/'. Vaporization studies yielded data on the thermal stability of the oxides: Li/sub 2/O, ..gamma..-LiAlO/sub 2/, ..beta..-Li/sub 5/AlO/sub 4/, LiAl/sub 5/O/sub 8/, Li/sub 2/ZrO/sub 3/, Li/sub 4/ZrO/sub 4/, Li/sub 8/ZrO/sub 6/, Li/sub 2/SiO/sub 3/ and Li/sub 4/SiO/sub 4/. Tritium diffusivity was studied in Li/sub 2/O, ..gamma..-LiAlO/sub 2/, ..beta..-Li/sub 5/AlO/sub 4/ and Li/sub 4/SiO/sub 4/. A large number of gaseous lithides were detected during these studies.

  19. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    Science.gov (United States)

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  20. Prospects for a dominantly microwave-diagnosed magnetically confined fusion reactor

    Science.gov (United States)

    Volpe, F. A.

    2017-01-01

    Compared to present experiments, tokamak and stellarator reactors will be subject to higher heat loads, sputtering, erosion and subsequent coating, tritium retention, higher neutron fluxes, and a number of radiation effects. Additionally, neutral beam penetration in tokamak reactors will only be limited to the plasma edge. As a result, several optical, beam-based and magnetic diagnostics of today's plasmas might not be applicable to tomorrow's reactors, but the present discussion suggests that reactors could largely rely on microwave diagnostics, including techniques based on mode conversions and Collective Thomson Scattering.

  1. Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

    Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

  2. What we miss in order to be able to design and build a commercially viable fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Andreani, R. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dipt. Energia

    1999-07-01

    The paper considers in a critical way the different areas in which work is required to provide sufficient information in view of designing a reliable and attractive fusion reactor. Four main areas of activity are considered: physics, technology, engineering, safety. In physics the trend is positive towards a better understanding of suitable plasma regimes to be confirmed through further experimentation on the operating machines. Engineering has already proven itself by the design and construction of a number of experimental machines. In addition a large data base obtained from design and operation of fission reactors is available. Safety is reaching very satisfactory results in the analysis of the impact of fusion on man and the environment. Where it is still a large unsolved problem is concerning materials capable of standing the harsh fusion environment for an adequate number of years. An intense neutron source is needed in order to allow the necessary developments. [Italian] Il rapporto considera in modo critico le differenti aree nelle quali si richiede ulteriore lavoro per fornire informazini sufficienti al fine di progettare un reattore a fusione affidabile ed economicamente competitivo. Vengono considerate quattro aree principali di attivita': fisica, tecnologia, ingegneria, sicurezza. Nella fisica, vi e' una positiva tendenza verso una migliore comprensione di regimi di plasma favorevoli da confermare attraverso ulteriore sperimentazione sulle macchine funzionanti. L' ingegneria ha gia' dato dimostrazione di se' col progetto e la costruzione di un notevole numero di macchine sperimentali. In aggiunta e' disponibile un gran numero di dati ottenuti dalla progettazione, realizzazione e funzionamento dei reattori a fissione. La sicurezza sta raggiungendo risultati molto soddisfacenti nell'analisi dell'impatto della fusione sull'uomo e sull'ambiente. Un grosso problema tuttora irresoluto e' quello dei

  3. Plasma-Facing Materials Research For Fusion Reactors At FOM Rijnhuizen

    NARCIS (Netherlands)

    Rapp, J.; De Temmerman, G.; van Rooij, G. J.; van Emmichoven, P. A. Zeijlma; Kleyn, A. W.

    2011-01-01

    In next generation magnetic fusion devices such as ITER, plasma-facing materials are exposed to unprecedented high ion, power and neutron fluxes. Those extreme conditions cannot be recreated in current fusion devices from the tokamak type. The plasma-surface interaction is still an area of great unc

  4. Plasma-facing materials research for fusion reactors at Fom Rijnhuizen

    NARCIS (Netherlands)

    Rapp, J.; De Temmerman, G.; van Rooij, G.J.; Zeijlmans van Emmichoven, P.A.; Kleijn, A.W.

    2011-01-01

    In next generation magnetic fusion devices such as ITER, plasma-facing materials are exposed to unprecedented high ion, power and neutron fluxes. Those extreme conditions cannot be recreated in current fusion devices from the tokamak type. The plasma-surface interaction is still an area of great unc

  5. Plasma-Facing Materials Research For Fusion Reactors At FOM Rijnhuizen

    NARCIS (Netherlands)

    Rapp, J.; De Temmerman, G.; van Rooij, G. J.; van Emmichoven, P. A. Zeijlma; Kleyn, A. W.

    2011-01-01

    In next generation magnetic fusion devices such as ITER, plasma-facing materials are exposed to unprecedented high ion, power and neutron fluxes. Those extreme conditions cannot be recreated in current fusion devices from the tokamak type. The plasma-surface interaction is still an area of great unc

  6. Plasma-facing materials research for fusion reactors at Fom Rijnhuizen

    NARCIS (Netherlands)

    Rapp, J.; De Temmerman, G.; van Rooij, G.J.; Zeijlmans van Emmichoven, P.A.; Kleijn, A.W.

    2011-01-01

    In next generation magnetic fusion devices such as ITER, plasma-facing materials are exposed to unprecedented high ion, power and neutron fluxes. Those extreme conditions cannot be recreated in current fusion devices from the tokamak type. The plasma-surface interaction is still an area of great unc

  7. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Science.gov (United States)

    Anuar, Nuraslinda; Muhamad Pauzi, Anas

    2016-01-01

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the βmin is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the βmin, resulting in a list of candidate designs that possess the β value that is larger than the βmin. The proposed methodology can also be applied to purposes other than technological foresight.

  8. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Energy Technology Data Exchange (ETDEWEB)

    Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my; Muhamad Pauzi, Anas, E-mail: anas@uniten.edu.my [College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β{sub min} is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β{sub min}, resulting in a list of candidate designs that possess the β value that is larger than the β{sub min}. The proposed methodology can also be applied to purposes other than technological foresight.

  9. Minimization of the external heating power by long fusion power rise-up time for self-ignition access in the helical reactor FFHR2m

    Science.gov (United States)

    Mitarai, O.; Sagara, A.; Chikaraishi, H.; Imagawa, S.; Watanabe, K.; Shishkin, A. A.; Motojima, O.

    2007-11-01

    Minimization of the external heating power to access self-ignition is advantageous to increase the reactor design flexibility and to reduce the capital and operating costs of the plasma heating device in a helical reactor. In this work we have discovered that a larger density limit leads to a smaller value of the required confinement enhancement factor, a lower density limit margin reduces the external heating power and over 300 s of the fusion power rise-up time makes it possible to reach a minimized heating power. While the fusion power rise-up time in a tokamak is limited by the OH transformer flux or the current drive capability, any fusion power rise-up time can be employed in a helical reactor for reducing the thermal stresses of the blanket and shields, because the confinement field is generated by the external helical coils.

  10. In-reactor tests of the nuclear light bulb rocket concept

    Science.gov (United States)

    Gauntt, R. O.; Slutz, S. A.; Latham, T. S.; Roman, W. C.; Rogers, R. J.

    1992-07-01

    An overview is given of the closed-cycle Gas Core Nuclear Rocket outlining scenarios for its use in short-duration Mars missions and results of Nuclear Light Bulb (NLB) tests. Isothermal and nonnuclear tests are described which confirmed the fundamental concepts behind the NLB. NLB reference-engine performance characteristics are given for hypothetical engines that could be used for manned Mars missions. Vehicle/propulsion sizing is based on a Mars mission with three trans-Mars impulse burns, capture and escape burns, and a total mission duration of 600 days. The engine would have a specific impulse of 1870 seconds, a 412-kN thrust, and a thrust/weight ratio of 1.3. Reactor tests including small-scale in-reactor tests are shown to be prerequisites for studying: (1) fluid mechanical confinement of the gaseous nuclear fuel; (2) buffer gas separation and circulation; and (3) the minimization of transparent wall-heat loading. The reactor tests are shown to be critical for establishing the feasibility of the NLB concept.

  11. A Review of Dangerous Dust in Fusion Reactors: from Its Creation to Its Resuspension in Case of LOCA and LOVA

    Directory of Open Access Journals (Sweden)

    Andrea Malizia

    2016-07-01

    Full Text Available The choice of materials for the future nuclear fusion reactors is a crucial issue. In the fusion reactors, the combination of very high temperatures, high radiation levels, intense production of transmuting elements and high thermomechanical loads requires very high-performance materials. Erosion of PFCs (Plasma Facing Components determines their lifetime and generates a source of impurities (i.e., in-vessel tritium and dust inventories, which cool down and dilute the plasma. The resuspension of dust could be a consequences of LOss of Coolant Accidents (LOCA and LOss of Vacuum Accidents (LOVA and it can be dangerous because of dust radioactivity, toxicity, and capable of causing an explosion. These characteristics can jeopardize the plant safety and pose a serious threat to the operators. The purpose of this work is to determine the experimental and numerical steeps to develop a numerical model to predict the dust resuspension consequences in case of accidents through a comparison between the experimental results taken from campaigns carried out with STARDUST-U and the numerical simulation developed with CFD codes. The authors in this work will analyze the candidate materials for the future nuclear plants and the consequences of the resuspension of its dust in case of accidents through the experience with STARDUST-U.

  12. Research and development of the tritium recovery system for the blanket of the fusion reactor in JAEA

    Science.gov (United States)

    Kawamura, Y.; Isobe, K.; Iwai, Y.; Kobayashi, K.; Nakamura, H.; Hayashi, T.; Yamanishi, T.

    2009-05-01

    A water-cooling solid breeder blanket is a prime candidate for the blanket of the fusion reactor in Japan. In this case, the blanket tritium recovery system will be composed of three processes: tritium recovery from helium sweep gas as hydrogen, that as water vapour and tritium recovery from coolant water. The authors have proposed a set of advanced systems. For tritium recovery as hydrogen, an electrochemical hydrogen pump with a ceramic proton conductor has been proposed. The correlation between the proton concentration in the ceramic and the hydrogen gas pressure has been investigated to describe the pumping performance specifically. A ceramic electrolysis cell has been proposed to process the tritiated water vapour. The authors have developed a new electrode containing cerium oxide, and it has shown fairly good electrolysis efficiency. For tritium recovery from coolant water, reduction in the processing water by tritium concentration is necessary. The authors have proposed to apply the fixed-bed adsorption process of synthetic zeolite, and have developed new zeolite. It showed unique characteristics for water adsorption and desorption. The authors have determined the potential of these systems for the blanket of the fusion DEMO reactor.

  13. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li, Min; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Lv, Zhongliang; Ye, Minyou

    2015-02-15

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%.

  14. Irradiation creep in austenitic and ferritic steels irradiated in a tailored neutron spectrum to induce fusion reactor levels of helium

    Energy Technology Data Exchange (ETDEWEB)

    Grossbeck, M.L.; Gibson, L.T. [Oak Ridge National Laboratory, TN (United States); Jitsukawa, S.

    1996-04-01

    Six austenitic stainless steels and two ferritic alloys were irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor where an atomic displacement level of 7.4 dpa was achieved and was then transferred to the High Flux Isotope Reactor for the remainder of the irradiation to a total displacement level of 19 dpa. Temperatures of 60 and 330{degree}C are reported on. At 330{degree}C irradiation creep was found to be linear in stress and fluence with rates in the range of 1.7 - 5.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. Annealed and cold-worked materials exhibited similar creep rates. There is some indication that austenitic alloys with TiC or TiO precipitates had a slightly higher irradiation creep rate than those without. The ferritic alloys HT-9 and Fe-16Cr had irradiatoin creep rates about 0.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. No meaningful data could be obtained from the tubes irradiated at 60{degree}C because of damage to the tubes.

  15. Optimization process for the design of the DCLL blanket for the European DEMOnstration fusion reactor according to its nuclear performances

    Science.gov (United States)

    Palermo, Iole; Rapisarda, David; Fernández-Berceruelo, Iván; Ibarra, Angel

    2017-07-01

    The research study focuses on the neutronic design analysis and optimization of one of the options for a fusion reactor designed as DCLL (dual coolant lithium-lead). The main objective has been to develop an efficient and technologically viable modular DCLL breeding blanket (BB) using the DEMO generic design specifications established within the EUROfusion Programme. The final neutronic design has to satisfy the requirements of: tritium self-sufficiency; BB thermal efficiency; preservation of plasma confinement; temperature limits imposed by materials; and radiation limits to guarantee the largest operational life for all the components. Therefore, a 3D fully heterogeneous DCLL neutronic model has been developed for the DEMO baseline 2014 determining its behaviour under the real operational conditions of the DEMO reactor. Consequent actions have been adopted to improve its performances. Neutronic assessments have specially addressed tritium breeding ratio, multiplication energy factor, power density distributions, damage and shielding responses. The model has then been adapted to the subsequent DEMO baseline 2015 (with a more powerful and bigger plasma, smaller divertor and bigger blanket segments), implying new design choices to improve the reactor nuclear performances.

  16. Adaptability of optimization concept in the context of cryogenic distribution for superconducting magnets of fusion machine

    Science.gov (United States)

    Sarkar, Biswanath; Bhattacharya, Ritendra Nath; Vaghela, Hitensinh; Shah, Nitin Dineshkumar; Choukekar, Ketan; Badgujar, Satish

    2012-06-01

    Cryogenic distribution system (CDS) plays a vital role for reliable operation of largescale fusion machines in a Tokamak configuration. Managing dynamic heat loads from the superconducting magnets, namely, toroidal field, poloidal field, central solenoid and supporting structure is the most important function of the CDS along with the static heat loads. Two concepts are foreseen for the configuration of the CDS: singular distribution and collective distribution. In the first concept, each magnet is assigned with one distribution box having its own sub-cooler bath. In the collective concept, it is possible to share one common bath for more than one magnet system. The case study has been performed with an identical dynamic heat load profile applied to both concepts in the same time domain. The choices of a combined system from the magnets are also part of the study without compromising the system functionality. Process modeling and detailed simulations have been performed for both the options using Aspen HYSYS®. Multiple plasma pulses per day have been considered to verify the residual energy deposited in the superconducting magnets at the end of the plasma pulse. Preliminary 3D modeling using CATIA® has been performed along with the first level of component sizing.

  17. Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Werner, R.W. (ed.)

    1982-11-01

    This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

  18. A Concept Exploration Program in Fast Ignition Inertial Fusion — Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Stephens, Richarad Burnite [General Atomics; Freeman, Richard R. [The Ohio State University; Van Woekom, L. D. [The Ohio State University; Key, M. [Lawrence Livermore National Laboratory; MacKinnon, Andrew J. [Lawrence Livermore National Laboratory; Wei, Mingsheng [General Atomics

    2014-02-27

    The Fast Ignition (FI) approach to Inertial Confinement Fusion (ICF) holds particular promise for fusion energy because the independently generated compression and ignition pulses allow ignition with less compression, resulting in (potentially) higher gain. Exploiting this concept effectively requires an understanding of the transport of electrons in prototypical geometries and at relevant densities and temperatures. Our consortium, which included General Atomics (GA), The Ohio State University (OSU), the University of California, San Diego (UCSD), University of California, Davis (UC-Davis), and Princeton University under this grant (~$850K/yr) and Lawrence Livermore National Laboratory (LLNL) under a companion grant, won awards in 2000, renewed in 2005, to investigate the physics of electron injection and transport relevant to the FI concept, which is crucial to understand electron transport in integral FI targets. In the last two years we have also been preparing diagnostics and starting to extend the work to electron transport into hot targets. A complementary effort, the Advanced Concept Exploration (ACE) program for Fast Ignition, was funded starting in 2006 to integrate this understanding into ignition schemes specifically suitable for the initial fast ignition attempts on OMEGA and National Ignition Facility (NIF), and during that time these two programs have been managed as a coordinated effort. This result of our 7+ years of effort has been substantial. Utilizing collaborations to access the most capable laser facilities around the world, we have developed an understanding that was summarized in a Fusion Science & Technology 2006, Special Issue on Fast Ignition. The author lists in the 20 articles in that issue are dominated by our group (we are first authors in four of them). Our group has published, or submitted 67 articles, including 1 in Nature, 2 Nature Physics, 10 Physical Review Letters, 8 Review of Scientific Instruments, and has been invited to

  19. A Study of the Flow Patterns of Expanding Impurity Aerosol Following a Disruption Event in a Fusion Reactor

    Science.gov (United States)

    Majumdar, Rudrodip

    The current study focuses on the adiabatic expansion of aerosol impurity in the post-disruption and thermal quench scenario inside the vacuum chamber of a fusion reactor. A pulsed electrothermal plasma (ET) capillary source has been used as a source term simulating the surface ablation of the divertor or other interior critical components of a tokamak fusion reactor under hard disruption-like conditions. The capillary source generates particulates from wall evaporation by depositing transient radiant high heat flux onto the inner liner of the capillary. The particulates form a plasma jet moving towards the capillary exit at high speed and high pressure. The first chapter discusses briefly the relevance of the study pertaining to the impurities in a fusion reactor based on the work available in the form of published literature. The second chapter discusses briefly the operating principle of a pulsed electrothermal plasma source (PEPS), the virtual integration of PEPS with 1-D electrothermal plasma flow solver ETFLOW and the use of capillary plasma sources in various industrial applications. The third chapter discusses about primitive computational work, backed by the data from actual electrothermal source experiments from the in-house facility "PIPE" (Plasma Interactions with Propellants Experiment), that shows the supersonic bulk flow patterns for the temperature, density, pressure, bulk velocity and the flow Mach number of the impurity particulates as they get ejected as a high-pressure, high-temperature and hyper-velocity jet from the simulated source term. It also shows the uniform steady-state subsonic expansion of bulk aerosol inside the expansion chamber. The fourth chapter discusses scaling laws in 1-D for the aforesaid bulk plasma parameters for ranges of axial length traversed by the flow, so that one can retrieve the flow parameters at some preferred locations. The fifth chapter discusses the effect of temperature and the non--linearity of the adiabatic

  20. Work Domain Analysis Methodology for Development of Operational Concepts for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This report describes a methodology to conduct a Work Domain Analysis in preparation for the development of operational concepts for new plants. This method has been adapted from the classical method described in the literature in order to better deal with the uncertainty and incomplete information typical of first-of-a-kind designs. The report outlines the strategy for undertaking a Work Domain Analysis of a new nuclear power plant and the methods to be used in the development of the various phases of the analysis. Basic principles are described to the extent necessary to explain why and how the classical method was adapted to make it suitable as a tool for the preparation of operational concepts for a new nuclear power plant. Practical examples are provided of the systematic application of the method and the various presentation formats in the operational analysis of advanced reactors.

  1. Minimally Invasive Direct Lateral Interbody Fusion (MIS-DLIF): Proof of Concept and Perioperative Results.

    Science.gov (United States)

    Abbasi, Hamid; Abbasi, Ali

    2017-01-14

    Minimally invasive direct lateral interbody fusion (MIS-DLIF) is a novel approach for fusions of the lumbar spine. In this proof of concept study, we describe the surgical technique and report our experience and the perioperative outcomes of the first nine patients who underwent this procedure. In this study we establish the safety and efficacy of this approach. MIS-DLIF was performed on 15 spinal levels in nine patients who failed to respond to conservative therapy for the treatment of a re-herniated disk, spondylolisthesis, or other severe disk disease of the lumbar spine. We recorded surgery time, blood loss, fluoroscopy time, patient-reported pain, and complications. Throughout the MIS-DLIF procedure, the surgeon is aided by biplanar fluoroscopic imaging to place an interbody graft or cage into the disc space through the interpleural space. A discectomy is performed in the same minimally invasive fashion. The procedure is usually completed with posterior pedicle screw fixation. MIS-DLIF took 44/85 minutes, on average, for 1/2 levels, with 54/112 ml of blood loss, and 0.3/1.7 days of hospital stay. Four of nine patients did not require overnight hospitalization and were discharged two to four hours after surgery. We did not encounter any clinically significant complications. At more than ninety days post surgery, the patients reported a statistically significant reduction of 4.5 points on a 10-point sliding pain scale. MIS-DLIF with pedicle screw fixation is a safe and clinically effective procedure for fusions of the lumbar spine. The procedure overcomes many of the limitations of the current minimally invasive approaches to the lumbar spine and is technically straightforward. MIS-DLIF has the potential to improve patient outcomes and reduce costs relative to the current standard of care and therefore warrants further investigation. We are currently expanding this study to a larger cohort and documenting long-term outcome data.

  2. Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

    Science.gov (United States)

    Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

    Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

  3. Economics and Environmental Compatibility of Fusion Reactors —Its Analysis and Coming Issues— 4.Economic Effect of Fusion in Energy Market 4.2Various Externalities of Energy Systems and the Integrated Evaluation

    Science.gov (United States)

    Ito, Keishiro

    The primacy of a nuclear fusion reactor in a competitive energy market remarkably depends on to what extent the reactor contributes to reduce the externalities of energy. The reduction effects are classified into two effects, which have quite dissimilar characteristics. One is an effect of environmental dimensions. The other is related to energy security. In this study I took up the results of EC's Extern Eproject studies as are presentative example of the former effect. Concerning the latter effect, I clarified the fundamental characteristics of externalities related to energy security and the conceptual framework for the purpose of evaluation. In the socio-economical evaluation of research into and development investments in nuclear fusions reactors, the public will require the development of integrated evaluation systems to support the cost-effect analysis of how well the reduction effects of externalities have been integrated with the effects of technological innovation, learning, spillover, etc.

  4. Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept

    Directory of Open Access Journals (Sweden)

    Haileyesus Tsige-Tamirat

    2015-01-01

    Full Text Available Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required.

  5. A concept of JAERI passive safety light water reactor system (JPSR)

    Energy Technology Data Exchange (ETDEWEB)

    Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  6. The snowflake divertor, physics of a new concept for power exhaust of fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lunt, Tilmann; Feng, Yuehe [Max-Planck-Institut fuer Plasmaphysik, Garching/Greifswald (Germany); Canal, Gustavo; Reimerdes, Holger [Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland)

    2014-07-01

    Fusion reactors based on the tokamak design will have to deal with very high heat loads on the divertor plates. One of the approaches to solve this heat load problem is the so called 'snowflake divertor', a magnetic configuration with two nearby x-points and two additional divertor legs. In this contribution we report on 'EMC3-Eirene' simulations of the plasma- and neutral particle transport in the scrape-off layer of the swiss tokamak TCV of a series of snowflake equilibria with different values of σ, the distance between the x-points normalized to the minor radius of the plasma. The constant anomalous transport coefficients were chosen such that the power- and particle deposition profiles at the primary inner strike point match the Langmuir probe measurements for the σ=0.1 case. At one of the secondary strike points, however, a significantly larger power flux than that predicted by the simulation was measured by the probes, indicating the presence of an enhanced transport across the primary separatrix. We discuss the possible reason for this enhanced transport as well as its scaling with machine size. Another prediction from the simulation is that the density as well as the radiation maximum are moving from the recycling region in front of the plates upwards to the x-point.

  7. Exchange Flow Characteristics in a Tokamak Vacuum Vessel of Fusion Reactor Under the Loss-of-Vacuum Conditions

    Science.gov (United States)

    Takase, Kazuyuki; Kunugi, Tomoaki; Seki, Yasushi

    1997-06-01

    When a Tokamak vacuum vessel of fusion reactor is broken, buoyancy-driven exchange flows will take place through breaches after the inside pressure of the vacuum vessel (VV) becomes equal to the outside pressure. The exchange flow may bring a mixture of activated dusts and tritium from the inside of the VV to the outside through the breaches. Moreover, the exchange flow may remove decay heat from the plasma-facing components. A preliminary LOVA (Loss Of VAcuum event) apparatus was constructed to investigate quantitative heat transfer characteristics of the exchange flows through the breaches under the LOVA conditions. The results of this study, the relationship between Froude numbers and breach locations in the VV was determined and empirical correlations for the average Froude numbers were derived.

  8. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy

    CERN Document Server

    Gsponer, A; Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This paper contains two parts: (I) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (II) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showing that while full access to the physics of thermonuclear weapons is the main implication of ICF, full access to large-scale tritium technology is the main proliferation impact of MCF. The conclusion of the paper is that siting ITER in a country such as Japan, which already has a large separated-plutonium stockpile, and an ambitious laser-driven ICF program (comparable in size and quality to those of the United States or France) will considerably increase its latent (or virtual) nuclear weapons proliferation status, and fo...

  9. Simulation of rapid heating in fusion reactor first walls using the Green's function approach

    Energy Technology Data Exchange (ETDEWEB)

    Hassanein, A.M.; Kulcinski, G.L.

    1984-08-01

    The solution of the heat conduction problem in moving boundary conditions is very important in predicting accurate thermal behavior of materials when very high energy deposition is expected. Such high fluxes are encountered on first wall materials nd other components in fusion reactors. A numerical method has been developed to solve this problem by the use of the Green's function. A comparison is made between this method and a finite difference one. The comparison in the finite difference method is made with and without the variation of the thermophysical properties with temperature. The agreement between Green's function and the finite difference method is found to be very good. The advantages and disadvantages of using the Green's function method and the importance of the variation of material thermal properties with temperature are discussed.

  10. R&D around a photoneutralizer-based NBI system (Siphore) in view of a DEMO Tokamak steady state fusion reactor

    Science.gov (United States)

    Simonin, A.; Achard, Jocelyn; Achkasov, K.; Bechu, S.; Baudouin, C.; Baulaigue, O.; Blondel, C.; Boeuf, J. P.; Bresteau, D.; Cartry, G.; Chaibi, W.; Drag, C.; de Esch, H. P. L.; Fiorucci, D.; Fubiani, G.; Furno, I.; Futtersack, R.; Garibaldi, P.; Gicquel, A.; Grand, C.; Guittienne, Ph.; Hagelaar, G.; Howling, A.; Jacquier, R.; Kirkpatrick, M. J.; Lemoine, D.; Lepetit, B.; Minea, T.; Odic, E.; Revel, A.; Soliman, B. A.; Teste, P.

    2015-11-01

    Since the signature of the ITER treaty in 2006, a new research programme targeting the emergence of a new generation of neutral beam (NB) system for the future fusion reactor (DEMO Tokamak) has been underway between several laboratories in Europe. The specifications required to operate a NB system on DEMO are very demanding: the system has to provide plasma heating, current drive and plasma control at a very high level of power (up to 150 MW) and energy (1 or 2 MeV), including high performances in term of wall-plug efficiency (η  >  60%), high availability and reliability. To this aim, a novel NB concept based on the photodetachment of the energetic negative ion beam is under study. The keystone of this new concept is the achievement of a photoneutralizer where a high power photon flux (~3 MW) generated within a Fabry-Perot cavity will overlap, cross and partially photodetach the intense negative ion beam accelerated at high energy (1 or 2 MeV). The aspect ratio of the beam-line (source, accelerator, etc) is specifically designed to maximize the overlap of the photon beam with the ion beam. It is shown that such a photoneutralized based NB system would have the capability to provide several tens of MW of D0 per beam line with a wall-plug efficiency higher than 60%. A feasibility study of the concept has been launched between different laboratories to address the different physics aspects, i.e. negative ion source, plasma modelling, ion accelerator simulation, photoneutralization and high voltage holding under vacuum. The paper describes the present status of the project and the main achievements of the developments in laboratories.

  11. On the energy gain enhancement of DT+D3He fuel configuration in nuclear fusion reactor driven by heavy ion beams

    Directory of Open Access Journals (Sweden)

    S Khoshbinfar

    2016-09-01

    Full Text Available It is expected that advanced fuels be employed in the second generation of nuclear fusion reactors. Theoretical calculations show that in such a fuel, a high plasma temperature about 100 keV is a requisite for reaction rate improvement of nuclear fusion. However, creating such a temporal condition requires a more powerful driver than we have today. Here, introducing an optimal fuel configuration consisting of DT and D-3He layers, suitable for inertial fusion reactors and driven by heavy ion beams, the optimal energy gain conditions have been simulated and derived for 1.3 MJ system. It was found that, in this new fuel configuration, the ideal energy gain, is 22 percent more comparing with energy gain in corresponding single DT fuel layer. Moreover, the inner DT fuel layer contributed as an ignition trigger, while the outer D3He fuel acts as particle and radiation shielding as well as fuel layer.

  12. Improved structure and long-life blanket concepts for heliotron reactors

    Science.gov (United States)

    Sagara, A.; Imagawa, S.; Mitarai, O.; Dolan, T.; Tanaka, T.; Kubota, Y.; Yamazaki, K.; Watanabe, K. Y.; Mizuguchi, N.; Muroga, T.; Noda, N.; Kaneko, O.; Yamada, H.; Ohyabu, N.; Uda, T.; Komori, A.; Sudo, S.; Motojima, O.

    2005-04-01

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R&D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated.

  13. Fusion reactor materials: Semiannual progress report for the period ending March 31, 1988

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1988-08-01

    This report contains papers on thermonuclear reactor materials. The general categories of these papers are: irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; development of structural alloys; solid breeding materials; ceramics; and radiation effects. Selected papers have been processed for inclusion in the energy database. (LSP)

  14. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

    1992-03-01

    This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

  15. Fusion reactor materials semiannual progress report for the period ending March 31, 1990

    Energy Technology Data Exchange (ETDEWEB)

    1990-08-01

    This report mainly discusses topics on the physical effects of radiation on thermonuclear reactor materials. The areas discussed are: irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; fundamental mechanical behavior; radiation effects; mechanistic studies, theory and modeling; development of structural alloys; solid breeding materials; and ceramics. (FI)

  16. Fusion Reactor Materials semiannual progress report for period ending September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1992-04-01

    This report contains papers on topic in the following areas of thermonuclear reactor materials: irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials and beryllium; and ceramics. These paper have been index separately elsewhere. (LSP).

  17. Progress in physics design of fusion-fission hybrid energy reactor%次临界能源堆物理设计进展

    Institute of Scientific and Technical Information of China (English)

    李茂生; 贾建平; 程和平; 蒋洁琼; 栗再新; 杨永伟; 吴宏春; 师学明; 刘荣; 鹿心鑫; 朱通华; 王新华; 余泳; 严钧; 唐涛

    2014-01-01

    聚变-裂变混合能源堆包括聚变中子源和次临界能源堆,主要目标是生产电能。回顾了国内外混合堆的发展历史,给出混合能源堆设计的边界条件和约束条件,说明次临界能源堆以铀锆合金为燃料、水为冷却剂的设计思想。利用输运燃耗耦合程序 MCORGS 计算了混合能源的燃耗,给出了中子有效增殖因数、能量放大倍数和氚增殖比等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。论述了混合堆的热工设计并进行了安全分析。对于燃耗数值模拟程序,通过多家对算,保证其计算结果的可信性。针对次临界能源堆的特点,利用贫铀球壳建立了贫铀聚乙烯装置和贫铀 LiH 装置,并且专门设计加工了天然铀装置,开展铀裂变率、造钚率、产氚率等中子学积分实验,验证了数值模拟的可靠性。%In this paper,we propose a preliminary design for a fusion-fission hybrid energy reactor (FFHER),based on cur-rent fusion science and technology and well-developed fission technology.Design rules are listed and a primary concept blanket with uranium alloy as fuel and water as coolant is put forward.The uranium fuel can be natural uranium,LWR spent fuel,or de-pleted uranium.The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the de-velopment of nuclear energy.The interaction between the fusion neutron and the uranium fuel with the aim of achieving greater energy multiplication and tritium sustainability is studied.Other concept hybrid reactor designs are also reviewed.Integral neu-tron experiments were carried out to verify the credibility of our proposed physical design.The combination of the physical design with the related thermal hydraulic design,alloy fuel manufacture,and nuclear fuel cycle programs provides the

  18. Local transport barrier formation and relaxation in reverse-shear plasmas on the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Synakowski, E. J.; Batha, S. H.; Beer, M. A.; Bell, M. G.; Bell, R. E.; Budny, R. V.; Bush, C. E.; Efthimion, P. C.; Hahm, T. S.; Hammett, G. W.; LeBlanc, B.; Levinton, F.; Mazzucato, E.; Park, H.; Ramsey, A. T.; Schmidt, G.; Rewoldt, G.; Scott, S. D.; Taylor, G.; Zarnstorff, M. C.

    1997-05-01

    The roles of turbulence stabilization by sheared E×B flow and Shafranov shift gradients are examined for Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] enhanced reverse-shear (ERS) plasmas. Both effects in combination provide the basis of a positive-feedback model that predicts reinforced turbulence suppression with increasing pressure gradient. Local fluctuation behavior at the onset of ERS confinement is consistent with this framework. The power required for transitions into the ERS regime are lower when high power neutral beams are applied earlier in the current profile evolution, consistent with the suggestion that both effects play a role. Separation of the roles of E×B and Shafranov shift effects was performed by varying the E×B shear through changes in the toroidal velocity with nearly steady-state pressure profiles. Transport and fluctuation levels increase only when E×B shearing rates are driven below a critical value that is comparable to the fastest linear growth rates of the dominant instabilities. While a turbulence suppression criterion that involves the ratio of shearing to linear growth rates is in accord with many of these results, the existence of hidden dependencies of the criterion is suggested in experiments where the toroidal field was varied. The forward transition into the ERS regime has also been examined in strongly rotating plasmas. The power threshold is higher with unidirectional injection than with balance injection.

  19. The role of the boundary plasma in defining the viability of a magnetic fusion reactor: A review

    Science.gov (United States)

    Whyte, Dennis

    2012-10-01

    The boundary of magnetic confinement devices, from the pedestal through to the surrounding surfaces, encompasses an enormous range of plasma and material physics, and their integrated coupling. It is becoming clear that due to fundamental limits of plasma stability and material response the boundary will largely define the viability of an MFE reactor. However we face an enormous knowledge deficit in stepping from present devices and ITER towards a demonstration power plant. We review the boundary and plasma-material interaction (PMI) research required to address this deficit as well as related theoretical/scaling methods for extending present results to future devices. The research activities and gaps are reviewed and organized to three major axes of challenges: power density, plasma duration, and material temperature. The boundary can also be considered a multi-scale system of coupled plasma and material science regulated through the non-linear interface of the sheath. Measurement, theory and modeling across these scales are reviewed. Dimensionless parameters, often used to organized core plasma transport on similarity arguments, can be extended to the boundary plasma, plasma-surface interactions and material response. The scaling methodology suggests intriguing ways forward to prescribe and understand the boundary issues of an eventual reactor in intermediate size devices. Finally, proposed technology and science innovations towards solving the extreme PMI/boundary challenges of magnetic fusion energy will be reviewed.

  20. Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

    Energy Technology Data Exchange (ETDEWEB)

    Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Gironimo, G. Di, E-mail: giuseppe.digironimo@unina.it [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Esposito, G. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Mäkinen, H. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Miccichè, G. [ENEA Brasimone, I:40032 Camugnano (Italy); Mozzillo, R. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy)

    2016-11-15

    Highlights: • Concept Studies in Divertor Remote Handling. • Prioritization of concept alternatives. • Comparison and evaluation of product alternatives using AHP. - Abstract: The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field.

  1. Fusion power production in International Thermonuclear Experimental Reactor baseline H-mode scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Kessel, C. E. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08540 (United States); Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States)

    2015-04-15

    Self-consistent simulations of 15 MA ITER H-mode DT scenarios, from ramp-up through flat-top, are carried out. Electron and ion temperatures, toroidal angular frequency, and currents are evolved, in simulations carried out using the predictive TRANSPort and integrated modeling code starting with initial profiles and equilibria obtained from tokamak simulation code studies. Studies are carried out examining the dependence and sensitivity of fusion power production on electron density, argon impurity concentration, choice of radio frequency heating, pedestal temperature without and with E × B flow shear effects included, and the degree of plasma rotation. The goal of these whole-device ITER simulations is to identify dependencies that might impact ITER fusion performance.

  2. Conceptual study of ferromagnetic pebbles for heat exhaust in fusion reactors with short power decay length

    Directory of Open Access Journals (Sweden)

    N. Gierse

    2015-03-01

    The key results of this study are that very high heat fluxes are accessible in the operation space of ferromagnetic pebbles, that ferromagnetic pebbles are compatible with tokamak operation and current divertor designs, that the heat removal capability of ferromagnetic pebbles increases as λq decreases and, finally, that for fusion relevant values of q∥ pebble diameters below 100 μm are required.

  3. Mirror Advanced Reactor Study (MARS). Final report. Volume 2. Commercial fusion synfuels plant

    Energy Technology Data Exchange (ETDEWEB)

    Donohue, M.L.; Price, M.E. (eds.)

    1984-07-01

    Volume 2 contains the following chapters: (1) synfuels; (2) physics base and parameters for TMR; (3) high-temperature two-temperature-zone blanket system for synfuel application; (4) thermochemical hydrogen processes; (5) interfacing the sulfur-iodine cycle; (6) interfacing the reactor with the thermochemical process; (7) tritium control in the blanket system; (8) the sulfur trioxide fluidized-bed composer; (9) preliminary cost estimates; and (10) fuels beyond hydrogen. (MOW)

  4. Transient Behaviour of Superconducting Magnet Systems of Fusion Reactor ITER during Safety Discharge

    Directory of Open Access Journals (Sweden)

    A. M. Miri

    2008-01-01

    Full Text Available To investigate the transient behaviour of the toroidal and poloidal field coils magnet systems of the International Thermonuclear Experimental Reactor during safety discharge, network models with lumped elements are established. Frequency-dependant values of the network elements, that is, inductances and resistances are calculated with the finite element method. That way, overvoltages can be determined. According to these overvoltages, the insulation coordination of coils has to be selected.

  5. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  6. Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

    Directory of Open Access Journals (Sweden)

    Dalin Zhang

    2012-06-01

    Full Text Available Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium during a core disruptive accident (CDA may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within the CP-ESFR European project, focusing on the phenomena affecting the initiation and the transition phases of an unprotected loss of flow (ULOF accident. Key phenomena for the initiation phase are coolant boiling onset and further voiding of the core that lead to a reactivity increase in the case of a positive void reactivity effect. Therefore, the first level of optimization involves the reduction, by design, of the positive void effect in order to avoid entering a severe accident. If the core disruption cannot be avoided, the accident enters into the transition phase, characterized by the progression of core melting and recriticalities due to fuel compaction. Dedicated features that enhance and guarantee a sufficient and timely fuel discharge are considered for the optimization of this phase.

  7. Minimally Invasive Direct Lateral Interbody Fusion (MIS-DLIF): Proof of Concept and Perioperative Results

    Science.gov (United States)

    Abbasi, Hamid

    2017-01-01

    Background Minimally invasive direct lateral interbody fusion (MIS-DLIF) is a novel approach for fusions of the lumbar spine. In this proof of concept study, we describe the surgical technique and report our experience and the perioperative outcomes of the first nine patients who underwent this procedure. Study design/setting In this study we establish the safety and efficacy of this approach. MIS-DLIF was performed on 15 spinal levels in nine patients who failed to respond to conservative therapy for the treatment of a re-herniated disk, spondylolisthesis, or other severe disk disease of the lumbar spine. We recorded surgery time, blood loss, fluoroscopy time, patient-reported pain, and complications. Methods Throughout the MIS-DLIF procedure, the surgeon is aided by biplanar fluoroscopic imaging to place an interbody graft or cage into the disc space through the interpleural space. A discectomy is performed in the same minimally invasive fashion. The procedure is usually completed with posterior pedicle screw fixation. Results MIS-DLIF took 44/85 minutes, on average, for 1/2 levels, with 54/112 ml of blood loss, and 0.3/1.7 days of hospital stay. Four of nine patients did not require overnight hospitalization and were discharged two to four hours after surgery. We did not encounter any clinically significant complications. At more than ninety days post surgery, the patients reported a statistically significant reduction of 4.5 points on a 10-point sliding pain scale. Conclusions MIS-DLIF with pedicle screw fixation is a safe and clinically effective procedure for fusions of the lumbar spine. The procedure overcomes many of the limitations of the current minimally invasive approaches to the lumbar spine and is technically straightforward. MIS-DLIF has the potential to improve patient outcomes and reduce costs relative to the current standard of care and therefore warrants further investigation. We are currently expanding this study to a larger cohort and

  8. Overview of the VISTA Spacecraft Concept Powered by Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Orth, C D

    2000-11-21

    VISTA was conceived through a detailed systems analysis as a viable, realistic, and defensible spacecraft concept based on advanced ICF technology but existing or near-term technology for other systems. It is a conical self-contained single-stage piloted spacecraft in which a magnetic thrust chamber directs the plasma emissions from inertial confinement fusion (ICF) targets into a rearward exhaust. VISTA's propulsion system is therefore unique because it is based on (1) a rather mature technology (ICF), which is known to work with sufficient driver input; (2) direct heating of all expellant by the fusion process, thus providing high mass flow rates without significant degradation of jet efficiency; and (3) a magnetic thrust chamber, which avoids the plasma thermalization and resultant degradation of specific impulse that are unavoidable with the use of mechanical thrust chambers. VISTA therefore has inherently high power/mass ratios and high specific impulses. With advanced ICF technology, ultra-fast roundtrips (RTs) to objects within the solar system are possible (e.g., {ge}145 days RT to Mars, {ge}7 years RT to Pluto). Such short-duration missions are imperative to minimize the human physiological deteriorations arising from zero gravity and the cosmic-radiation. In addition, VISTA offers on-board artificial gravity and propellant-based shielding from cosmic rays, thus reducing the physiological deteriorations to insignificant levels. In this paper, we give an overview of the various vehicle systems for this concept, estimate the general missions performance capabilities for interplanetary missions, and describe in detail the performance for the baseline mission of a piloted roundtrip to Mars with a 100-ton payload. Items requiring further research include a reduction of the wet mass from its baseline value of 6,000 metric tons, and the development of fast ignition or its equivalent to provide target gains in excess of several hundred. With target gains well

  9. Numerical simulations of in-situ neutron detector calibration experiments on the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ku, L.P.; Hendel, H.W.; Liew, S.L.; Strachan, J.D.

    1990-02-01

    Accurate determinations of fusion neutron yields on the TFTR require that the neutron detectors be absolutely calibrated in-situ, using neutron sources of known strengths. For such calibrations, numerical simulations of neutron transport can be powerful tools in the design of experiments and the study of measurement results. On the TFTR, numerical calibration experiments' have been frequently used to complement actual detector calibrations. We present calculational approaches and transport models used in these numerical simulations, and summarize the results from simulating the calibration of {sup 235}U fission detectors carried out in December 1988. 12 refs., 9 figs., 6 tabs.

  10. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  11. Helium desorption in EFDA iron materials for use in nuclear fusion reactors; Desorcion de helio en materiales de fierro EFDA para su aplicacion en los reactores de fusion nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Salazar R, A. R.; Pinedo V, J. L. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Sanchez, F. J.; Ibarra, A.; Vila, R., E-mail: arsr2707@hotmail.com [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Av. Complutense No. 40, 28040 Madrid (Spain)

    2015-09-15

    In this paper the implantation with monoenergetic ions (He{sup +}) was realized with an energy of 5 KeV in iron samples (99.9999 %) EFDA (European Fusion Development Agreement) using a collimated beam, after this a Thermal Desorption Spectrometry of Helium (THeDS) was made using a leak meter that detects amounts of helium of up to 10{sup -}- {sup 12} mbar l/s. Doses with which the implantation was carried out were 2 x 10{sup 15} He{sup +} /cm{sup 2}, 1 x 10{sup 16} He{sup +} /cm{sup 2}, 2 x 10{sup 16} He{sup +} /cm{sup 2}, 1 x 10{sup 17} He{sup +} /cm{sup 2} during times of 90 s, 450 s, 900 s and 4500 s, respectively. Also, using the SRIM program was calculated the depth at which the helium ions penetrate the sample of pure ion, finding that the maximum distance is 0.025μm in the sample. For this study, 11 samples of Fe EFDA were prepared to find defects that are caused after implantation of helium in order to provide valuable information to the manufacture of materials for future fusion reactors. However understand the effects of helium in the micro structural evolution and mechanical properties of structural materials are some of the most difficult questions to answer in materials research for nuclear fusion. When analyzing the spectra of THeDS was found that five different groups of desorption peaks existed, which are attributed to defects of He caused in the material, these defects are He{sub n} V (2≤n≤6), He{sub n} V{sub m}, He V for the groups I, II and IV respectively. These results are due to the comparison of the peaks presented in the desorption spectrum of He, with those of other authors who have made theoretical calculations. Is important to note that the thermal desorption spectrum of helium was different depending on the dose with which the implantation of He{sup +} was performed. (Author)

  12. A Concept for a Low Pressure Noble Gas Fill Intervention in the IFE Fusion Test Facility (FTF) Target Chamber

    Energy Technology Data Exchange (ETDEWEB)

    Gentile, C. A.; Blanchard, W. R.; Kozub, T. A.; Aristova, M.; McGahan, C.; Natta, S.; Pagdon, K.; Zelenty, J.

    2010-01-14

    An engineering evaluation has been initiated to investigate conceptual engineering methods for implementing a viable gas shield strategy in the Fusion Test Facility (FTF) target chamber. The employment of a low pressure noble gas in the target chamber to thermalize energetic helium ions prior to interaction with the wall could dramatically increase the useful life of the first wall in the FTF reactor1. For the purpose of providing flexibility, two target chamber configurations are addressed: a five meter radius sphere and a ten meter radius sphere. Experimental studies at Nike have indicated that a low pressure, ambient gas resident in the target chamber during laser pulsing does not appear to impair the ability of laser light from illuminating targets2. In addition, current investigations into delivering, maintaining, and processing low pressure gas appear to be viable with slight modification to current pumping and plasma exhaust processing technologies3,4. Employment of a gas fill solution for protecting the dry wall target chamber in the FTF may reduce, or possibly eliminate the need for other attenuating technologies designed for keeping He ions from implanting in first wall structures and components. The gas fill concept appears to provide an effective means of extending the life of the first wall while employing mostly commercial off the shelf (COTS) technologies. Although a gas fill configuration may provide a methodology for attenuating damage inflicted on chamber surfaces, issues associated with target injection need to be further analyzed to ensure that the gas fill concept is viable in the integrated FTF design5. In the proposed system, the ambient noble gas is heated via the energetic helium ions produced by target detonation. The gas is subsequently cooled by the chamber wall to approximately 800oC, removed from the chamber, and processed by the chamber gas processing system (CGPS). In an optimized scenario of the above stated concept, the chamber

  13. Preliminary Safeguards Assessment for the Pebble-Bed Fluoride High-Temperature Reactor (PB-FHR) Concept

    Energy Technology Data Exchange (ETDEWEB)

    Disser, Jay; Arthur, Edward; Lambert, Janine

    2016-09-01

    This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.

  14. The science program of the TCV tokamak: exploring fusion reactor and power plant concepts

    Science.gov (United States)

    Coda, S.; TCV Team

    2015-10-01

    TCV is acquiring a new 1 MW neutral beam and 2 MW additional third-harmonic electron cyclotron resonance heating (ECRH) to expand its operational range. Its existing shaping and ECRH launching versatility was amply exploited in an eclectic 2013 campaign. A new sub-ms real-time equilibrium reconstruction code was used in ECRH control of NTMs and in a prototype shape controller. The detection of visible light from the plasma boundary was also successfully used in a position-control algorithm. A new bang-bang controller improved stability against vertical displacements. The RAPTOR real-time transport simulator was employed to control the current density profile using electron cyclotron current drive. Shot-by-shot internal inductance optimization was demonstrated by iterative learning control of the current reference trace. Systematic studies of suprathermal electrons and ions in the presence of ECRH were performed. The L-H threshold power was measured to be ˜50-75% higher in both H and He than D, to increase with the length of the outer separatrix, and to be independent of the current ramp rate. Core turbulence was found to decrease from positive to negative edge triangularity deep into the core. The geodesic acoustic mode was studied with multiple diagnostics, and its axisymmetry was confirmed by a full toroidal mapping of its magnetic component. A new theory predicting a toroidal rotation component at the plasma edge, driven by inhomogeneous transport and geodesic curvature, was tested successfully. A new high-confinement mode (IN-mode) was found with an edge barrier in density but not in temperature. The edge gradients were found to govern the scaling of confinement with current, power, density and triangularity. The dynamical interplay of confinement and magnetohydrodynamic modes leading to the density limit in TCV was documented. The heat flux profile decay lengths and heat load profile on the wall were documented in limited plasmas. In the snowflake (SF) divertor configuration the heat flux profiles were documented on all four strike points. SF simulations with the EMC3-EIRENE code, including the physics of the secondary separatrix, underestimate the flux to the secondary strike points, possibly resulting from steady-state E × B drifts. With neon injection, radiation in a SF was 15% higher than in a conventional divertor. The novel triple-null and X-divertor configurations were also achieved in TCV.

  15. Laser-induced mobilization of dust produced during fusion reactors operation; Mise en suspension par laser de poussieres generees lors du fonctionnement des reacteurs de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Vatry, A.

    2010-11-16

    During tokamak operation, plasma-wall interactions lead to material erosion process and dusts production. These dusts are mainly composed by carbon and tungsten, with sizes ranging from 10 nm to 100 {mu}m. For safety reasons and to guarantee an optimum reactor functioning, the dusts have to be kept in reasonable quantity. The dusts mobilization is a first step to collect them, and the laser is a promising technique for this application. To optimize the cleaning, physical mechanisms responsible for dust ejection induced by laser have been identified. Some particles, such as aggregates, are directly ablated by the laser. The metal droplets are ejected intact by an electrostatic force, induced by the photoelectrons. We also characterized the particles ejection to choose an appropriate collection device. (author) [French] Lors du fonctionnement d'une machine de fusion, les interactions plasma-parois conduisent a des processus d'erosion des materiaux et a la production de particules. Ces poussieres sont principalement composees de carbone et de tungstene. Pour des raisons de surete et afin de garantir un fonctionnement optimum du reacteur, il est important de garder en quantite raisonnable les poussieres dont la taille varie entre 10 nm et 100 {mu}m. La mise en suspension de ces poussieres est une etape preliminaire a leur recuperation, et le laser est une technique prometteuse pour cette application. Afin d'optimiser le nettoyage, les mecanismes physiques a l'origine de l'ejection induite par laser de ces poussieres ont ete identifies. Les agregats sont directement ablates par le laser et les gouttelettes metalliques sont ejectees intactes par une force electrostatique induite par les photoelectrons. Nous avons egalement caracterise l'ejection des particules pour choisir un systeme de recuperation adapte

  16. The use of LBB concept in French fast reactors: Application to SPX plant

    Energy Technology Data Exchange (ETDEWEB)

    Turbat, A.; Deschanels, H.; Sperandio, M. [and others

    1997-04-01

    The leak before break (LBB) concept was not used at the design level for SUPERPHENIX (SPX), but different studies have been performed or are in progress concerning different components : Main Vessel (MV), pipings. These studies were undertaken to improve the defense in depth, an approach used in all French reactors. In a first study, the LBB approach has been applied to the MV of SPX plant to verify the absence of risk as regards the core supporting function and to help in the definition of in-service inspection (ISI) program. Defining a reference semi-elliptic defect located in the welds of the structure, it is verified that the crack growth is limited and that the end-of-life defect is smaller than the critical one. Then it is shown that the hoop welds (those which are the most important for safety) located between the roof and the triple point verify the leak-before-break criteria. However, generally speaking, the low level of membrane primary stresses which is favorable for the integrity of the vessel makes the application of the leak-before-break concept more difficult due to small crack opening areas. Finally, the extension of the methodology to the secondary pipings of SPX incorporating recent European works of DCRC is briefly presented.

  17. Inertial confinement fusion with direct electric generation by magnetic flux comparession

    Energy Technology Data Exchange (ETDEWEB)

    Lasche, G.P.

    1983-01-01

    A high-power-density laser-fusion-reactor concept in investigated in which directed kinetic enery imparted to a large mass of liquid lithium--in which the fusion target is centrally located--is maximized. In turn, this kinetic energy is converted directly to electricity with, potentially, very high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the concept maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall can be many orders of magnitude less than is typical of D-T fusion reactor concepts.

  18. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

  19. Application of railgun principle to high-velocity hydrogen pellet injection for magnetic fusion reactor refueling

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K.

    1991-08-01

    This report contains three documents describing the progress made by the University of Illinois electromagnetic railgun program sponsored by the Office of Fusion Energy of the United States Department of Energy during the period from July 16, 1990 to August 16, 1991. The first document contains a brief summary of the tasks initiated, continued, or completed, the status of major tasks, and the research effort distribution, estimated and actual, during the period. The second document contains a description of the work performed on time resolved laser interferometric density measurement of the railgun plasma-arc armature. The third document is an account of research on the spectroscopic measurement of the electron density and temperature of the railgun plasma arc.

  20. Self-ignition of an advanced fuel field-reversed configuration reactor by fusion product heating

    Energy Technology Data Exchange (ETDEWEB)

    Ohnishi, M.; Ohi, S.; Okamoto, M.; Momota, H.; Wakabayashi, J.

    1987-09-01

    A self-ignition of a deuterium-deuterium (D-D)-/sup 3/He fuel field-reversed configuration (FRC) plasma by fusion product heating is studied by using the point plasma model, where an FRC plasma equilibrium is taken into account. It is numerically demonstrated that the D-D-/sup 3/He plasma can be evolved from a deuterium-tritium burning plasma in a controlled manner by means of a compression-decompression control as well as a fueling control. It is also indicated that the increase of a trapped flux is effective for suppressing the excessive elongation of a plasma during the transition. The proposed method may provide a solution to the problem on plasma heating to attain a D-D-/sup 3/He self-ignition.

  1. Fusion reactor systems studies. Progress report, November 1, 1992--October 31, 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-09-09

    Fusion Technology Institute personnel actively participated in the ARIES/PULSAR project during the present contract period. Numerous presentations were made at PULSAR project meetings, major contributions were written for the ARIES-II/IV Final Report presentations and papers were given at technical conferences contributions were written for the ARIES Lessons Learned report and a very large number of electronic-mail and regular-mail communications were sent. The remaining sections of this progress report win summarize the work accomplished and in progress for the PULSAR project during the contract period. The main areas of effort are: PULSAR Research; ARIES-II/IV Report Contributions; ARIES Lessons Learned Report Contributions; and Stellarator Study.

  2. Conception of high safety reactor MAVR, technical and economical fuel cycle characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Kotov, V.M.; Cherepnin, Yu.S.

    1993-12-31

    Operation safety of reactor MAVR under nominal and emergency situations is based on creation of conditions for the minimum time of fuel operation in the core at the minimum quantity of the fissionable material. The variants of core elements construction, of the reactor control systems, and the possible scheme of fuel cycles of the reactor MAVR are considered.

  3. Prospects and problems using vanadium alloys as a structural material of the first wall and blanket of fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Votinov, S.N. [RSRC, Moscow (Russian Federation). A.A. Bochvar Inst. of Inorg. Mater.; Solonin, M.I. [RSRC, Moscow (Russian Federation). A.A. Bochvar Inst. of Inorg. Mater.; Kazennov, Yu.I. [RSRC, Moscow (Russian Federation). A.A. Bochvar Inst. of Inorg. Mater.; Kondratjev, V.P. [RSRC, Moscow (Russian Federation). A.A. Bochvar Inst. of Inorg. Mater.; Nikulin, A.D. [RSRC, Moscow (Russian Federation). A.A. Bochvar Inst. of Inorg. Mater.; Tebus, V.N. [RSRC, Moscow (Russian Federation). A.A. Bochvar Inst. of Inorg. Mater.; Adamov, E.O. [RDIPE, Moscow (Russian Federation); Bougaenko, S.E. [RDIPE, Moscow (Russian Federation); Strebkov, Yu.S. [RDIPE, Moscow (Russian Federation); Sidorenkov, A.V. [RDIPE, Moscow (Russian Federation); Ivanov, V.B. [Nauchno-Issledovatel`skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation); Kazakov, V.A. [Nauchno-Issledovatel`skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation); Evtikhin, V.A. [SE ``Krasnaya Zvezda``, Moscow (Russian Federation); Lyublinski, I.E. [SE ``Krasnaya Zvezda``, Moscow (Russian Federation); Trojanov, V.M. [SSC- IPPE, Obninsk (Russian Federation); Rusanov, A.E. [SSC- IPPE, Obninsk (Russian Federation); Chernov, V.M. [SSC- IPPE, Obninsk (Russian Federation); Birgevoj, G.A. [SSC- IPPE, Obninsk (Russian Federation)

    1996-10-01

    Vanadium-based alloys are most promising as low activation structural materials for DEMO. It was previously established that high priority is to be given to V-alloys of the V-Ti-Cr system as structural materials of a tritium breeding blanket and the first wall of a fusion reactor. However, there is some uncertainty in selecting a specific element ratio between the alloy components in this system. This is primarily explained by the fact that the properties of V-alloys are dictated not only by the ratio between the main alloying elements (here Ti and Cr), but also by impurities, both metallic and oxygen interstitials. Based on a number of papers today one can say that V-Ti-Cr alloys with insignificant variations in the contents of the main constituents within 5-10 mass% Ti and 4-6 mass% Cr must be taken as a base for subsequent optimization of chemical composition and thermomechanical working. However, the database is obviously insufficient to assess the ecological acceptability (activation), physical and mechanical properties, corrosion and irradiation resistance and, particularly, the commercial production of alloys. Therefore, there is a need for comprehensive studies of promising V-alloys, namely V-4Ti-4Cr and V-10Ti-5Cr. (orig.).

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

    Energy Technology Data Exchange (ETDEWEB)

    Takase, K.; Kunugi, T.; Shibata, M.; Seki, Y. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)

    1998-09-01

    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.) 6 refs.

  5. Microinstability properties of negative magnetic shear discharges in the Tokamak Fusion Test Reactor and DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Rewoldt, G.; Tang, W.M. [Princeton Univ., NJ (United States). Princeton Plasma Physics Lab.; Lao, L.L. [General Atomics, San Diego, CA (United States)

    1997-03-01

    The microinstability properties of discharges with negative (reversed) magnetic shear in the Tokamak Fusion Test Reactor (TFTR) and DIII-D experiments with and without confinement transitions are investigated. A comprehensive kinetic linear eigenmode calculation employing the ballooning representation is employed with experimentally measured profile data, and using the corresponding numerically computed magnetohydrodynamic (MHD) equilibria. The instability considered is the toroidal drift mode (trapped-electron-{eta}{sub i} mode). A variety of physical effects associated with differing q-profiles are explained. In addition, different negative magnetic shear discharges at different times in the discharge for TFTR and DIII-D are analyzed. The effects of sheared toroidal rotation, using data from direct spectroscopic measurements for carbon, are analyzed using comparisons with results from a two-dimensional calculation. Comparisons are also made for nonlinear stabilization associated with shear in E{sub r}/RB{sub {theta}}. The relative importance of changes in different profiles (density, temperature, q, rotation, etc.) on the linear growth rates is considered.

  6. Design and Analysis of HIP joined W and Ferritic-Martensitic Steel Mockup for Fusion Reactor Divertor Development

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. W.; Shin, K. I.; Kim, S. K.; Jin, H. G.; Lee, E. H.; Yoon, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Moon, S. Y.; Hong, B. G. [Chonbuk National Univ., Jeonju (Korea, Republic of)

    2013-10-15

    Korea has developed a Helium Cooled Ceramic Reflector (HCCR) based Test Blanket System (TBS) for an ITER, which consists of the First Wall (FW), Breeding Zone (BZ), Side Wall (SW), and BZ box. Among them, the FW is an important component which faces the plasma directly and, therefore, it is subjected to high heat and neutron loads. The FW of the TBM is considered to be composed of a beryllium (Be) armor as a plasma-facing material and Ferritic-Martensitic (FM) steel as a structure material, or a tungsten (W) armor and FM steel, or bare FM steel. Since Be/FMS and bare FMS were developed and proved by high heat flux (HHF) test, W armor and FM steel joining, fabricated mock-ups, and preparation of the high heat flux (HHF) test for integrity investigation are introduced in the present study. For the application to fusion reactor, joining methods with W to FMS has been developed. The W mock-up was fabricated with HIP considering Ti interlayer and PHHT condition. And the HHF test was prepared by performing the preliminary analysis to determine the test conditions. From the analysis heating and cooling conditions were determined for 0.5 and 1.0 MW/m2 heat fluxes. In the near future, the thermal life-time will be evaluated to determine the test period of the mockups by the mechanical analysis with ANSYS.

  7. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    Science.gov (United States)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-10-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 °C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 °C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI).

  8. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    Science.gov (United States)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  9. A review of nuclear data needs and their status for fusion reactor technology with some suggestions on a strategy to satisfy the requirements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L. [Argonne National Lab., IL (United States); Cheng, E.T. [TSI Research, Inc., Solana Beach, CA (United States)

    1991-09-01

    A review was performed on the needs and status of nuclear data for fusion-reactor technology. Generally, the status of nuclear data for fusion has been improved during the past two decades due to the dedicated effort of the nuclear data developers. However, there are still deficiencies in the nuclear data base, particularly in the areas of activation and neutron scattering cross sections. Activation cross sections were found to be unsatisfactory in 83 of the 153 reactions reviewed. The scattering cross sections for fluorine and boron will need to be improved at energies above 1 MeV. Suggestions concerning a strategy to address the specific fusion nuclear data needs for dosimetry and activation are also provided.

  10. Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1976-11-01

    This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn/sub 3/ conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended.

  11. Microstructural evolution in an austenitic stainless steel fusion reactor first wall

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.; Odette, G.R.

    1986-01-01

    A detailed rate-theory-based model of microstructural evolution under fast neutron irradiation has been developed. The prominent new aspect of this model is a treatment of dislocation evolution in which Frank faulted loops nucleate, grow and unfault to provide a source for network dislocations while the dislocation network can be simultaneously annihilated by a climb/glide process. The predictions of this model compare very favorably with the observed dose and temperature dependence of these key microstructural features over a broad range. This new description of dislocation evolution has been coupled with a previously developed model of cavity evolution and good agreement has been obtained between the predictions of the composite model and fast reactor swelling data as well. The results from the composite model also reveal that the various components of the irradiation-induced microstructure evolve in a highly coupled manner. The predictions of the composite model are more sensitive to parametric variations than more simple models. Hence, its value as a tool in data analysis and extrapolation is enhanced.

  12. Ceramics for fusion reactors: The role of the lithium orthosilicate as breeder

    Energy Technology Data Exchange (ETDEWEB)

    Carella, Elisabetta, E-mail: elisabetta.carella@ciemat.es [National Laboratory for Magnetic Fusion, CIEMAT, Madrid (Spain); Hernandez, Teresa [National Laboratory for Magnetic Fusion, CIEMAT, Madrid (Spain)

    2012-11-15

    Lithium-based oxide ceramics are studied as breeder blanket materials for the controlled thermonuclear reactors (CTR). Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is one of the most promising candidates because of its lithium concentration (0.54 g/cm{sup 3}), its high melting temperature (1523 K) and its excellent tritium release behavior. It is reported that the diffusion of tritium is closely related to that of lithium, so it is possible to find an indirect measure of the trend of tritium studying the diffusivity of Li{sup +}. In the present work, the synthesis of the Li{sub 4}SiO{sub 4} is carried out by Spray drying followed by pyrolysis. The study of the Li{sup +} ion diffusion on the sintered bodies, is investigated by means of electrical conductivity measurements. The effect of the {gamma}-ray irradiation is evaluated by the impedance spectroscopy method (EIS) from room temperature to 1173 K. The results indicate that the sintesis process employed can produce Li{sub 4}SiO{sub 4} in the form of pebbles, finally the best ion species for the electrical conduction is the Li{sup +} and is shown that the g-irradiation to a dose of 5MGy, facilitate its mobility through the creation of defects, without change in its conduction process.

  13. Engineering solutions for components facing the plasma in experimental fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Farfaletti-Casali, F.

    1986-07-01

    An analysis is made of the engineering problems related to the structures facing the plasma in experimental tokamak-type reactors. Attention is focused on the so-called ''current first wall'', i.e. the wall side of the blanket segments facing the plasma, and on the collector plates of the impurity control system. The design of a first wall, developed at the JRC-Ispra for INTOR/NET and based on the idea of conceiving it as one of the sides, of a box which envelopes a blanket segment, is described. The progress in the structural analysis of the first wall box under operating and abnormal (plasma disruption) conditions is presented and discussed. The design of the collector plates of the single-null divertor of INTOR/NET, as developed at the JRC-Ispra, is described. This design is based on a W-Re protective layer and a water-cooled heat sink, including cooling channels iun Cu-alloys and a Cu-matrix for bonding. The results of the elastic and elasto-plastic evaluations are discussed, together with a layout of the experimental activity in progress. It is concluded that, even if the uncertainties related to the plasma-wall interaction are still relevant, the engineering solutions identified look manageable, although they require a large research and development effort.

  14. Is Helium-3 hype, hyperbole or a hopeful fuel for the future. [Lunar He-3 extraction/production for earth fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mackowski, M.J.

    1989-08-01

    Sixty kilowatts of thermal power have been reached with deuterium/He-3 reaction on the JET reactor, and full scale study of the environmental impact of a tokamak D/He-3 reactor is now underway for NASA. He-3 is obtained from the decaying process undergone by tritium, but in nature, the source of He-3 is the sun. It is found in abundance in the lunar regolith. He-3 combines with deuterium in a fusion reaction generating very high amounts of energy, He-4 and protons. He-3 is economical; it could be moon mined and sold at a price comparable to oil. The energy released is roughly 70 percent efficient, and can be directly converted to electricity with solid-state converters. Also, reactors can be built cheaper, placed closer to cities, and maintained and decomissioned more easily than any other type of fission or fusion reactor, thus allowing faster commercialization and lower energy costs. He-3 is not very radioactive; however, the physics of its nuclear structure presents barriers to getting it to fuse. Other advantages of producing He-3 on the moon include obtaining gases necessary in moon colonies, and fuel for hydrogen rockets. Water, nickel and carbon-oxygen compounds can also be obtained that way.

  15. Proceedings of the Second Fusion-Fission Energy Systems Review Meeting

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-11-02

    The agenda of the meeting was developed to address, in turn, the following major areas: specific problem areas in nuclear energy systems for application of fusion-fission concepts; current and proposed fusion-fission programs in response to the identified problem areas; target costs and projected benefits associated with fusion-fission energy systems; and technical problems associated with the development of fusion-fission concepts. The greatest emphasis was placed on the characteristics of and problems, associated with fuel producing fusion-fission hybrid reactors.

  16. Proceedings of the Second Fusion-Fission Energy Systems Review Meeting

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-11-02

    The agenda of the meeting was developed to address, in turn, the following major areas: specific problem areas in nuclear energy systems for application of fusion-fission concepts; current and proposed fusion-fission programs in response to the identified problem areas; target costs and projected benefits associated with fusion-fission energy systems; and technical problems associated with the development of fusion-fission concepts. The greatest emphasis was placed on the characteristics of and problems, associated with fuel producing fusion-fission hybrid reactors.

  17. Application of carbon-aluminum nanostructures in divertor coatings from fusion reactor

    Science.gov (United States)

    Ciupina, V.; Lungu, C. P.; Vladoiu, R.; Epure, T. D.; Prodan, G.; Porosnicu, C.; Prodan, M.; Stanescu, I. M.; Contulov, M.; Mandes, A.; Dinca, V.; Zarovschi, V.

    2012-10-01

    Nanostructured carbon materials have increasingly attracted the interest of the scientific community, because of their fascinating physical properties and potential applications in high-tech devices. In the current ITER design, the tiles made of carbon fiber composites (CFCs) are foreseen for the strike point zone and tungsten (W) for other parts of the divertor region. This choice is a compromise based mainly on experience with individual materials in many different tokamaks. Also Carbon-Aluminum composites are the candidate material for the First Wall in ITER. In order to prepare nanostructured carbon-aluminum nanocomposite for the divertor part in fusion applications, the original method thermionic vacuum arc (TVA) was used in two electronic guns configuration. One of the main advantages of this technology is the bombardment of the growing thin film just by the ions of the depositing film. Moreover, the energy of ions can be controlled. Thermo-electrons emitted by an externally heated cathode and focused by a Wehnelt focusing cylinder are strongly accelerated towards the anode whose material is evaporated and bright plasma is ignited by a high voltage DC supply. The nanostructured C-Al films were characterized by Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM). Tribological properties in dry sliding were evaluated using a CSM ball-on-disc tribometer. The carbon - aluminum films were identified as a nanocrystals complex (from 2nm to 50 nm diameters) surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films. The friction coefficients (0.1 - 0.2, 0.5) of the C-Al coatings was decreased more than 2-5 times in comparison with the uncoated substrates proving excellent tribological properties. C-Al nanocomposites coatings were designed to have excellent tribological properties while the structure is composed by nanocrystals complex surrounded by amorphous structures

  18. Electron microscopy characterization of some carbon based nanostructures with application in divertors coatings from fusion reactor

    Science.gov (United States)

    Ciupina, V.; Morjan, I.; Lungu, C. P.; Vladoiu, R.; Prodan, G.; Prodan, M.; Zarovschi, V.; Porosnicu, C.; Stanescu, I. M.; Contulov, M.; Mandes, A.; Dinca, V.; Sugiyama, K.

    2011-10-01

    Nanostructured carbon materials have increasingly attracted the interest of the scientific community, because of their fascinating physical properties and potential applications in high-tech devices. In the current ITER design, the tiles made of carbon fiber composites (CFCs) are foreseen for the strike point zone and tungsten (W) for other parts of the divertor region. This choice is a compromise based mainly on experience with individual materials in many different tokamaks. Also Beryllium is the candidate material for the First Wall in ITER. In order to prepare nanostructured carbon-tungsten nanocomposite for the divertor part in fusion applications, the original method thermionic vacuum arc (TVA) was used in two electronic guns configuration. One of the main advantages of this technology is the bombardment of the growing thin film just by the ions of the depositing film. The nanostructured C-W and C-Be films were characterized by Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM) and Atomic Force Microscopy (AFM). The C-W films were identified as a nanocrystals complex (5 nm average diameter) surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films. The C-Be films are polycrystalline with mean grain size about 15 nm. The friction coefficients (0.15 - 0.35) of the C-W coatings was decreased more than 3-5 times in comparison with the uncoated substrates proving excellent tribological properties. C-W nanocomposites coatings were designed to have excellent tribological properties while the structure is composed by nanocrystals complex surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films.&updat

  19. Coupled anaerobic-aerobic treatment of whey wastewater in a sequencing batch reactor: proof of concept.

    Science.gov (United States)

    Frigon, J C; Bruneau, T; Moletta, R; Guiot, S R

    2007-01-01

    A proof of concept was performed in order to verify if the coupling of anaerobic and aerobic conditions inside the same digester could efficiently treat a reconstituted whey wastewater at 21 degrees C. The sequencing batch reactor (SBR) cycles combined initial anaerobic phase and final aerobic phase with reduced aeration. A series of 24 h cycles in 0.5 L digesters, with four different levels of oxygenation (none, 54, 108 and 182 mgO2 per gram of chemical oxygen demand (COD)), showed residual soluble chemical oxygen demand (sCOD) of 683 +/- 46, 720 +/- 33, 581 +/- 45, 1239 +/- 15 mg L(-1), respectively. Acetate and hydrogen specific activities were maintained for the anaerobic digester, but decreased by 10-25% for the acetate and by 20-50% for the hydrogen, in the coupled digesters. The experiment was repeated using 48 h cycles with limited aeration during 6 or 16 hours at 54 and 108 mgO2gCODinitial(-1), displaying residual sCOD of 177 +/- 43, 137 +/- 38, 104 +/- 22 and 112 +/- 9 mgL(-1) for the anaerobic and the coupled digesters, respectively. The coupled digesters recovered after a pH shock with residual sCOD as low as 132 mg L(-1) compared to 636 mg L(-1) for the anaerobic digester. With regard to the obtained results, the feasibility of the anaerobic-aerobic coupling in SBR digesters for the treatment of whey wastewater was demonstrated.

  20. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  1. Modelling of turbulent hydrocarbon combustion. Test of different reactor concepts for describing the interactions between turbulence and chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Kremer, H. [Ruhr-Universitaet Bochum, Lehrstuhl fuer Energieanlagentechnik, Bochum (Germany); Kilpinen, P.; Hupa, M. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group

    1997-12-31

    The detailed modelling of turbulent reactive flows with CFD-codes is a major challenge in combustion science. One method of combining highly developed turbulence models and detailed chemistry in CFD-codes is the application of reactor based turbulence chemistry interaction models. In this work the influence of different reactor concepts on methane and NO{sub x} chemistry in turbulent reactive flows was investigated. Besides the classical reactor approaches, a plug flow reactor (PFR) and a perfectly stirred reactor (PSR), the Eddy-Dissipation Combustion Model (EDX) and the Eddy Dissipation Concept (EDC) were included. Based on a detailed reaction scheme and a simplified 2-step mechanism studies were performed in a simplified computational grid consisting of 5 cells. The investigations cover a temperature range from 1273 K to 1673 K and consider fuel-rich and fuel-lean gas mixtures as well as turbulent and highly turbulent flow conditions. All test cases investigated in this study showed a strong influence of the reactor residence time on the species conversion processes. Due to this characteristic strong deviations were found for the species trends resulting from the different reactor approaches. However, this influence was only concentrated on the `near burner region` and after 4-5 cells hardly any deviation and residence time dependence could be found. The importance of the residence time dependence increased when the species conversion was accelerated as it is the case for overstoichiometric combustion conditions and increased temperatures. The study focused furthermore on the fine structure in the EDC. Unlike the classical approach this part of the cell was modelled as a PFR instead of a PSR. For high temperature conditions there was hardly any difference between both reactor types. However, decreasing the temperature led to obvious deviations. Finally, the effect of the selective species transport between the cells on the conversion process was investigated

  2. Application of the LBB concept to nuclear power plants with WWER 440 and WWER 1000 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L. [Nuclear Research Institute Rez (Czech Republic)

    1997-04-01

    Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors.

  3. Fusion neutronics

    CERN Document Server

    Wu, Yican

    2017-01-01

    This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...

  4. SABR Fusion-Fission Hybrid Studies

    Science.gov (United States)

    Stewart, Chris

    2012-03-01

    The Subcritical Advanced Burner Reactor (SABR) concept is a fast reactor comprised of a tokamak fusion neutron source based on ITER surrounded by an annular fission core adapted from Integral Fast Reactor designs. Previous work has examined SABR used to help close the nuclear fuel cycle by fissioning the transuranics from spent nuclear fuel. One focus of the present work is a SABR Breeder Reactor to achieve tritium self-sufficieny and a Pu breeding ratio significantly above 1 in order to provide fuel for SABR as well as for MOX-fueled LWR's and other fast reactors. Another focus of this research is the dynamic safety simulation of lloss-of-flow loss-of-heat-sink, loss-of-power, and positive reactivity accidents in the TRU fuel SABR burner reactor. The reactivity effect of thermal-induced bowing of fuel pins has been modeled, which is expected to provide passive safety.

  5. Recent developments concerning the fusion; Developpements recents sur la fusion

    Energy Technology Data Exchange (ETDEWEB)

    Jacquinot, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint Paul lez Durance (France); Andre, M. [CEA/DAM Ile de France, 91 - Bruyeres Le Chatel (France); Aymar, R. [ITER Joint Central Team Garching, Muenchen (Germany)] [and others

    2000-09-04

    Organized the 9 march 2000 by the SFEN, this meeting on the european program concerning the fusion, showed the utility of the exploitation and the enhancement of the actual technology (JET, Tore Supra, ASDEX) and the importance of the Europe engagement in the ITER program. The physical stakes for the magnetic fusion have been developed with a presentation of the progresses in the knowledge of the stability limits. A paper on the inertial fusion was based on the LMJ (Laser MegaJoule) project. The two blanket concepts chosen in the scope of the european program on the tritium blankets, have been discussed. These concepts will be validated by irradiation tests in the ITER-FEAT and adapted for a future reactor. (A.L.B.)

  6. Optimization and simplification of the concept of non-moderated Thorium Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, Elsa; Heuer, Daniel; Allibert, Michel; Doligez, Xavier; Ghetta, Veronique; Le Brun, Christian [LPSC-IN2P3-CNRS/UJF/INPG, LPSC 53 avenue des Martyrs, 38026 Grenoble Cedex (France)

    2008-07-01

    Molten salt reactors, in the configuration presented here and called Thorium Molten Salt Reactor (TMSR), are particularly well suited to fulfil the criteria defined by the Generation IV forum, and may be operated in simplified and safe conditions in the Th/{sup 233}U fuel cycle with fluoride salts. The characteristics of the non-moderated TMSR based on a fast neutron spectrum are detailed in this paper: we aimed at designing an optimised TMSR with the simplest configuration. Using a simple kinetic-point model, we analyze the reactor's transient as the total reactivity margins are introduced in the core. We thus confirm, beyond the classical advantages of molten salt reactors, the satisfactory behaviour of the TMSR in terms of safety and the excellent level of stability which can be achieved in such reactors. (authors)

  7. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    Science.gov (United States)

    Fletcher, C. D.

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes.

  8. Nuclear Fusion Research Understanding Plasma-Surface Interactions

    CERN Document Server

    Clark, Robert E.H

    2005-01-01

    It became clear in the early days of fusion research that the effects of the containment vessel (erosion of "impurities") degrade the overall fusion plasma performance. Progress in controlled nuclear fusion research over the last decade has led to magnetically confined plasmas that, in turn, are sufficiently powerful to damage the vessel structures over its lifetime. This book reviews current understanding and concepts to deal with this remaining critical design issue for fusion reactors. It reviews both progress and open questions, largely in terms of available and sought-after plasma-surface interaction data and atomic/molecular data related to these "plasma edge" issues.

  9. Compact magnetic confinement fusion: Spherical torus and compact torus

    Directory of Open Access Journals (Sweden)

    Zhe Gao

    2016-05-01

    Full Text Available The spherical torus (ST and compact torus (CT are two kinds of alternative magnetic confinement fusion concepts with compact geometry. The ST is actually a sub-category of tokamak with a low aspect ratio; while the CT is a toroidal magnetic configuration with a simply-connected geometry including spheromak and field reversed pinch. The ST and CT have potential advantages for ultimate fusion reactor; while at present they can also provide unique fusion science and technology contributions for mainstream fusion research. However, some critical scientific and technology issues should be extensively investigated.

  10. Optical transmittance investigation of 1-keV ion-irradiated sapphire crystals as potential VUV to NIR window materials of fusion reactors

    Directory of Open Access Journals (Sweden)

    Keisuke Iwano

    2016-10-01

    Full Text Available We investigate the optical transmittances of ion-irradiated sapphire crystals as potential vacuum ultraviolet (VUV to near-infrared (NIR window materials of fusion reactors. Under potential conditions in fusion reactors, sapphire crystals are irradiated with hydrogen (H, deuterium (D, and helium (He ions with 1-keV energy and ∼ 1020-m-2 s-1 flux. Ion irradiation decreases the transmittances from 140 to 260 nm but hardly affects the transmittances from 300 to 1500 nm. H-ion and D-ion irradiation causes optical absorptions near 210 and 260 nm associated with an F-center and an F+-center, respectively. These F-type centers are classified as Schottky defects that can be removed through annealing above 1000 K. In contrast, He-ion irradiation does not cause optical absorptions above 200 nm because He-ions cannot be incorporated in the crystal lattice due to the large ionic radius of He-ions. Moreover, the significant decrease in transmittance of the ion-irradiated sapphire crystals from 140 to 180 nm is related to the light scattering on the crystal surface. Similar to diamond polishing, ion irradiation modifies the crystal surface thereby affecting the optical properties especially at shorter wavelengths. Although the transmittances in the VUV wavelengths decrease after ion irradiation, the transmittances can be improved through annealing above 1000 K. With an optical transmittance in the VUV region that can recover through simple annealing and with a high transparency from the ultraviolet (UV to the NIR region, sapphire crystals can therefore be used as good optical windows inside modern fusion power reactors in terms of light particle loadings of hydrogen isotopes and helium.

  11. Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

    Science.gov (United States)

    Martin, James J.; Reid, Robert S.

    2004-01-01

    A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  13. Design of a Fast Neutral He Beam System for Feasibility Study of Charge-Exchange Alpha-Particle Diagnostics in a Thermonuclear Fusion Reactor

    CERN Document Server

    Shinto, Katsuhiro; Kitajima, Sumio; Kiyama, Satoru; Nishiura, Masaki; Sasao, Mamiko; Sugawara, Hiroshi; Takenaga, Mahoko; Takeuchi, Shu; Wada, Motoi

    2005-01-01

    For alpha-particle diagnostics in a thermonuclear fusion reactor, neutralization using a fast (~2 MeV) neutral He beam produced by the spontaneous electron detachment of a He- is considered most promising. However, the beam transport of produced fast neutral He has not been studied, because of difficulty for producing high-brightness He- beam. Double-charge-exchange He- sources and simple beam transport systems were developed and their results were reported in the PAC99* and other papers.** To accelerate an intense He- beam and verify the production of the fast neutral He beam, a new test stand has been designed. It consists of a multi-cusp He+

  14. Development of a real-time simulation tool towards self-consistent scenario of plasma start-up and sustainment on helical fusion reactor FFHR-d1

    Science.gov (United States)

    Goto, T.; Miyazawa, J.; Sakamoto, R.; Suzuki, Y.; Suzuki, C.; Seki, R.; Satake, S.; Huang, B.; Nunami, M.; Yokoyama, M.; Sagara, A.; the FFHR Design Group

    2017-06-01

    This study closely investigates the plasma operation scenario for the LHD-type helical reactor FFHR-d1 in view of MHD equilibrium/stability, neoclassical transport, alpha energy loss and impurity effect. In 1D calculation code that reproduces the typical pellet discharges in LHD experiments, we identify a self-consistent solution of the plasma operation scenario which achieves steady-state sustainment of the burning plasma with a fusion gain of Q ~ 10 was found within the operation regime that has been already confirmed in LHD experiment. The developed calculation tool enables systematic analysis of the operation regime in real time.

  15. Towards standardization of the dissemination measures and tritium solubility in materials of fusion reactors; Hacia la estandarizacion de las medidas de difusion y solubilidad de tritio en materiales de reactores de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Alberto, G.; Penalva, I.; Aranburu, I.; Sarrionandia-Ibarra, A.; Legarda, F.; Martinez, P. M.; Sedano, L.; Moral, N.

    2011-07-01

    The standardization of the measurements of hydrogen isotope interaction with different materials is a challenge and goal of fusion technology programs worldwide. For decades the programs have promoted the need for a reference laboratory for measurements of hydrogen transport to the evolution of fusion technology, but that goal is still pending, in contrast to the situation in other goals I+D.

  16. 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).

  17. Thermal-Fatigue Analysis of W-joined Ferritic-Martensitic Steel Mockup for Fusion Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon; Park, Seong Dae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shin, Kyu In [Gentec Co., Daejeon (Korea, Republic of); Moon, Se Yeon; Hong, Bong Guen [Chonbuk National University, Chonbuk (Korea, Republic of)

    2015-10-15

    Through the ITER blanket first wall (BFW) development project in Korea, the joining methods were developed with a beryllium (Be) layer as a plasma-facing material, a copper alloy (CuCrZr) layer as a heat sink, and type 316L austenitic stainless steel (SS316L) as a structural material. And joining methods were developed such as Be as an armor and FMS as a structural material, or W as an armor and FMS as a structural material were developed through the test blanket module (TBM) program. As a candidate of PFC for DEMO, W/FMS joining methods have been developed and a new Ti interlayer was applied differently from the previous work. In the present study, the W/FMS PFC development was introduced with the following procedure to apply to the PFCs for a fusion reactor: (1) Three W/FMS mockups were fabricated using the developed HIP followed by a post-HIP heat treatment (PHHT). (2) Because the High Heat Flux (HHF) test should be performed over the thermal lifetime of the mockup under the proper test conditions to confirm the joint's integrity, the test conditions were determined through a preliminary analysis. In this study, commercial ANSYS-CFX for thermalhydraulic analysis and ANSYS-mechanical for the thermo-mechanical analysis are used to evaluate the thermal-lifetime of the mockup to determine the test conditions. Also, the Korea Heat Load Test facility with an Electron Beam (KoHLT-EB) will be used and its water cooling system is considered to perform the thermal-hydraulic analysis especially for considering the two-phase analysis with a higher heat flux conditions. From the analysis, the heating and the cooling conditions were determined for 0.5- and 1.0-MW/m{sup 2} heat fluxes, respectively. Elastic-plastic analysis is performed to determine the lifetime and finally, the 1.0 MW/m{sup 2} heat flux conditions are determined up to 4,306 cycles. The test will be done in the near future and the measured temperatures will be compared with the present simulation results.

  18. Developing maintainability for fusion power systems. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Zahn, H.S.; Mantz, H.C.; Curtis, C.T.; Buchheit, R.J.; Green, W.M.; Zuckerman, D.S.

    1979-11-01

    The overall purpose of the study is to identify design features of fusion power reactors which contribute to the achievement of high levels of maintainability. Previous phases evaluated several commercial tokamak reactor design concepts. This final phase compares the maintainability of a tandem mirror reactor (TMR) commercial conceptual design with the most maintainable tokamak concept selected from earlier work. A series of maintainability design guidelines and desirable TMR design features are defined. The effects of scheduled and unscheduled maintenance for most of the reactor subsystems are defined. The comparison of the TMR and tokamak reactor maintenance costs and availabilities show that both reactors have similar costs for scheduled maintenance at 19.4 and 20.8 million dollars annually and similar scheduled downtime availability impacts, achieving approximate availabilities of 79% at optimized maintenance intervals and cost of electricity.

  19. Scrap tyre recycling process with molten zinc as direct heat transfer and solids separation fluid: A new reactor concept.

    Science.gov (United States)

    Riedewald, Frank; Goode, Kieran; Sexton, Aidan; Sousa-Gallagher, Maria J

    2016-01-01

    Every year about 1.5 billion tyres are discarded worldwide representing a large amount of solid waste, but also a largely untapped source of raw materials. The objective of the method was to prove the concept of a novel scrap tyre recycling process which uses molten zinc as the direct heat transfer fluid and, simultaneously, uses this media to separate the solids products (i.e. steel and rCB) in a sink-float separation at an operating temperature of 450-470 °C. This methodology involved: •construction of the laboratory scale batch reactor,•separation of floating rCB from the zinc,•recovery of the steel from the bottom of the reactor following pyrolysis.

  20. Space Molten Salt Reactor Concept for Nuclear Electric Propulsion and Surface Power

    Science.gov (United States)

    Eades, M.; Flanders, J.; McMurray, N.; Denning, R.; Sun, X.; Windl, W.; Blue, T.

    Students at The Ohio State University working under the NASA Steckler Grant sought to investigate how molten salt reactors with fissile material dissolved in a liquid fuel medium can be applied to space applications. Molten salt reactors of this kind, built for non-space applications, have demonstrated high power densities, high temperature operation without pressurization, high fuel burn up and other characteristics that are ideal for space fission systems. However, little research has been published on the application of molten salt reactor technology to space fission systems. This paper presents a conceptual design of the Space Molten Salt Reactor (SMSR), which utilizes molten salt reactor technology for Nuclear Electric Propulsion (NEP) and surface power at the 100 kWe to 15 MWe level. Central to the SMSR design is a liquid mixture of LiF, BeF2 and highly enriched U235F4 that acts as both fuel and core coolant. In brief, some of the positive characteristics of the SMSR are compact size, simplified core design, high fuel burn up percentages, proliferation resistant features, passive safety mechanisms, a considerable body of previous research, and the possibility for flexible mission architecture.

  1. Manufacturing W fibre-reinforced Cu composite pipes for application as heat sink in divertor targets of future nuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Alexander v.; You, Jeong-Ha [Max-Planck-Institut fuer Plasmaphysik, 85748 Garching (Germany); Ewert, Dagmar [Institut fuer Textil- und Verfahrenstechnik Denkendorf, 73770 Denkendorf (Germany); Siefken, Udo [Louis Renner GmbH, 85221 Dachau (Germany)

    2016-07-01

    An important plasma-facing component (PFC) in future nuclear fusion reactors is the so-called divertor which allows power exhaust and removal of impurities from the main plasma. The most highly loaded parts of a divertor are the target plates which have to withstand intense particle bombardment. This intense particle bombardment leads to high heat fluxes onto the target plates which in turn lead to severe thermomechanical loads. With regard to future nuclear fusion reactors, an improvement of the performance of divertor targets is desirable in order to ensure reliable long term operation of such PFCs. The performance of a divertor target is most closely linked to the properties of the materials that are used for its design. W fibre-reinforced Cu (Wf/Cu) composites are regarded as promising heat sink materials in this respect. These materials do not only feature adequate thermophysical and mechanical properties, they do also offer metallurgical flexibility as their microstructure and hence their macroscopic properties can be tailored. The contribution will point out how Wf/Cu composites can be used to realise an advanced design of a divertor target and how these materials can be fabricated by means of liquid Cu infiltration.

  2. Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O2 cannot be ignored, especially for the FHR, in which environment the product, SiO2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

  3. Capabilities of the ATHENA computer code for modeling the SP-100 space reactor concept

    Science.gov (United States)

    Fletcher, C. D.

    1985-09-01

    The capability to perform thermal-hydraulic analyses of an SP-100 space reactor was demonstrated using the ATHENA computer code. The preliminary General Electric SP-100 design was modeled using Athena. The model simulates the fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of this design. Two ATHENA demonstration calculations were performed simulating accident scenarios. A mask for the SP-100 model and an interface with the Nuclear Plant Analyzer (NPA) were developed, allowing a graphic display of the calculated results on the NPA.

  4. Porous Photocatalytic Membrane Microreactor (P2M2): A new reactor concept for photochemistry

    NARCIS (Netherlands)

    Aran, H.C.; Salamon, D.; Rijnaarts, T.; Mul, G.; Wessling, M.; Lammertink, R.G.H.

    2011-01-01

    In this study, a new membrane microreactor concept for multiphase photocatalytic reactions is demonstrated. Microfabrication, photocatalyst immobilization and surface modification steps were performed to develop a Porous Photocatalytic Membrane Microreactor (P2M2). This concept benefits from a stabl

  5. Porous Photocatalytic Membrane Microreactor (P2M2): A new reactor concept for photochemistry

    NARCIS (Netherlands)

    Aran, H.C.; Salamon, David; Rijnaarts, Timon; Rijnaarts, T.; Mul, Guido; Wessling, Matthias; Lammertink, Rob G.H.

    2011-01-01

    In this study, a new membrane microreactor concept for multiphase photocatalytic reactions is demonstrated. Microfabrication, photocatalyst immobilization and surface modification steps were performed to develop a Porous Photocatalytic Membrane Microreactor (P2M2). This concept benefits from a

  6. Realizing "2001: A Space Odyssey": Piloted Spherical Torus Nuclear Fusion Propulsion

    Science.gov (United States)

    Williams, Craig H.; Dudzinski, Leonard A.; Borowski, Stanley K.; Juhasz, Albert J.

    2005-01-01

    A conceptual vehicle design enabling fast, piloted outer solar system travel was created predicated on a small aspect ratio spherical torus nuclear fusion reactor. The initial requirements were satisfied by the vehicle concept, which could deliver a 172 mt crew payload from Earth to Jupiter rendezvous in 118 days, with an initial mass in low Earth orbit of 1,690 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including artificial gravity payload, central truss, nuclear fusion reactor, power conversion, magnetic nozzle, fast wave plasma heating, tankage, fuel pellet injector, startup/re-start fission reactor and battery bank, refrigeration, reaction control, communications, mission design, and space operations. Detailed fusion reactor design included analysis of plasma characteristics, power balance/utilization, first wall, toroidal field coils, heat transfer, and neutron/x-ray radiation. Technical comparisons are made between the vehicle concept and the interplanetary spacecraft depicted in the motion picture 2001: A Space Odyssey.

  7. Helias reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Beidler, C.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Grieger, G. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Harmeyer, E. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kisslinger, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Karulin, N. [Nuclear Fusion Institute, Moscow (Russian Federation); Maurer, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany); Nuehrenberg, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rau, F. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Sapper, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Wobig, H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1995-10-01

    The present status of Helias reactor studies is characterised by the identification and investigation of specific issues which result from the particular properties of this type of stellarator. On the technical side these are issues related to the coil system, while physics studies have concentrated on confinement, alpha-particle behaviour and ignition conditions. The usual assumptions have been made in those fields which are common to all toroidal fusion reactors: blanket and shield, refuelling and exhaust, safety and economic aspects. For blanket and shield sufficient space has been provided, a detailed concept will be developed in future. To date more emphasis has been placed on scoping and parameter studies as opposed to fixing a specific set of parameters and providing a detailed point study. One result of the Helias reactor studies is that physical dimensions are on the same order as those of tokamak reactors. However, it should be noticed that this comparison is difficult in view of the large spectrum of tokamak reactors ranging from a small reactor like Aries, to a large device such as SEAFP. The notion that the large aspect ratio of 10 or more in Helias configurations also leads to large reactors is misleading, since the large major radius of 22 m is compensated by the average plasma radius of 1.8 m and the average coil radius of 5 m. The plasma volume of 1400 m{sup 3} is about the same as the ITER reactor and the magnetic energy of the coil system is about the same or even slightly smaller than envisaged in ITER. (orig.)

  8. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  9. Power Flattening and Rejuvenation of PWR Spent Fuel Blanket for Hybrid Fusion-Fission Reactor%功率展平的压水堆乏燃料发电包层中子学初步研究

    Institute of Scientific and Technical Information of China (English)

    马续波; 陈义学; 王继亮; 王悦; 韩静茹; 陆道纲

    2011-01-01

    The hybrid fusion-fission reactor has advantages of breeding of the nuclear fuel and transmutation of the long-life nuclear waste and having inherent safety. Meanwhile, the engineering and technological demand of hybrid reactor is significantly reduced comparing with that of pure fusion reactor. A generating electricity blanket concept using the PWR spent fuel directly was proposed, which was based on ITER parameter level achieved. Different volume fractions of the fuel in blanket enabled to realize a power flattening in the fissile zone. The results show that the peak-to-average power factor becomes less than no power flattening, and the output power of the fuel zone raises more than 21. 7%. At the end of the operation, the maximum fuel enrichment is 5. 23%. The blanket is feasible from the neutronics viewpoint.%聚变裂变混合堆在增殖核燃料、嬗变长寿命核废料及固有安全性等方面具有较大优势,同时,它比纯聚变堆在工程及技术方面要求低,因此较聚变堆更易实现.本工作基于目前国际聚变实验堆(ITER)所能达到的技术水平,提出一种直接利用乏燃料进行发电的聚变裂变混合堆包层概念,利用在不同位置放置不同乏燃料体积分数的方法对燃料增殖区实现了功率展平.计算结果表明:功率展平后的包层功率不均匀系数更小,且包层中燃料区的能量输出要比不展平情况下的能量输出高约21.7%.燃料富集度到运行末期最大可达5.23%.从中子学角度初步论证了该包层的可行性.

  10. Burn Control of Magnetically Confined Fusion Plasma 2. Burn Control in Tokamak Fusion Reactors 2.3 Burn Control in ITER

    Science.gov (United States)

    Fujisawa, Noboru

    The issue of burn control in FDR-ITER, the design of which was completed in 1998, is introduced, Controllability was studied based on the ID transport code, PRETOR, during the burn phase with self-ignition, as well as during start-up and shut-down. The results of the present study have helped us to identify the importance of controlling the fuel supply, impurity injection, and heating power to maintain fusion power and power to the divertor.

  11. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  12. Plasma physics for controlled fusion

    CERN Document Server

    Miyamoto, Kenro

    2016-01-01

    This new edition presents the essential theoretical and analytical methods needed to understand the recent fusion research of tokamak and alternate approaches. The author describes magnetohydrodynamic and kinetic theories of cold and hot plasmas in detail. The book covers new important topics for fusion studies such as plasma transport by drift turbulence, which depend on the magnetic configuration and zonal flows. These are universal phenomena of microturbulence. They can modify the onset criterion for turbulent transport, instabilities driven by energetic particles as well as alpha particle generation and typical plasma models for computer simulation. The fusion research of tokamaks with various new versions of H modes are explained. The design concept of ITER, the international tokamak experimental reactor, is described for inductively driven operations as well as steady-state operations using non-inductive drives. Alternative approaches of reversed-field pinch and its relaxation process, stellator includi...

  13. Silicon carbide as an advanced material for long-term fusion reactor development programs; Il Carburo di silicio come materiale avanzato per i programmi long-term relativi ai futuri reattori a fusione nucleare: Review sulle tecnologie applicative e sullo stato della ricerca

    Energy Technology Data Exchange (ETDEWEB)

    Donato, A.

    1991-10-01

    Silicon carbide is being considered a possible candidate low activation material for the construction of important components (e.g., first wall coating, limitors, divertors and beam stops) of future commercial fusion reactors which will require excellent thermal and mechanical behaviour in high temperature environments and very high irradiation fields. In this paper a literature review covering the last ten years (up to 1990) is reported. It concerns silicon carbide properties, its relative manufacturing technologies and technological developments. One hundred and twenty-seven papers published in scientific magazines were selected and reviewed, out of 400 concerning silicon carbide in general. Several important analogies and possible industrial synergies for its use both in fusion reactors and in aerospace devices were found. The material as produced today, the design methodology and the technology are relatively new and seem to need further development. Moreover, extensive research work is necessary to better evaluate silicon carbide behaviour in fusion reactor operating conditions.

  14. The Long way Towards Inertial Fusion Energy (lirpp Vol. 13)

    Science.gov (United States)

    Velarde, Guillermo

    2016-10-01

    In 1955 the first Geneva Conference was held in which two important events took place. Firstly, the announcement by President Eisenhower of the Program Atoms for Peace declassifying the information concerning nuclear fission reactors. Secondly, it was forecast that due to the research made on stellerators and magnetic mirrors, the first demo fusion facility would be in operation within ten years. This forecasting, as all of us know today, was a mistake. Forty years afterwards, we can say that probably the first Demo Reactor will be operative in some years more and I sincerely hope that it will be based on the inertial fusion concept...

  15. Analysis of dashpot performance for rotating control drums of a lithium cooled fast reactor concept

    Science.gov (United States)

    Wenzler, C. J.

    1972-01-01

    A dashpot was incorporated in the design of the drive train of the rotating control drum to prevent shock damage to the control drum and drive train at the termination of a scram action. A rotating vane dashpot using reactor coolant lithium as a damping fluid appears to be the best candidate of the various damping devices explored. A performance analysis, results and discussion of vane type dashpots are presented.

  16. Hybrid reactor based on combined cavitation and ozonation: from concept to practical reality.

    Science.gov (United States)

    Gogate, P R; Mededovic-Thagard, S; McGuire, D; Chapas, G; Blackmon, J; Cathey, R

    2014-03-01

    The present work gives an in depth discussion related to the development of a hybrid advanced oxidation reactor, which can be effectively used for the treatment of various types of water. The reactor is based on the principle of intensifying degradation/disinfection using a combination of hydrodynamic cavitation, acoustic cavitation, ozone injection and electrochemical oxidation/precipitation. Theoretical studies have been presented to highlight the uniform distribution of the cavitational activity and enhanced generation of hydroxyl radicals in the cavitation zone, as well as higher turbulence in the main reactor zone. The combination of these different oxidation technologies have been shown to result in enhanced water treatment ability, which can be attributed to the enhanced generation of hydroxyl radicals, enhanced contact of ozone and contaminants, and the elimination of mass transfer resistances during electrochemical oxidation/precipitation. Compared to the use of individual approaches, the hybrid reactor is expected to intensify the treatment process by 5-20 times, depending on the application in question, which can be confirmed based on the literature illustrations. Also, the use of Ozonix® has been successfully proven while processing recycled fluids at commercial sites on over 750 oil and natural gas wells during hydraulic operations around the United States. The superiority of the hybrid process over conventional chemical treatments in terms of bacteria and scale reduction as well as increased water flowability and better chemical compatibility, which is a key requirement for oil and gas applications, has been established. Copyright © 2013 Elsevier B.V. All rights reserved.

  17. Parameter space for the collective laser coupling in the laser fusion driver based on the concept of fiber amplification network.

    Science.gov (United States)

    Huang, Zhihua; Lin, Honghuan; Xu, Dangpeng; Li, Mingzhong; Wang, Jianjun; Deng, Ying; Zhang, Rui; Zhang, Yongliang; Tian, Xiaocheng; Wei, Xiaofeng

    2013-07-15

    Collective laser coupling of the fiber array in the inertial confinement fusion (ICF) laser driver based on the concept of fiber amplification network (FAN) is researched. The feasible parameter space is given for laser coupling of the fundamental, second and third harmonic waves by neglecting the influence of the frequency conversion on the beam quality under the assumption of beam quality factor conservation. Third harmonic laser coupling is preferred due to its lower output energy requirement from a single fiber amplifier. For coplanar fiber array, the energy requirement is around 0.4 J with an effective mode field diameter of around 500 μm while maintaining the fundamental mode operation which is more than one order of magnitude higher than what can be achieved with state-of-the-art technology. Novel waveguide structure needs to be developed to enlarge the fundamental mode size while mitigating the catastrophic self-focusing effect.

  18. Concept of a BNCT line with in-pool fission converter at MARIA reactor in Swierk

    Science.gov (United States)

    Pytel, Krzysztof; Andrzejewski, Krzysztof; Golnik, Natalia; Osko, Jakub

    2009-01-01

    BNCT facility in the Institute of Atomic Energy in Otwock-Swierk is under construction at the horizontal channel H2 of the research reactor MARIA. Measurements of the neutron energy spectrum performed at the front of the H2 experimental channel, have shown that flux of epithermal neutrons (above 10 keV) at the BNCT irradiation port was below 109 n cm-2 s-1 i.e. it was too low to be directly used for the BNCT treatment. Therefore, a fission converter will be placed between the reactor core and the periphery of the graphite reflector of MARIA reactor. The uranium converter will be powered by the densely packed EK-10 fuel elements with 10% enrichment. Preliminary calculations have shown that the total neutron flux in the converter will be about 1013 n cm-2 s-1 and flux of epithermal neutrons at the entrance to the filter/moderator of the beam will be about 2·1013 n cm-2 s-1.

  19. A new concept of high flow rate non-thermal plasma reactor for air treatment

    Energy Technology Data Exchange (ETDEWEB)

    Goujard, V.; Tatibouet, J.M. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique

    2010-07-01

    Although several non-thermal plasma reactors have been tested for air treatment at the laboratory scale, up-scaling to pilot or industrial scale remains a challenge because several parameters must be considered, such as hydrodynamic behaviour, maximum voltage in an industrial environment, and maintenance of the system. This paper presented a newly developed reactor which consists to a DBD plasma generated on individual supports that could be directly inserted in gas pipes where air flow must be treated. Elimination of 40 percent of 15 ppm of propene was obtained with a energy density as low as 10 J/L. The propene conversion increased when a manganese oxide based catalyst was used because the ozone produced by the plasma was used as an as an oxidant. A simple model of the plasma-catalyst reactor behaviour showed that more than 90 percent of propene conversion can be expected for an input energy density of 10 J/L and residual ozone concentration less than 100 ppb.

  20. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW{sub th} and an electricity generation mode of 100 kW{sub th}, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and