Science.gov

Sample records for advanced fusion reactor

  1. Advanced Fusion Reactors for Space Propulsion and Power Systems

    SciTech Connect

    Chapman, John J.

    2011-06-15

    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.

  2. Advanced Fusion Reactors for Space Propulsion and Power Systems

    NASA Technical Reports Server (NTRS)

    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. Radioactive waste produced by DEMO and commercial fusion reactors extrapolated from ITER and advanced data bases

    SciTech Connect

    Stacey, W.M.; Hertel, N.E.; Hoffman, E.A.

    1994-07-01

    The radioactive wastes that would be produced in demonstration (DEMO) and commercial (CFR) fusion reactors which could be extrapolated from the design data base that will be provided by ITER and its supporting R&D and from a design data base supplemented by advanced physics and advanced materials R&D programs are identified and characterized in terms of a number of possible criteria for near-surface burial. The results indicate that there is a possibility that all fusion wastes could satisfy a ``low level`` waste criterion for ``near-surface`` burial.

  4. Reactor applications of the Compact Fusion Advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

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

    1988-03-01

    A preliminary design of a D-T fusion reactor blanket and MHD power conversion system is made based on the CFAR concept, and it was found that 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 boiling 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 temperatures, and only a relatively small natural-draft heat exchanger 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 a cost analysis has not yet been performed, preliminary cost estimates indicate low capital costs and a very attractive cost of electricity.

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

    SciTech Connect

    Tsai, H.; Gomes, I.C.; Smith, D.L.; Palmer, A.J.; Ingram, F.W.; Wiffen, F.W.

    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.

  6. Advanced fueling system for steady-state operation of a fusion reactor

    SciTech Connect

    Raman, R.

    2008-07-15

    Steady-state Advanced Tokamak scenarios rely on optimized density and pressure profiles to maximize the bootstrap current fraction. Under this mode of operation, the fuelling system must deposit small amounts of fuel where it is needed, and as often as needed, so as to compensate for fuel losses, but not to adversely alter the established density and pressure profiles. A precision fuelling system has the capability for controlling the fusion burn by maintaining the required pressure profile to maximize the bootstrap current fraction. An advanced fuelling system based on Compact Toroid (CT) injection has the potential to meet these needs while simultaneously simplifying the requirements of the tritium handling systems. Simpler engineering systems would reduce reactor construction and maintenance cost through increased reliability. A CT fueling system is described together with the associated tritium handling requirements. (authors)

  7. Radioactive waste produced by DEMO and commerical fusion reactors extrapolated from ITER and advanced data bases

    SciTech Connect

    Stacey, W.M.; Hertel, N.E.; Hoffman, E.A.

    1994-12-31

    The potential for providing energy with minimal environmental impact is a powerful motivation for the development of fusion and is the long-term objective of most fusion programs. However, the societal acceptability of magnetic fusion may well be decided in the near-term when decisions are taken on the construction of DEMO to follow ITER (if not when the construction decision is taken on ITER). Component wastes were calculated for DEMOs based on each data base by first calculating reactor sizes needed to satisfy the physics, stress and radiation attenuation requirements, and then calculating component replacement rates based on radiation damage and erosion limits. Then, radioactive inventories were calculated and compared to a number of international criteria for {open_quote}near-surface{close_quote} burial. None of the components in either type of design would meet the Japanese LLW criterion (<1 Ci/m{sup 3}) within 10 years of shutdown, although the advanced (V/Li) blanket would do so soon afterwards. The vanadium first wall, divertor and blanket would satisfy the IAEA LLW criterion (<2 mSv/h contact dose) within about 10 years after shutdown, but none of the stainless steel or copper components would. All the components in the advanced data base designs except the stainless steel vacuum vessel and shield readily satisfy the US extended 10CFR61 intruder dose criterion, but none of the components in the {open_quotes}ITER data base{close_quotes} designs do so. It seems unlikely that a stainless steel first wall or a copper divertor plate could satisfy the US (class C) criterion for near surface burial, much less the more stringent international, criteria. On the other hand, the first wall, divertor and blanket of the V/Li system would still satisfy the intruder dose concentration limits even if the dose criterion was reduced by two orders of magnitude.

  8. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    NASA Astrophysics Data System (ADS)

    Velarde, G.; Perlado, J. M.; Alonso, E.; Alonso, M.; Domínguez, E.; Rubiano, J. G.; Gil, J. M.; Gómez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martínez-Val, J. M.; Mínguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P.

    2001-05-01

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT x fuel with a small tritium initial content ( x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures (≫100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower Te and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in Si

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

    NASA Astrophysics Data System (ADS)

    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.

  10. Spherical torus fusion reactor

    DOEpatents

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  11. Prospects for Tokamak Fusion Reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.

    1995-04-01

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

  12. Prospects for toroidal fusion reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.D.

    1994-06-01

    Work on the International Thermonuclear Experimental Reactor (ITER) tokamak has refined understanding of the realities of a deuterium-tritium (D-T) burning magnetic fusion reactor. An ITER-like tokamak reactor using ITER costs and performance would lead to a cost of electricity (COE) of about 130 mills/kWh. Advanced tokamak physics to be tested in the Toroidal Physics Experiment (TPX), coupled with moderate components in engineering, technology, and unit costs, should lead to a COE comparable with best existing fission systems around 60 mills/kWh. However, a larger unit size, {approximately}2000 MW(e), is favored for the fusion system. Alternative toroidal configurations to the conventional tokamak, such as the stellarator, reversed-field pinch, and field-reversed configuration, offer some potential advantage, but are less well developed, and have their own challenges.

  13. Fusion reactor materials

    SciTech Connect

    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.

  14. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-04-04

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  15. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  16. Fusion reactor pumped laser

    DOEpatents

    Jassby, Daniel L.

    1988-01-01

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  17. An evaluation of potential material coolant compatibility for applications in advanced fusion reactors

    NASA Astrophysics Data System (ADS)

    Kondo, T.; Watanabe, Y.; Yi, Y. S.; Hishinuma, A.

    1998-10-01

    In assessing possible potential issues for fusion applications, the compatibility of several metallic structural materials was examined using high temperature/pressure steam as test environment. High corrosion resistance associated with protective oxide film formation was regarded as essential for the function of protecting from tritium permeation and corrosion damage. A Ti-Al-based intermetallic compound with V addition, recently developed, showed excellent performance. A low-activation ferritic/martensitic steel, F82-H, was comparable with the current advanced materials for modern supercritical fossil boilers, while some potential vanadium alloys, although not intended for use in steam, were found less compatible.

  18. Fusion reactor pumped laser

    DOEpatents

    Jassby, D.L.

    1987-09-04

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  19. Development of new generation reduced activation ferritic-martenstic steels for advanced fusion reactors

    DOE PAGES

    Tan, Lizhen; Snead, Lance Lewis; Katoh, Yutai

    2016-05-26

    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 experimentalmore » 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. Furthermore, 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.« less

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

    NASA Astrophysics Data System (ADS)

    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.

  1. Development of advanced blanket materials for a solid breeder blanket of a fusion reactor

    NASA Astrophysics Data System (ADS)

    Kawamura, H.; Ishitsuka, E.; Tsuchiya, K.; Nakamichi, M.; Uchida, M.; Yamada, H.; Nakamura, K.; Ito, H.; Nakazawa, T.; Takahashi, H.; Tanaka, S.; Yoshida, N.; Kato, S.; Ito, Y.

    2003-08-01

    The design of an advanced solid breeding blanket in a DEMO reactor requires a tritium breeder and a neutron multiplier that can withstand high temperatures and high neutron fluences, and the development of such advanced blanket materials has been carried out by collaboration between JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by a wet process is a reference material as a tritium breeder, but its stability at high temperatures has to be improved for its application in a DEMO blanket. One of these improved materials, TiO2-doped Li2TiO3 pebbles, was successfully fabricated and studied. For the advanced neutron multiplier, beryllides that have a high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that it had lower swelling and tritium inventory than beryllium metal. Pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. These activities have shown that there is a bright prospect in realizing a DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides.

  2. Computational mathematics and physics of fusion reactors

    PubMed Central

    Garabedian, Paul R.

    2003-01-01

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors. PMID:14614129

  3. Overview of fusion reactor safety

    NASA Astrophysics Data System (ADS)

    Cohen, S.; Crocker, J. G.

    Use of deuterium-tritium fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control; (2) neutron activation of structural materials, fluid streams and reactor hall environment; (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions; (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices; and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power.

  4. Proton Collimators for Fusion Reactors

    NASA Technical Reports Server (NTRS)

    Miley, George H.; Momota, Hiromu

    2003-01-01

    Proton collimators have been proposed for incorporation into inertial-electrostatic-confinement (IEC) fusion reactors. Such reactors have been envisioned as thrusters and sources of electric power for spacecraft and as sources of energetic protons in commercial ion-beam applications.

  5. Mirror Advanced Reactor Study (MARS). Final report. Volume 2. Commercial fusion synfuels plant

    SciTech Connect

    Donohue, M.L.; Price, M.E.

    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)

  6. The EU programme for modelling radiation effects in fusion reactor materials: An overview of recent advances and future goals

    NASA Astrophysics Data System (ADS)

    Dudarev, S. L.; Boutard, J.-L.; Lässer, R.; Caturla, M. J.; Derlet, P. M.; Fivel, M.; Fu, C.-C.; Lavrentiev, M. Y.; Malerba, L.; Mrovec, M.; Nguyen-Manh, D.; Nordlund, K.; Perlado, M.; Schäublin, R.; Van Swygenhoven, H.; Terentyev, D.; Wallenius, J.; Weygand, D.; Willaime, F.

    2009-04-01

    The EU fusion materials modelling programme was initiated in 2002 with the objective of developing a comprehensive set of computer modelling techniques and approaches, aimed at rationalising the extensive available experimental information on properties of irradiated fusion materials, developing capabilities for predicting the behaviour of materials under conditions not yet accessible to experimental tests, assessing results of tests involving high dose rates, and extrapolating these results to the fusion-relevant conditions. The programme presently gives emphasis to modelling a single class of materials, which are ferritic-martensitic EUROFER-type steels, and focuses on the investigation of key physical phenomena and interpretation of experimental observations. The objective of the programme is the development of computational capabilities for predicting changes in mechanical properties, hardening and embrittlement, as well as changes in the microstructure and phase stability of EUROFER and FeCr model alloys occurring under fusion reactor relevant irradiation conditions.

  7. Gas Evolution Measurements on Reactor Irradiated Advanced Fusion Magnet Insulation Systems

    NASA Astrophysics Data System (ADS)

    Humer, K.; Seidl, E.; Weber, H. W.; Fabian, P. E.; Feucht, S. W.; Munshi, N. A.

    2006-03-01

    Glass-fiber reinforced plastics (GFRPs) are used as insulation materials for the superconducting magnet coils of the International Thermonuclear Experimental Reactor (ITER). The radiation environment present at the magnet location will lead to gas production, swelling and weight loss of the laminate, which may result in a pressure rise combined with undefined stresses on the magnet coil casing. Consequently, these effects are important parameters for the engineering and design criteria of superconducting magnet coil structures. In this study, newly developed epoxy and cyanate-ester (CE) based S2-glass fiber reinforced insulation systems were irradiated at ambient temperature in the TRIGA-Mark II reactor (Vienna) to a fast neutron fluence of 1 and 5×1021 m-2 (E>0.1 MeV) prior to measurements of gas evolution, swelling and weight loss. The CE based laminates show increased radiation resistance, i.e. less gas evolution. The highest radiation hardness up to the highest dose was observed in a pure CE system. In addition, the effects of swelling and weight loss are either negligible or less pronounced for all systems. The results prove that the newly developed CE based composites are serious candidate insulation systems for ITER.

  8. The Cascade inertial confinement fusion reactor concept

    NASA Astrophysics Data System (ADS)

    Pitts, J. H.; Hogan, W. J.; Tobin, M. T.; Bourque, R. F.; Meier, W. R.

    1990-12-01

    The Cascade reactor concept has the potential of converting inertial confinement fusion (ICF) energy into electrical power safely, efficiently, and with low activation. Its flexibility permits a number of options for materials, blankets, fuel-pellet designs and drivers. This report documents a theoretical and experimental study culminating in a reference Cascade conceptual design that produces 890 MW of electrical power with a net plant efficiency of 47 percent if a heavy-ion driver if used. The reactor is double-cone shaped and rotates at 50 rpm. A ceramic-granule blanket flows through the reactor held against the reactor wall by the rotation. The blanket serves several functions: it absorbs energy from fusion reactions that occur at 5 Hz in the center of the reactor, thereby protecting the reactor wall and extending its lifetime to that of power plant; it acts as a heat-exchange medium to transfer fusion energy to high-pressure helium gas used in power conversion; and it produces tritium to replace that burned in the fusion process. Cascade's illumination geometry is restricted, so that good energy coupling to presently-envisioned fuel pellets is practical only with heavy-ion drivers. Laser drivers would require the use of fuel pellets with advanced design features. We discuss the reactor concept, heat exchanger, balance of plant, other systems that would be necessary for a full-scale production of electrical power, and experiments that prove the feasibility of a flowing granular blanket. A cost study predicts that Cascade, using a heavy-ion driver, could produce electricity for between 5.5 and 6.8 cents/kWh-comparable to the cost of power using modular high-temperature gas-cooled reactors, pressurized-water reactors, or coal-fired power plants. Finally, we include an annotated bibliography of the over 50 reports which have been written about Cascade.

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

    NASA Technical Reports Server (NTRS)

    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.

  10. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  11. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  12. (Meeting on fusion reactor materials)

    SciTech Connect

    Jones, R.H. ); Klueh, R.L.; Rowcliffe, A.F.; Wiffen, F.W. ); Loomis, B.A. )

    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.

  13. Cost assessment of a generic magnetic fusion reactor

    SciTech Connect

    Sheffield, J.; Dory, R.A.

    1984-01-01

    A generic magnetic fusion reactor model is used to determine the conditions under which electricity generation from fusion would be economically viable. The use of a generic model helps to circumvent problems associated with present perceptions of magnetic configurations. It helps also to decouple those limitations set by generic constraints such as nuclear cross sections from those set by the state of development today. The model shows that only moderate advances are required in reactor characteristics over current designs to make an economically attractive magnetic fusion reactor.

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

    NASA Astrophysics Data System (ADS)

    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.

  15. Modular stellarator fusion reactor concept

    NASA Astrophysics Data System (ADS)

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

    1981-08-01

    A steady-state ignited, DT-fueled, magnetic fusion reactor is proposed for use as a central electric-power station. The MSR concept combines the physics of the classic stellarator confinement topology with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an 1 = 2 system with a plasma aspect ratio of 11. The physical basis of the design point is described together with supporting magnetics, coil-force, and stress computations.

  16. Investigation of materials for fusion power reactors

    NASA Astrophysics Data System (ADS)

    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.

  17. Materials issues in fusion reactors

    NASA Astrophysics Data System (ADS)

    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

  18. Mirror Advanced Reactor Study interim design report

    SciTech Connect

    Not Available

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  19. SABR fusion-fission hybrid transmutation reactor design concept

    NASA Astrophysics Data System (ADS)

    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.

  20. Laser-driven fusion reactor

    DOEpatents

    Hedstrom, J.C.

    1973-10-01

    A laser-driven fusion reactor consisting of concentric spherical vessels in which the thermonuclear energy is derived from a deuterium-tritium (D + T) burn within a pellet'', located at the center of the vessels and initiated by a laser pulse. The resulting alpha -particle energy and a small fraction of the neutron energy are deposited within the pellet; this pellet energy is eventually transformed into sensible heat of lithium in a condenser outside the vessels. The remaining neutron energy is dissipated in a lithium blanket, located within the concentric vessels, where the fuel ingredient, tritium, is also produced. The heat content of the blanket and of the condenser lithium is eventually transferred to a conventional thermodynamic plant where the thermal energy is converted to electrical energy in a steam Rankine cycle. (Official Gazette)

  1. Generic Magnetic Fusion Reactor Revisited

    NASA Astrophysics Data System (ADS)

    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.

  2. Radiation Effects in Fission and Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Odette, G. Robert; Wirth, Brian D.

    Since the prediction of "Wigner disease" [1] and the subsequent observation of anisotropic growth of the graphite used in the Chicago Pile, the effects of radiation on materials has been an important technological concern. The broad field of radiation effects impacts many critical advanced technologies, ranging from semiconductor processing to severe materials degradation in nuclear reactor environments. Radiation effects also occur in many natural environments, ranging from deep space to inside the Earth's crust. As selected examples that involve many basic phenomena that cross-cut and illustrate the broader impacts of radiation exposure on materials, this article focuses on modeling microstructural changes in iron-based ferritic alloys under high-energy neutron irradiation relevant to light water fission reactor pressure vessels. We also touch briefly on radiation effects in structural alloys for fusion reactor first wall and blanket structures; in this case the focus is on modeling the evolution of self-interstitial atom clusters and dislocation loops. Note, since even the narrower topic of structural materials for nuclear energy applications encompass a vast literature dating from 1942, the references included in this article are primarily limited to these two narrower subjects. Thus, the references cited here are presented as examples, rather than comprehensive bibliographies. However, the interested reader is referred to proceedings of continuing symposia series that have been sponsored by several organizations, several monographs [2-4] and key journals (e.g., Journal of Nuclear Materials, Radiation Effects and Defects in Solids).

  3. Advanced fusion concepts: project summaries

    SciTech Connect

    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)

  4. Trends in fusion reactor safety research

    SciTech Connect

    Herring, J.S.; Holland, D.F.; Piet, S.J.

    1991-01-01

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g. for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions. 18 refs.

  5. Compact Fusion Advanced Rankine (CFARII) power cycle

    SciTech Connect

    Logan, B.G.

    1991-08-23

    The Compact Fusion Advanced Rankine (CFARII) power cycle is a direct plasma energy conversion scheme for inertial fusion (ICF) and magnetically-insulated, inertially confined fusion (MICF) reactors utilizing: (1) conversion of plasma thermal ionization and thermal energy into kinetic energy of a supersonic plasma jet, (2) conversion of the plasma jet kinetic energy into DC electricity by slowing down in an ``impulse`` type of magnetohydrodynamic (MHD) generator, and (3) condensation and heat rejection of the exhaust plasma on droplets of recirculating condensate (``raindrop`` condensor). A preliminary evaluation of a particular reference case CFARII Balance-of-Plant (BoP) is found sufficiently attractive (52% gross cycle efficiency, 40 million 1991 $ BoP for 1 GWe gross electric) to warrant further work on several design issues.

  6. Compact Fusion Advanced Rankine (CFARII) power cycle

    SciTech Connect

    Logan, B.G.

    1991-08-23

    The Compact Fusion Advanced Rankine (CFARII) power cycle is a direct plasma energy conversion scheme for inertial fusion (ICF) and magnetically-insulated, inertially confined fusion (MICF) reactors utilizing: (1) conversion of plasma thermal ionization and thermal energy into kinetic energy of a supersonic plasma jet, (2) conversion of the plasma jet kinetic energy into DC electricity by slowing down in an impulse'' type of magnetohydrodynamic (MHD) generator, and (3) condensation and heat rejection of the exhaust plasma on droplets of recirculating condensate ( raindrop'' condensor). A preliminary evaluation of a particular reference case CFARII Balance-of-Plant (BoP) is found sufficiently attractive (52% gross cycle efficiency, 40 million 1991 $ BoP for 1 GWe gross electric) to warrant further work on several design issues.

  7. IPFR: Integrated Pool Fusion Reactor concept

    SciTech Connect

    Sze, D.K.

    1986-01-01

    The IPFR (Integrated Pool Fusion Reactor) concept is to place a fusion reactor into a pool of molten Flibe. The Flibe will serve the multiple functions of breeding, cooling, shielding, and moderating. Therefore, the only structural material between the superconducting magnets and the plasma is the first wall. The first wall is a stand-alone structure with no coolant connection and is cooled by Flibe at the atmospheric pressure. There is also no need of the primary coolant loop. The design is expected to improve the safety, reliability, and maintainability aspects of the fusion system.

  8. Status and problems of fusion reactor development.

    PubMed

    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.

  9. Evaluation of performance of select fusion experiments and projected reactors

    NASA Technical Reports Server (NTRS)

    Miley, G. H.

    1978-01-01

    The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters.

  10. Generic Stellarator-like Magnetic Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Sheffield, John; Spong, Donald

    2015-11-01

    The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

  11. Inertial confinement fusion reactor systems

    SciTech Connect

    Frank, T.G.; Bohachevsky, I.O.; Pendergrass, J.H.

    1980-01-01

    A variety of reactor cavity concepts, drivers, and energy conversion mechanisms are being considered to realize commercial applications of ICF. Presented in this paper are: (1) a review of reactor concepts with estimates of practically achievable pulse repetition rates; (2) a survey of drivers with estimates of the requirements on reactor conditions imposed by beam propagation characteristics; and (3) an assessment of compatible driver-reactor combinations.

  12. Packed fluidized bed blanket for fusion reactor

    DOEpatents

    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.

  13. Neutronic analysis of a fusion hybrid reactor

    SciTech Connect

    Kammash, T.

    2012-07-01

    In a PHYSOR 2010 paper(1) we introduced a fusion hybrid reactor whose fusion component is the gasdynamic mirror (GDM), and whose blanket was made of thorium - 232. The thrust of that study was to demonstrate the performance of such a reactor by establishing the breeding of uranium - 233 in the blanket, and the burning thereof to produce power. In that analysis, we utilized the diffusion equation for one-energy neutron group, namely, those produced by the fusion reactions, to establish the power distribution and density in the system. Those results should be viewed as a first approximation since the high energy neutrons are not effective in inducing fission, but contribute primarily to the production of actinides. In the presence of a coolant, however, such as water, these neutrons tend to thermalize rather quickly, hence a better assessment of the reactor performance would require at least a two group analysis, namely the fast and thermal groups. We follow that approach and write an approximate set of equations for the fluxes of these groups. From these relations we deduce the all-important quantity, k{sub eff}, which we utilize to compute the multiplication factor, and subsequently, the power density in the reactor. We show that k{sub eff} can be made to have a value of 0.99, thus indicating that 100 thermal neutrons are generated per fusion neutron, while allowing the system to function as 'subcritical.' Moreover, we show that such a hybrid reactor can generate hundreds of megawatts of thermal power per cm of length depending on the flux of the fusion neutrons impinging on the blanket. (authors)

  14. Selection of a toroidal fusion reactor concept for a magnetic fusion production reactor

    NASA Astrophysics Data System (ADS)

    Jassby, D. L.

    1987-03-01

    The basic fusion driver requirements of a toroidal materials production reactor are considered. The tokamak, stellarator, bumpy torus, and reversed-field pinch are compared with regard to their demonstrated performance, probable near-term development, and potential advantages and disadvantages if used as reactors for materials production. Of the candidate fusion drivers, the tokamak is determined to be the most viable for a near-term production reactor. Four tokamak reactor concepts (TORFA/FED-R, AFTR/ZEPHYR, Riggatron, and Superconducting Coil) of approximately 500-MW fusion power are compared with regard to their demands on plasma performance, required fusion technology development, and blanket configuration characteristics. Because of its relatively moderate requirements on fusion plasma physics and technology development, as well as its superior configuration of production blankets, the TORFA/FED-R type of reactor operating with a fusion power gain of about 3 is found to be the most suitable tokamak candidate for implementation as a near-term production reactor.

  15. Advanced Concepts: Aneutronic Fusion Power and Propulsion

    NASA Technical Reports Server (NTRS)

    Chapman, John J.

    2012-01-01

    Aneutronic Fusion for In-Space thrust, power. Clean energy & potential nuclear gains. Fusion plant concepts, potential to use advanced fuels. Methods to harness ionic momentum for high Isp thrust plus direct power conversion into electricity will be presented.

  16. First wall for polarized fusion reactors

    DOEpatents

    Greenside, H.S.; Budny, R.V.; Post, D.E. Jr.

    1985-01-29

    A first-wall or first-wall coating for use in a fusion reactor having polarized fuel may be formed of a low-Z non-metallic material having slow spin relaxation, i.e., a depolarization rate greater than 1 sec/sup -1/. Materials having these properties include hydrogenated and deuterated amorphous semiconductors. A method for preventing the rapid depolarization of a polarized plasma in a fusion device may comprise the step of providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec/sup -1/.

  17. Conceptual design of Fusion Experimental Reactor

    NASA Astrophysics Data System (ADS)

    Seki, Yasushi; Takatsu, Hideyuki; Iida, Hiromasa

    1991-08-01

    Safety analysis and evaluation have been made for the FER (Fusion Experimental Reactor) as well as for the ITER (International Thermonuclear Experimental Reactor) which are basically the same in terms of safety. This report describes the results obtained in fiscal years 1988 - 1990, in addition to a summary of the results obtained prior to 1988. The report shows the philosophy of the safety design, safety analysis and evaluation for each of the operation conditions, namely, normal operation, repair and maintenance, and accident. Considerations for safety regulations and standards are also added.

  18. First wall for polarized fusion reactors

    DOEpatents

    Greenside, Henry S.; Budny, Robert V.; Post, Jr., Douglass E.

    1988-01-01

    Depolarization mechanisms arising from the recycling of the polarized fuel at the limiter and the first-wall of a fusion reactor are greater than those mechanisms in the plasma. Rapid depolarization of the plasma is prevented by providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec.sup.-1.

  19. Choice of coils for a fusion reactor

    PubMed Central

    Alexander, Romeo; Garabedian, Paul R.

    2007-01-01

    In a fusion reactor a hot plasma of deuterium and tritium is confined by a strong magnetic field to produce helium ions and release energetic neutrons. The 3D geometry of a stellarator provides configurations for such a device that reduce net toroidal current that might lead to disruptions. We construct smooth coils generating an external magnetic field designed to prevent the plasma from deteriorating. PMID:17640879

  20. Colliding Beam Fusion Reactor Space Propulsion System

    NASA Astrophysics Data System (ADS)

    Cheung, A.; Binderbauer, M.; Liu, F.; Qerushi, A.; Rostoker, N.; Wessel, F. J.

    2004-02-01

    The Colliding Beam Fusion Reactor Space Propulsion System, CBFR-SPS, is an aneutronic, magnetic-field-reversed configuration, fueled by an energetic-ion mixture of hydrogen and boron11 (H-B11). Particle confinement and transport in the CBFR-SPS are classical, hence the system is scaleable. Fusion products are helium ions, α-particles, expelled axially out of the system. α-particles flowing in one direction are decelerated and their energy recovered to ``power'' the system; particles expelled in the opposite direction provide thrust. Since the fusion products are charged particles, the system does not require the use of a massive-radiation shield. This paper describes a 100 MW CBFR-SPS design, including estimates for the propulsion-system parameters and masses. Specific emphasis is placed on the design of a closed-cycle, Brayton-heat engine, consisting of heat-exchangers, turbo-alternator, compressor, and finned radiators.

  1. Materials needs for compact fusion reactors

    SciTech Connect

    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. The spheromak as a compact fusion reactor

    SciTech Connect

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

    1987-03-01

    After summarizing the economic and utility-based rationale for compact, higher-power-density fusion reactors, the gun-sustained spheromak concept is explored as one of a number of poloidal-field-dominated confinement configurations that might improve the prospects for economically attractive and operationally simplified fusion power plants. Using a comprehensive physics/engineering/costing model for the spheromak, guided by realistic engineering constraints and physics extrapolation, a range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported. The results presented herein provide the basis for conceptual engineering designs of key fusion-power-core (FPC) subsystems and more detailed plasma modeling of this promising, high mass-power-density concept, which stresses single-piece FPC maintenance, steady-state current drive through electrostatic magnetic helicity injection, a simplified co-axial electrode-divertor, and efficient resistive-coal equilibrium-field coils. The optimal FPC size and the cost estimates project a system that competes aggressively with the best offered by alternative energy sources while simplifying considerably the complexity that has generally been associated with most approaches to magnetic fusion energy.

  3. Innovative energy production in fusion reactors

    NASA Astrophysics Data System (ADS)

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

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

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

    SciTech Connect

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

    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.

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

    SciTech Connect

    Not Available

    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.

  6. Fusion-fission hybrids for nuclear waste transmutation : a synergistic step between Gen-IV fission and fusion reactors.

    SciTech Connect

    Olson, Craig Lee; Mehlhorn, Thomas Alan; Cipiti, Benjamin B.; Rochau, Gary Eugene

    2007-09-01

    Energy demand and GDP per capita are strongly correlated, while public concern over the role of energy in climate change is growing. Nuclear power plants produce 16% of world electricity demands without greenhouse gases. Generation-IV advanced nuclear energy systems are being designed to be safe and economical. Minimizing the handling and storage of nuclear waste is important. NIF and ITER are bringing sustainable fusion energy closer, but a significant gap in fusion technology development remains. Fusion-fission hybrids could be a synergistic step to a pure fusion economy and act as a technology bridge. We discuss how a pulsed power-driven Z-pinch hybrid system producing only 20 MW of fusion yield can drive a sub-critical transuranic blanket that transmutes 1280 kg of actinide wastes per year and produces 3000 MW. These results are applicable to other inertial and magnetic fusion energy systems. A hybrid system could be introduced somewhat sooner because of the modest fusion yield requirements and can provide both a safe alternative to fast reactors for nuclear waste transmutation and a maturation path for fusion technology. The development and demonstration of advanced materials that withstand high-temperature, high-irradiation environments is a fundamental technology issue that is common to both fusion-fission hybrids and Generation-IV reactors.

  7. Demountable vacuum seals for fusion reactor applications

    SciTech Connect

    Batzer, T.H.; Call, W.R.

    1987-10-16

    Demountable vacuum seals for fusion reactor applications must be compatible with the reactor environment, easily scalable, very reliable and readily maintained by remote handling methods. We are investigating gate valves as well as flanges in our efforts to provide such seals. They are all metal and scalable without becoming massive and require no axial fasteners. Preliminary tests on an initial 30 cm aluminum flange using no soft metal coatings or gaskets have given several vacuum tight closures. Weld fatigue of this preliminary design caused degradation of the seal with further cycling to leakage levels of 10/sup -6/ Tl/sec, which is acceptable with differential pumping for either valves or flanges. Additional flange pairs using slightly altered geometry, fabrication techniques, and seal plating materials will be tested and reported on.

  8. Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  9. Pulse Star inertial confinement fusion reactor

    SciTech Connect

    Blink, J.A.; Hogan, W.J.

    1985-04-15

    Pulse Star is a pool-type ICF reactor that emphasizes low cost and high safety levels. The reactor consists of a vacuum chamber (belljar) submerged in a compact liquid metal (Li/sub 17/Pb/sub 83/ or lithium) pool which also contains the heat exchangers and liquid metal pumps. The shielding efficiency of the liquid metal pool is high enough to allow hands-on maintenance of (removed) pumps and heat exchangers. Liquid metal is allowed to spray through the 5.5 m radius belljar at a controlled rate, but is prohibited from the target region by a 4 m radius mesh first wall. The wetted first wall absorbs the fusion x-rays and debris while the spray region absorbs the fusion neutrons. The mesh allows vaporized liquid metal to blow through to the spray region where it can quickly cool and condense. Preliminary calculations show that a 2 m thick first wall could handle the mechanical (support, buckling, and x-ray-induced hoop) loads. Wetting and gas flow issues are in an initial investigation stage.

  10. Natural fueling of a tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Wan, Weigang; Parker, Scott E.; Chen, Yang; Perkins, Francis W.

    2010-04-01

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is presented. In H-mode plasmas dominated by ion-temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward toward the core. This mechanism is due to the quasineutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection is augmented by an inward pinch of cold DT fuel. The natural fueling mechanism is investigated using the gyrokinetic turbulence code GEM [Y. Chen and S. E. Parker, J. Comput. Phys. 220, 839 (2007)] and is analyzed using quasilinear theory. Profiles similar to those used for conservative International Thermonuclear Experimental Reactor [R. Aymar et al., Nucl. Fusion 41, 1301 (2001)] transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rate and energy transport. Natural fueling requires a two-component plasma and ion-ion and charge-exchange collisions set limits on this favorable effect.

  11. New concept for muon catalyzed fusion reactor

    SciTech Connect

    Tajima, T.; Eliezer, S.; Kulsrud, R.M.

    1988-12-27

    A new concept for a muon catalyzed pure fusion reactor is considered. To our best knowledge this constitutes a first plausible configuration to make energy gain without resorting to fissile matter breeding by fusion neutrons, although a number of crucial physical and engineering questions as well as details have yet to be resolved. A bundle of DT ice ribbons (with a filling factor f) is immersed in the magnetic field. The overall magnetic field in the mirror configuration confines pions created by the injected high energy deuterium (or tritium) beam. The DT materials is long enough to be inertially confined along the axis of mirror. The muon catalyzed mesomolecule formation and nuclear fusion take place in the DT target, leaving ..cap alpha../sup + +/ and occasionally (..cap alpha mu..)/sup +/ (muon sticking). The stuck muons are stripped fast enough in the target, while they are accelerated by ion cyclotron resonance heating when they circulate in the vaccum (or dilute plasma). The ribbon is (eventually) surrounded and pressure-confined by this coronal plasma, whereas the corona is magnetically confined. The overall bundle of ribbons (a pellet) is inertially confined. This configuration may also be of use for stripping stuck muons via the plasma mechanism of Menshikov and Ponomarev.

  12. Improvement in fusion reactor performance due to ion channeling

    SciTech Connect

    Emmert, G.A.; El-Guebaly, L.A.; Kulcinski, G.L.; Santarius, J.F.; Sviatoslavsky, I.N.; Meade, D.M.

    1994-11-01

    Ion channeling is a recent idea for improving the performance of fusion reactors by increasing the fraction of the fusion power deposited in the ions. In this paper the authors assess the effect of ion channeling on D-T and D-{sup 3}He reactors. The figures of merit used are the fusion power density and the cost of electricity. It is seen that significant ion channeling can lead to about a 50-65% increase in the fusion power density. For the Apollo D-{sup 3}He reactor concept the reduction in the cost of electricity can be as large as 30%.

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

    NASA Astrophysics Data System (ADS)

    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.

  14. Multivariable optimization of fusion reactor blankets

    SciTech Connect

    Meier, W.R.

    1984-04-01

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% /sup 6/Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO/sub 2/ breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO/sub 2/ breeding blanket enriched to 34% /sup 6/Li.

  15. The need and prospects for improved fusion reactors

    NASA Astrophysics Data System (ADS)

    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.

  16. Beryllium Use in the Advanced Test Reactor

    SciTech Connect

    Glen R. Longhurst

    2007-12-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) began operation in 1967. It makes use of a unique serpentine fuel core design and a beryllium reflector. Reactor control is achieved with rotating beryllium cylinders to which have been fastened plates of hafnium. Over time, the beryllium develops rather high helium content because of nuclear transmutations and begins to swell. The beryllium must be replaced at nominally 10-year intervals. Determination of when the replacement is made is by visual observation using a periscope to examine the beryllium surface for cracking and swelling. Disposition of the irradiated beryllium was once accomplished in the INL’s Radioactive Waste Management Complex, but that is no longer possible. Among contributing reasons are high levels of specific radioactive contaminants including transuranics. The INL is presently considering disposition pathways for this irradiated beryllium, but presently is storing it in the canal adjacent to the reactor. Numerous issues are associated with this situation including (1) Is there a need for ultra-low uranium material? (2) Is there a need to recover tritium from irradiated beryllium either because this is a strategic material resource or in preparation for disposal? (3) Is there a need to remove activation and fission products from irradiated beryllium? (4) Will there be enough material available to meet requirements for research reactors (fission and fusion)? In this paper will be discussed the present status of considerations on these issues.

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

    SciTech Connect

    McAdams, R.

    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.

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

    PubMed

    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.

  19. Burning nuclear wastes in fusion reactors

    SciTech Connect

    Meldner, H.W.; Howard, W.M.

    1980-02-20

    We have studied actinide burn-up in ICF reactor pellets; i.e., 14 MeV neutron fission of the very long-lived actinides that pose storage problems. A major advantage of pellet fuel region burn-up is safety: only milligrams of highly toxic and active material need to be present in the fusion chamber, whereas blanket burn-up requires the continued presence of tons of actinides in a small volume. The actinide data tables required for Monte Carlo calculations of the burn-up of /sup 241/Am and /sup 243/Am are discussed in connection with a study of the sensitivity to cross section uncertainties. More accurate and complete cross sections are required for realistic quantitative calculations.

  20. Thermomagnetic burn control for magnetic fusion reactor

    DOEpatents

    Rawls, John M.; Peuron, Unto A.

    1982-01-01

    Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors (30a, 30b, etc.) formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma (12) and a toroidal field coil (18). A mechanism (60) for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

  1. Thermomagnetic burn control for magnetic fusion reactor

    DOEpatents

    Rawls, J.M.; Peuron, A.U.

    1980-07-01

    Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma and a toroidal field coil. A mechanism for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

  2. Laser-fusion targets for reactors

    DOEpatents

    Nuckolls, John H.; Thiessen, Albert R.

    1987-01-01

    A laser target comprising a thermonuclear fuel capsule composed of a centrally located quantity of fuel surrounded by at least one or more layers or shells of material for forming an atmosphere around the capsule by a low energy laser prepulse. The fuel may be formed as a solid core or hollow shell, and, under certain applications, a pusher-layer or shell is located intermediate the fuel and the atmosphere forming material. The fuel is ignited by symmetrical implosion via energy produced by a laser, or other energy sources such as an electron beam machine or ion beam machine, whereby thermonuclear burn of the fuel capsule creates energy for applications such as generation of electricity via a laser fusion reactor.

  3. Magnetized Target Fusion in Advanced Propulsion Research

    NASA Technical Reports Server (NTRS)

    Cylar, Rashad

    2003-01-01

    The Magnetized Target Fusion (MTF) Propulsion lab at NASA Marshall Space Flight Center in Huntsville, Alabama has a program in place that has adopted to attempt to create a faster, lower cost and more reliable deep space transportation system. In this deep space travel the physics and development of high velocity plasma jets must be understood. The MTF Propulsion lab is also in attempt to open up the solar system for human exploration and commercial use. Fusion, as compared to fission, is just the opposite. Fusion involves the light atomic nuclei combination to produce denser nuclei. In the process, the energy is created by destroying the mass according to the distinguished equation: E = mc2 . Fusion energy development is being pursued worldwide as a very sustainable form of energy that is environmentally friendly. For the purposes of space exploration fusion reactions considered include the isotopes of hydrogen-deuterium (D2) and tritium (T3). Nuclei have an electrostatic repulsion between them and in order for the nuclei to fuse this repulsion must be overcome. One technique to bypass repulsion is to heat the nuclei to very high temperatures. The temperatures vary according to the type of reactions. For D-D reactions, one billion degrees Celsius is required, and for D-T reactions, one hundred million degrees is sufficient. There has to be energy input for useful output to be obtained form the fusion To make fusion propulsion practical, the mass, the volume, and the cost of the equipment to produce the reactions (generally called the reactor) need to be reduced by an order of magnitude or two from the state-of-the-art fusion machines. Innovations in fusion schemes are therefore required, especially for obtaining thrust for propulsive applications. Magnetized target fusion (MTF) is one of the innovative fusion concepts that have emerged over the last several years. MSFC is working with Los Alamos National Laboratory and other research groups in studying the

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

    ERIC Educational Resources Information Center

    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…

  5. Alternative fusion concepts and the prospects for improved reactors

    NASA Astrophysics Data System (ADS)

    Krakowski, R. A.

    1985-05-01

    Past trends, present status, and future directions in the search for an improved fusion reactor are reviewed, and promising options available to both the principle tokamak and other supporting concept are summarized.

  6. Advanced light water reactor requirements document: Chapter 4, Reactor systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the following for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR): reactor pressure vessel, nozzles and safe-ends, reactor internals, in-vessel portions of fluid systems (including reactor internal pumps (Emergency Core Cooling System (ECCS) piping and spargers), nuclear fuel, and the control rods and control rod drive system (including hydraulic supply and accumulators). Special tools required for reactor system maintenance, inspection and testing are also covered.

  7. Helium-3 blankets for tritium breeding in fusion reactors

    NASA Technical Reports Server (NTRS)

    Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

    1988-01-01

    It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

  8. Nuclear design of a very-low-activation fusion reactor

    SciTech Connect

    Cheng, E.T.; Hopkins, G.R.

    1983-06-01

    An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design.

  9. Advanced Reactor Innovation Evaluation Study (ARIES) Properties Archive

    DOE Data Explorer

    ARIES stands for Advanced Reactor Innovation Evaluation Study. It is a program and a team that explores the commercial potential of fusion as an energy resource. Though it is a multi-institutional program, ARIES is led by the University of California at San Diego. ARIES studies both magnetic fusion energy (MFE) and inertial fusion energy (IFE), using an approach that integrates theory, experiments, and technology. The ARIES team proposes fusion reactor designs and works to understand how technology, materials and plasma physics processes interact and influence each other. A 2005 report to the Fusion Energy Sciences Advisory Committee ("Scientific Challenges, Opportunities, and Priorities for the U.S. Fusion Energy Sciences Program") noted on page 98 an example of the importance of this materials properties aspect: "For instance, effects on plasma edge by various plasma facing materials and effects on various plasma stabilization and control techniques by highly conducting liquid metal blankets are being considered by physicists." This web page is an archive of material properties collected here for the use of the ARIES Fusion Power Plant Studies Team.

  10. Heat transfer in inertial confinement fusion reactor systems

    SciTech Connect

    Hovingh, J.

    1980-04-23

    The short time and deposition distance for the energy from inertial fusion products results in local peak power densities on the order of 10/sup 18/ watts/m/sup 3/. This paper presents an overview of the various inertial fusion reactor designs which attempt to reduce these peak power intensities and describes the heat transfer considerations for each design.

  11. Fission-suppressed hybrid reactor: the fusion breeder

    SciTech Connect

    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.

  12. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    SciTech Connect

    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.

  13. HYLIFE-2 inertial confinement fusion reactor design

    NASA Astrophysics Data System (ADS)

    Moir, Ralph W.

    1990-10-01

    The HYLIFE-II inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-I used liquid lithium. HYLIFE-II avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li2BeF4) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-I. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-I, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost.

  14. HYLIFE-2 inertial confinement fusion reactor design

    SciTech Connect

    Moir, R.W.

    1990-10-04

    The HYLIFE-II inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x-rays, and blast to provide a 30-y lifetime. HYLIFE-I used liquid lithium. HYLIFE-II avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li{sub 2}BeF{sub 4}) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-I. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-I, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09$/kW{center dot}h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost. 12 refs., 9 figs., 5 tabs.

  15. Design of the DEMO Fusion Reactor Following ITER

    PubMed Central

    Garabedian, Paul R.; McFadden, Geoffrey B.

    2009-01-01

    Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

  16. Challenges in the Development of Advanced Reactors

    SciTech Connect

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  17. The TITAN Reversed-Field Pinch fusion reactor study

    SciTech Connect

    Not Available

    1988-03-01

    The TITAN Reversed-Field Pinch (RFP) fusion reactor study is a multi-institutional research effort to determine the technical feasibility and key developmental issues of an RFP fusion reactor, especially at high power density, and to determine the potential economics, operations, safety, and environmental features of high-mass-power-density fusion systems. The TITAN conceptual designs are DT burning, 1000 MWe power reactors based on the RFP confinement concept. The designs are compact, have a high neutron wall loading of 18 MW/m{sup 2} and a mass power density of 700 kWe/tonne. The inherent characteristics of the RFP confinement concept make fusion reactors with such a high mass power density possible. Two different detailed designs have emerged: the TITAN-I lithium-vanadium design, incorporating the integrated-blanket-coil concept; and the TITAN-II aqueous loop-in-pool design with ferritic steel structure. This report contains a collection of 16 papers on the results of the TITAN study which were presented at the International Symposium on Fusion Nuclear Technology. This collection describes the TITAN research effort, and specifically the TITAN-I and TITAN-II designs, summarizing the major results, the key technical issues, and the central conclusions and recommendations. Overall, the basic conclusions are that high-mass power-density fusion reactors appear to be technically feasible even with neutron wall loadings up to 20 MW/m{sup 2}; that single-piece maintenance of the FPC is possible and advantageous; that the economics of the reactor is enhanced by its compactness; and the safety and environmental features need not to be sacrificed in high-power-density designs. The fact that two design approaches have emerged, and others may also be possible, in some sense indicates the robustness of the general findings.

  18. Evaluation of charcoal sorbents for helium cryopumping in fusion reactors

    SciTech Connect

    Tobin, A.G.; Sedgley, D.W.; Batzer, T.H.; Call, W.R.

    1987-01-01

    Improved methods for cryopumping helium were developed for application to fusion reactors where high helium generation rates are expected. In this study, small coconut charcoal granules were utilized as the sorbent, and braze alloys and low temperature curing cements were used as the bonding agents for attachment to a copper support structure. Problems of scale-up of the bonding agent to a 40 cm diam panel were also investigated. Our results indicate that acceptable helium pumping performance of braze bonded and cement bonded charcoals can be achieved over the range of operating conditions expected in fusion reactors.

  19. Fission-suppressed blankets for fissile fuel breeding fusion reactors

    NASA Astrophysics Data System (ADS)

    Lee, J. D.; Moir, R. W.

    1981-07-01

    Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions. The two blankets described use beryllium for neutron multiplication. One blanket uses two separate circulating molten salts: one salt for tritium breeding and the other salt for U-233 breeding. The other uses separate solid forms of lithium and thorium for breeding and helium for cooling.

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

    NASA Astrophysics Data System (ADS)

    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.

  1. Vanadium-base alloys for fusion reactor applications

    SciTech Connect

    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.

  2. Fusion energy calorimeter for the tokamak fusion test reactor

    SciTech Connect

    Jassby, D.L.; Imel, G.R.

    1981-04-01

    One and two-dimensional neutronic analyses treating the transport and scattering of neutrons and the production and transport of gamma rays in the TFTR demonstrate that the fusion energy production in a D-T pulse in the TFTR can be determined with an uncertainty of +- 15% or less, simply by integrating the measured profile of temperature increase along the central radial axis of a large hydrocarbon moderator that fills the bay between adjacent toroidal-field coils, just outside the vacuum vessel. Limitations in thermopile temperature measurements dictate a minimum fusion-neutron fluence at the vacuum vessel of the order of 10/sup 12/ n/cm/sup 2/ per pulse (a source strength of 10/sup 18/ n/pulse in TFTR), in order that this simple calorimeter can provide useful accuracy.

  3. Fuel provision for nonbreeding deuterium-tritium fusion reactors

    SciTech Connect

    Jassby, D.L.; Katsurai, M.

    1980-01-01

    Nonbreeding D-T reactors have decisive advantages in minimum size, unit cost, variety of applications, and ease of heat removal over reactors using any other fusion cycle, and significant advantages in environmental and safety characteristics over breeding D-T reactors. Considerations of relative energy production demonstrate that the most favorable source of tritium for a widely deployed system of nonbreeding D-T reactors is the very large (approx. 10 GW thermal) semi-catalyzed-deuterium (SCD), or sub-SCD reactor, where none of the escaping /sup 3/He (> 95%) or tritium (< 25%) is reinjected for burn-up. Feasibility of the ignited SCD tokamak reactor requires spatially averaged betas of 15 to 20% with a magnetic field at the TF coils of 12 to 13 Tesla.

  4. Inertial-fusion-reactor studies at Lawrence Livermore National Laboratory

    SciTech Connect

    Monsler, M.J.; Meier, W.R.

    1982-08-01

    We present results of our reactor studies for inertial-fusion energy production. Design studies of liquid-metal wall chambers have led to reactors that are remarkably simple in design, and that promise long life and low cost. Variants of the same basic design, called HYLIFE, can be used for electricity production, as a fissile-fuel factory, a dedicated tritium breeder, or hybrids of each.

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

    NASA Astrophysics Data System (ADS)

    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.

  6. CASL: The Consortium for Advanced Simulation of Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Kothe, Douglas B.

    2010-11-01

    Like the fusion community, the nuclear engineering community is embarking on a new computational effort to create integrated, multiphysics simulations. The Consortium for Advanced Simulation of Light Water Reactors (CASL), one of 3 newly-funded DOE Energy Innovation Hubs, brings together an exceptionally capable team that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated the Virtual Reactor (VR), will: 1) Enable the use of leadership-class computing for engineering design and analysis to improve reactor capabilities, 2) Promote an enhanced scientific basis and understanding by replacing empirically based design and analysis tools with predictive capabilities, 3) Develop a highly integrated multiphysics environment for engineering analysis through increased fidelity methods, and 4) Incorporate UQ as a basis for developing priorities and supporting, application of the VR tools for predictive simulation. In this presentation, we present the plans for CASL and comment on the similarity and differences with the proposed Fusion Simulation Project (FSP).

  7. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.

  8. FIREBIRD: A conceptual design of a field reversed configuration Compact Torus Fusion Reactor (CTFR)

    NASA Astrophysics Data System (ADS)

    Raman, Roger; Zubrin, Robert M.

    Work carried out by the Nuclear Engineering 512 design team at the University of Washington on a conceptual design study of a Compact-Torus (field-reverse) Fusion Reactor Configuration (CTFR) is summarized. The primary objective was to develop a reactor design for high engineering power density, modest recirculating power, and competitive cost of electrical power. A conceptual design was developed for a translating field-reversed configuration reactor; based on the physics developed by Tuszewski and Lindford at LANL and by Hoffman and Milroy at MSNW. Furthermore, it also appears possible to operate a simplified form of this reactor using a pure D-D fuel cycle after an initial D-T ignition ramp to reach the advanced fuel operating regime. One optimistic reactor so designed has a length of about 35 meters, producing a net electrical power of about 375 MWe.

  9. Design considerations for an interial confinement fusion reactor power plant

    NASA Astrophysics Data System (ADS)

    Massey, J. V.; Simpson, J. E.

    1981-08-01

    A conceptual design study to further define the engineering and economic concerns for inertial confinement fusion reactors is presented. Alternatives to the Livermore HYLIFE concept were examined and information from liquid metal cooled fast breeder reactor power plant studies was incorporated into the design. Laser and target physics models were employed in a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the JADE concept. In addition to a power plant design developed using the JADE example, the applicability of the energy absorbing gas lithium ejector concept was investigated.

  10. The neutronics studies of fusion fission hybrid power reactor

    SciTech Connect

    Zheng Youqi; Wu Hongchun; Zu Tiejun; Yang Chao; Cao Liangzhi

    2012-06-19

    In this paper, a series of neutronics analysis of hybrid power reactor is proposed. The ideas of loading different fuels in a modular-type fission blanket is analyzed, fitting different level of fusion developments, i.e., the current experimental power output, the level can be obtained in the coming future and the high-power fusion reactor like ITER. The energy multiplication of fission blankets and tritium breeding ratio are evaluated as the criterion of design. The analysis is implemented based on the D-type simplified model, aiming to find a feasible 1000MWe hybrid power reactor for 5 years' lifetime. Three patterns are analyzed: 1) for the low fusion power, the reprocessed fuel is chosen. The fuel with high plutonium content is loaded to achieve large energy multiplication. 2) For the middle fusion power, the spent fuel from PWRs can be used to realize about 30 times energy multiplication. 3) For the high fusion power, the natural uranium can be directly used and about 10 times energy multiplication can be achieved.

  11. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

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

    1999-11-01

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

  12. Performance of advanced missions using fusion propulsion

    NASA Technical Reports Server (NTRS)

    Friedlander, Alan; Mcadams, Jim; Schulze, Norm

    1989-01-01

    A quantitive evaluation of the premise that nuclear fusion propulsion offers benefits as compared to other propulsion technologies for carrying out a program of advanced exploration of the solar system and beyond is presented. Using a simplified analytical model of trajectory performance, numerical results of mass requirements versus trip time are given for robotic missions beyond the solar system that include flyby and rendezvous with the Oort cloud of comets and with the star system Alpha Centauri. Round trip missions within the solar system, including robotic sample returns from the outer planet moons and multiple asteroid targets, and manned Mars exploration are also described.

  13. Principles and rationale of the Fusion-Fission Hybrid burner reactor

    NASA Astrophysics Data System (ADS)

    Stacey, Weston M.

    2012-06-01

    The potential advantages of Fusion-Fission Hybrid (FFH) reactors (relative to critical fast reactors) for closing the back end of the nuclear fuel cycle are discussed. The choices of fission and fusion technologies for FFH burner reactors that would fission the transuranics remaining in spent fuel discharged from nuclear power reactors are summarized. The conceptual design and fuel cycle performance of the SABR FFH burner reactor are presented, and a fusion power development schedule with a symbiotic dual FFH path is outlined.

  14. Advanced PPA Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Raymond; Aske, James; Abney, Morgan B.; Miller, Lee A.; Greenwood, Zachary

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA s Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development work.

  15. Advanced Catalytic Hydrogenation Retrofit Reactor

    SciTech Connect

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  16. Future engineering needs of mirror fusion reactors

    SciTech Connect

    Thomassen, K.I.

    1982-07-30

    Fusion research has matured during the last decade and significant insight into the future program needs has emerged. While some will properly note that the crystal ball is cloudy, it is equally important to note that the shape and outline of our course is discernable. In this short summary paper, I will draw upon the National Mirror Program Plan for mirror projects and on available design studies of these projects to put the specific needs of the mirror program in perspective.

  17. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  18. Modular stellarator reactor: a fusion power plant

    SciTech Connect

    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.

  19. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    SciTech Connect

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-31

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

  20. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  1. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  2. Advanced fission and fossil plant economics-implications for fusion

    SciTech Connect

    Delene, J.G.

    1994-09-01

    In order for fusion energy to be a viable option for electric power generation, it must either directly compete with future alternatives or serve as a reasonable backup if the alternatives become unacceptable. This paper discusses projected costs for the most likely competitors with fusion power for baseload electric capacity and what these costs imply for fusion economics. The competitors examined include advanced nuclear fission and advanced fossil-fired plants. The projected costs and their basis are discussed. The estimates for these technologies are compared with cost estimates for magnetic and inertial confinement fusion plants. The conclusion of the analysis is that fusion faces formidable economic competition. Although the cost level for fusion appears greater than that for fission or fossil, the costs are not so high as to preclude fusion`s potential competitiveness.

  3. Plant maintenance and advanced reactors, 2005

    SciTech Connect

    Agnihotri, Newal

    2005-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: First U.S. EPRs in 2015, by Ray Ganthner, Framatome ANP; Pursuing several opportunities, by William E. (Ed) Cummins, Westinghouse Electric Company; Vigorous plans to develop advanced reactors, by Yuliang Sun, Tsinghua University, China; Multiple designs, small and large, by Kumiaki Moriya, Hitachi Ltd., Japan; Sealed and embedded for safety and security, by Handa Norihiko, Toshiba Corporation, Japan; Scheduled online in 2010, by Johan Slabber, PMBR (Pty) Ltd., South Africa; Multi-application reactors, by Nikolay G. Kodochigov, OKBM, Russia; Six projects under budget and on schedule, by David F. Togerson, AECL, Canada; Creating a positive image, by Scott Peterson, Nuclear Energy Institute (NEI); Advanced plans for nuclear power's renaissance, by John Cleveland, International Atomic Energy Agency, Austria; and, Plant profile: last five outages in less than 20 days, by Beth Rapczynski, Exelon Nuclear.

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

    PubMed

    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.

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

    PubMed

    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. PMID:22112596

  6. Frontier of Fusion Research: Path to the Steady State Fusion Reactor by Large Helical Device

    NASA Astrophysics Data System (ADS)

    Motojima, Osamu

    2006-12-01

    The ITER, the International Thermonuclear Experimental Reactor, which will be built in Cadarache in France, has finally started this year, 2006. Since the thermal energy produced by fusion reactions divided by the external heating power, i.e., the Q value, will be larger than 10, this is a big step of the fusion research for half a century trying to tame the nuclear fusion for the 6.5 Billion people on the Earth. The source of the Sun's power is lasting steadily and safely for 8 Billion years. As a potentially safe environmentally friendly and economically competitive energy source, fusion should provide a sustainable future energy supply for all mankind for ten thousands of years. At the frontier of fusion research important milestones are recently marked on a long road toward a true prototype fusion reactor. In its own merits, research into harnessing turbulent burning plasmas and thereby controlling fusion reaction, is one of the grand challenges of complex systems science. After a brief overview of a status of world fusion projects, a focus is given on fusion research at the National Institute for Fusion Science (NIFS) in Japan, which is playing a role of the Inter University Institute, the coordinating Center of Excellence for academic fusion research and by the Large Helical Device (LHD), the world's largest superconducting heliotron device, as a National Users' facility. The current status of LHD project is presented focusing on the experimental program and the recent achievements in basic parameters and in steady state operations. Since, its start in a year 1998, a remarkable progress has presently resulted in the temperature of 140 Million degree, the highest density of 500 Thousand Billion/cc with the internal density barrier (IDB) and the highest steady average beta of 4.5% in helical plasma devices and the largest total input energy of 1.6 GJ, in all magnetic confinement fusion devices. Finally, a perspective is given of the ITER Broad Approach program

  7. Frontier of Fusion Research: Path to the Steady State Fusion Reactor by Large Helical Device

    SciTech Connect

    Motojima, Osamu

    2006-12-01

    The ITER, the International Thermonuclear Experimental Reactor, which will be built in Cadarache in France, has finally started this year, 2006. Since the thermal energy produced by fusion reactions divided by the external heating power, i.e., the Q value, will be larger than 10, this is a big step of the fusion research for half a century trying to tame the nuclear fusion for the 6.5 Billion people on the Earth. The source of the Sun's power is lasting steadily and safely for 8 Billion years. As a potentially safe environmentally friendly and economically competitive energy source, fusion should provide a sustainable future energy supply for all mankind for ten thousands of years. At the frontier of fusion research important milestones are recently marked on a long road toward a true prototype fusion reactor. In its own merits, research into harnessing turbulent burning plasmas and thereby controlling fusion reaction, is one of the grand challenges of complex systems science.After a brief overview of a status of world fusion projects, a focus is given on fusion research at the National Institute for Fusion Science (NIFS) in Japan, which is playing a role of the Inter University Institute, the coordinating Center of Excellence for academic fusion research and by the Large Helical Device (LHD), the world's largest superconducting heliotron device, as a National Users' facility. The current status of LHD project is presented focusing on the experimental program and the recent achievements in basic parameters and in steady state operations. Since, its start in a year 1998, a remarkable progress has presently resulted in the temperature of 140 Million degree, the highest density of 500 Thousand Billion/cc with the internal density barrier (IDB) and the highest steady average beta of 4.5% in helical plasma devices and the largest total input energy of 1.6 GJ, in all magnetic confinement fusion devices. Finally, a perspective is given of the ITER Broad Approach program

  8. Diamond neutral particle spectrometer for fusion reactor ITER

    SciTech Connect

    Krasilnikov, V.; Amosov, V.; Kaschuck, Yu.; Skopintsev, D.

    2014-08-21

    A compact diamond neutral particle spectrometer with digital signal processing has been developed for fast charge-exchange atoms and neutrons measurements at ITER fusion reactor conditions. This spectrometer will play supplementary role for Neutral Particle Analyzer providing 10 ms time and 30 keV energy resolutions for fast particle spectra in non-tritium ITER phase. These data will also be implemented for independent studies of fast ions distribution function evolution in various plasma scenarios with the formation of a single fraction of high-energy ions. In tritium ITER phase the DNPS will measure 14 MeV neutrons spectra. The spectrometer with digital signal processing can operate at peak counting rates reaching a value of 10{sup 6} cps. Diamond neutral particle spectrometer is applicable to future fusion reactors due to its high radiation hardness, fast response and high energy resolution.

  9. Diamond neutral particle spectrometer for fusion reactor ITER

    NASA Astrophysics Data System (ADS)

    Krasilnikov, V.; Amosov, V.; Kaschuck, Yu.; Skopintsev, D.

    2014-08-01

    A compact diamond neutral particle spectrometer with digital signal processing has been developed for fast charge-exchange atoms and neutrons measurements at ITER fusion reactor conditions. This spectrometer will play supplementary role for Neutral Particle Analyzer providing 10 ms time and 30 keV energy resolutions for fast particle spectra in non-tritium ITER phase. These data will also be implemented for independent studies of fast ions distribution function evolution in various plasma scenarios with the formation of a single fraction of high-energy ions. In tritium ITER phase the DNPS will measure 14 MeV neutrons spectra. The spectrometer with digital signal processing can operate at peak counting rates reaching a value of 106 cps. Diamond neutral particle spectrometer is applicable to future fusion reactors due to its high radiation hardness, fast response and high energy resolution.

  10. Fusion reactor high vacuum pumping: Charcoal cryosorber tritium exposure results

    SciTech Connect

    Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M. )

    1991-01-01

    Recent experiments, have shown the practically of using activated charcoal (coconut charcoal) at 4{degrees}K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were shown to be satisfactory. The long term effects of tritium on the charcoal/cement system developed by Grumman and LLNL were not known and a program was undertaken to see what, if any, effect long term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77{degrees}K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately half way through and after the exposure. Modest effects were noted which would not seriously restrict charcoal's use as a cryosorber for fusion reactor high vacuum pumping applications. 4 refs., 8 figs.

  11. Present status of liquid metal research for a fusion reactor

    NASA Astrophysics Data System (ADS)

    Tabarés, Francisco L.

    2016-01-01

    Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

  12. Hydrogen isotopes transport parameters in fusion reactor materials

    NASA Astrophysics Data System (ADS)

    Serra, E.; Benamati, G.; Ogorodnikova, O. V.

    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.

  13. Thermonuclear Fusion Research Progress and the Way to the Reactor

    NASA Astrophysics Data System (ADS)

    Koch, Raymond

    2006-06-01

    The paper reviews the progress of fusion research and its prospects for electricity generation. It starts with a reminder of the principles of thermonuclear fusion and a brief discussion of its potential role in the future of the world energy production. The reactions allowing energy production by fusion of nuclei in stars and on earth and the conditions required to sustain them are reviewed. At the high temperatures required for fusion (hundred millions kelvins), matter is completely ionized and has reached what is called its 4th state: the plasma state. The possible means to achieve these extreme temperatures is discussed. The remainder of the paper focuses on the most promising of these approaches, magnetic confinement. The operating principles of the presently most efficient machine of this type — the tokamak — is described in some detail. On the road to producing energy with fusion, a number of obstacles have to be overcome. The plasma, a fluid that reacts to electromagnetic forces and carries currents and charges, is a complex medium. Fusion plasma is strongly heated and is therefore a good example of a system far from equilibrium. A wide variety of instabilities can grow in this system and lead to self-organized structures and spontaneous cycles. Turbulence is generated that degrades the confinement and hinders easy achievement of long lasting hot plasmas. Physicists have learned how to quench turbulence, thereby creating sort of insulating bottles inside the plasma itself to circumvent this problem. The recent history of fusion performance is outlined and the prospect of achieving power generation by fusion in a near future is discussed in the light of the development of the "International Tokamak Experimental Reactor" project ITER.

  14. Reliability assurance for regulation of advanced reactors

    SciTech Connect

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1991-01-01

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics.

  15. Reliability assurance for regulation of advanced reactors

    SciTech Connect

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1991-12-31

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics.

  16. On the fusion triple product and fusion power gain of tokamak pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.

    2016-06-01

    The energy confinement time of tokamak plasmas scales positively with plasma size and so it is generally expected that the fusion triple product, nTτ E, will also increase with size, and this has been part of the motivation for building devices of increasing size including ITER. Here n, T, and τ E are the ion density, ion temperature and energy confinement time respectively. However, tokamak plasmas are subject to operational limits and two important limits are a density limit and a beta limit. We show that when these limits are taken into account, nTτ E becomes almost independent of size; rather it depends mainly on the fusion power, P fus. In consequence, the fusion power gain, Q fus, a parameter closely linked to nTτ E is also independent of size. Hence, P fus and Q fus, two parameters of critical importance in reactor design, are actually tightly coupled. Further, we find that nTτ E is inversely dependent on the normalised beta, β N; an unexpected result that tends to favour lower power reactors. Our findings imply that the minimum power to achieve fusion reactor conditions is driven mainly by physics considerations, especially energy confinement, while the minimum device size is driven by technology and engineering considerations. Through dedicated R&D and parallel developments in other fields, the technology and engineering aspects are evolving in a direction to make smaller devices feasible.

  17. Characteristics of irradiation creep in the first wall of a fusion reactor

    SciTech Connect

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available.

  18. BOOK REVIEW: Advanced Diagnostics for Magnetic and Inertial Confinement Fusion

    NASA Astrophysics Data System (ADS)

    Stott, PE; Wootton, A.; Gorini, G.; Sindoni, E.; Batani, D.

    2003-02-01

    parameters approach ignition. Spectroscopic systems and their recent developments are well represented, whereas edge diagnostics are somewhat thin on the ground. A dedicated section is devoted to the latest tests on radiation effects and technological issues. The problems of damage to optical components and the difficulties presented by the determination of the tritium inventory are described. In the last part, the new diagnostic systems of the most recent experiments (under construction or recently operated) are reported. Various aspects of some diagnostics not included in the three previous sections are also covered, with particular emphasis on microwaves and infrared diagnostics. The book is well suited for specialists and, more generally, for people involved in nuclear fusion, who need information about the most recent developments in the field of plasma diagnostics. The papers cover many aspects of the challenges and possible solutions for performing measurements in fusion machines approaching reactor conditions. On the other hand, the contributions are in general quite advanced and would be challenging for people without a significant background in plasma diagnostics and nuclear fusion. The quality of the paper is more than satisfactory both from the point of view of clarity and of graphics. Moreover, at the beginning of the book, several papers make a considerable effort to put diagnostic issues in the wider context of present day nuclear fusion research. For those topics, which are too involved to be completely described in a conference contribution, in general adequate references are provided for deeper investigation. A Murari Approximately one third of the papers included in this volume deal with diagnostics related to inertial confinement fusion plasmas (i.e., laser-produced plasmas and pulsed-power). These papers discuss recent developments in charged particle diagnostics, neutron diagnostics, optical and x-ray measurements along with laser and particle probing

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

    NASA Astrophysics Data System (ADS)

    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.

  20. Design considerations for an inertial confinement fusion reactor power plant

    SciTech Connect

    Massey, J.V.; Simpson, J.E.

    1981-08-10

    To further define the engineering and economic concerns for inertial confinement fusion reactors (ICR's), a conceptual design study was performed by Bechtel Group Incorporated under the direction of Lawrence Livermore National Laboratory (LLNL). The study examined alternatives to the LLNL HYLIFE concept and expanded the previous balance of plant design to incorporate information from recent liquid metal cooled fast breeder reactor (LMFBR) power plant studies. The majority of the effort was to incorporate present laser and target physics models into a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the LLNL JADE concept. In addition to producing a power plant design for LLNL using the JADE example, Bechtel has also examined the applicability of the EAGLE (Energy Absorbing Gas Lithium Ejector) concept.

  1. Material options for a commercial fusion reactor first wall

    SciTech Connect

    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.

  2. Advanced Test Reactor National Scientific User Facility

    SciTech Connect

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  3. The advanced neutron source reactor: An overview

    SciTech Connect

    West, C.D.

    1990-01-01

    The Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, including neutron scattering, materials testing, materials analysis, isotope production and nuclear physics experiments. The centerpiece of the facility is to be the world's highest flux beam reactor. There will be beams of hot, cold and thermal neutrons for more than 40 simultaneous scattering and nuclear physics experiments. In addition, there will be irradiation positions and rabbit tubes for in-pile experiments, testing and isotopes production (including transuranium isotopes). To reduce technical risks and to minimize safety issues, the reactor design is based on technology already employed in existing research reactors. The fuel elements are annular assemblies of aluminum clad involute fuel plates, similar to the design of the High Flux Isotope Reactor (HFIR) at Oak Ridge and the Institut Laue-Langevin (ILL) Reactor in Grenoble. As is common with many other research reactors, the core is cooled, moderated and reflected by heavy water. The preferred fuel is U{sub 3}Si{sub 2} - a high-density fuel form developed by Argonne National Laboratory and Babcock and Wilcox that has been extensively tested in reactors in the United States, Europe and Japan. 7 figs., 2 tabs.

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

    SciTech Connect

    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.

  5. Nuclear fuel cycle analysis of the SABR fusion-fission hybrid transmutation reactor

    NASA Astrophysics Data System (ADS)

    Sommer, Chris; Stacey, Weston; Petrovic, Bojan

    2009-11-01

    Various fuel cycles have been designed and analyzed for the Subcritical Advanced Burner Reactor (SABR). SABR is a sodium cooled fast reactor fueled with transuranics (TRU) from spent fuel of light water reactors and driven by a tokamak fusion neutron source based on ITER physics and technology. SABR employs a four batch fuel cycle using an out-to-in shuffling pattern, with the fuel being reprocessed at the end of each cycle. The reprocessing method assumes recovery rates of 99.9% of the actinides and 0.1% of the fission products remain in the recycled fuel. The reprocessing fuel cycles were analyzed to find an optimal cycle length in terms of burn up, power distribution, and materials limitations. Fuel cycles are analyzed using CEA's ERANOS2.0 code, with fuel residence times limited by radiation damage at 100, 150 and 200 dpa.

  6. Electronuclear ion fusion in an ion cyclotron resonance reactor

    SciTech Connect

    Cowgill, Donald F.

    1996-12-01

    A method and apparatus for generating nuclear fusion by ion cyclotron resonance in an ion trap reactor. The reactor includes a cylindrical housing having an axial axis, an internal surface, and first and second ends. First and second end plates that are charged are respectively located at the first and second ends of the cylindrical housing. A gas layer is adsorbed on the internal surface of the cylindrical housing. Ions are desorbed from the gas layer, forming a plasma layer adjacent to the cylindrical housing that includes first ions that have a same charge sign as the first and second end plates. A uniform magnetic field is oriented along the axial axis of the cylindrical housing. Second ions, that are unlike the first ions, but have the same charge sign, are injected into the cylindrical housing along the axial axis of the cylindrical housing. A radio frequency field resonantly accelerates the injected second ions at the cyclotron resonance frequency of the second ions. The second ions circulate in increasing helical orbits and react with the first ions, at the optimum energy for nuclear fusion. The amplitude of the radio frequency field is adjusted to accelerate the second ions at a rate equal to the rate of tangential energy loss of the second ions by nuclear scattering in the first ions, causing the ions to continually interact until fusion occurs.

  7. Review of progress on fusion materials technology, Harwell, December 1980. Irradiation effects in fusion reactor materials

    NASA Astrophysics Data System (ADS)

    Harries, D. R.

    1981-03-01

    The evolution of the radiation damage structure, void and gas bubble swelling, and surface blistering effects in both model and potential first wall materials for a D-T fusion reactor system of the TOKAMAK type was investigated along with radiation effects in inorganic insulator materials. In addition, investigations were carried out into the effects of irradiation on organic insulators and on the performance of rubber seals. The principal achievements are summarized and a list of 50 references is given.

  8. Uncertainty quantification approaches for advanced reactor analyses.

    SciTech Connect

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  9. Tritium retention in fusion reactor plasma facing components

    SciTech Connect

    Langley, R.A.

    1995-03-01

    The IAEA has proposed a coordinated research program to address tritium retention and release in fusion reactor plasma facing components. This program will address materials which are mainly of interest to the design and construction of ITER, namely beryllium, carbon based materials and medium and high-Z metals, e.g. tungsten, vanadium and molybdenum, but will not be limited to these materials. Experimental data are needed for: recycling models, tritium inventory estimates, tritium permeation calculations and hydrogen embrittlement characterization. The ultimate use of the data would be to influence the formation of models for use by fusion reactor designers. Judicious material choices must be made by the designers and accurate predictive codes are required in order to make these choices. The proposed coordinated research program will provide a forum for discussions between experimentalists, theoreticians, modelers and reactor designers, provide financial support for relevant research projects and collect and evaluate experimental and theoretical data. This paper briefly reviews existing data, addresses the data gaps and points out experiments designed to obtain the needed data. 18 refs., 3 figs., 1 tab.

  10. Innovative approaches to inertial confinement fusion reactors: Final report

    SciTech Connect

    Bourque, R.F.; Schultz, K.R.

    1986-11-01

    Three areas of innovative approaches to inertial confinement fusion (ICF) reactor design are given. First, issues pertaining to the Cascade reactor concept are discussed. Then, several innovative concepts are presented which attempt to directly recover the blast energy from a fusion target. Finally, the Turbostar concept for direct recovery of that energy is evaluated. The Cascade issues discussed are combustion of the carbon granules in the event of air ingress, the use of alternate granule materials, and the effect of changes in carbon flow on details of the heat exchanger. Carbon combustion turns out to be a minor problem. Four ICF innovative concepts were considered: a turbine with ablating surfaces, a liquid piston system, a wave generator, and a resonating pump. In the final analysis, none show any real promise. The Turbostar concept of direct recovery is a very interesting idea and appeared technically viable. However, it shows no efficiency gain or any decrease in capital cost compared to reactors with conventional thermal conversion systems. Attempts to improve it by placing a close-in lithium sphere around the target to increase gas generation increased efficiency only slightly. It is concluded that these direct conversion techniques require thermalization of the x-ray and debris energy, and are Carnot limited. They therefore offer no advantage over existing and proposed methods of thermal energy conversion or direct electrical conversion.

  11. Advanced reactor physics methods for heterogeneous reactor cores

    NASA Astrophysics Data System (ADS)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  12. Radiation Hydrodynamic Simulations of an Inertial Fusion Energy Reactor Chamber

    NASA Astrophysics Data System (ADS)

    Sacks, Ryan Foster

    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 carried out using the 1-D BUCKY radiation hydrodynamics code for a variety of differing chamber fills, radii, chamber obstructions and first wall materials. Discussion of the first wall temperature rise, x-ray spectrum incident on the wall, shock timing and maximum overpressure are presented. An additional discussion of the impact of different gas opacities and their effect on overall chamber dynamics, including the formation of two shock fronts, is also presented. This work is performed under collaboration with Lawrence Livermore National Laboratory at the University of Wisconsin-Madison's Fusion Technology Institute.

  13. Novel Doppler laser radar for diagnostics in fusion reactors

    SciTech Connect

    Menon, Madhavan; Slotwinski, Anthony

    2004-10-01

    We describe the development of a novel Doppler laser radar (DOLAR) for remote measurement of flow velocity (0-10 m/s) and film thickness of liquid metal walls, currently being studied for their superior heat handling and self-healing characteristics. Small fluctuations in flow velocity({approx}mm/s) and flow thickness ({approx}50 {mu}m) that may arise during plasma discharges can also be measured. The DOLAR is also designed for non intrusive mapping of features of plasma-facing solid surfaces with very high precision ({approx}50 {mu}m). It can also measure the motion of structural components of a fusion reactor during plasma discharges and during plasma disruptions. The device utilizes frequency modulation laser radar principles for precision range measurements. Compensation of Doppler frequency shift is used to measure flow velocity. The DOLAR probe head is designed with acousto-optic and piezoelectric devices for operation in the harsh fusion environment.

  14. Biomagnetic effects: a consideration in fusion reactor development.

    PubMed Central

    Mahlum, D D

    1977-01-01

    Fusion reactors will utilize powerful magnetic fields for the confinement and heating of plasma and for the diversion of impurities. Large dipole fields generated by the plasma current and the divertor and transformer coils will radiate outward for several hundred meters, resulting in magnetic fields up to 450 gauss in working areas. Since occupational personnel could be exposed to substantial magnetic fields in a fusion power plant, an attempt has been made to assess the possible biological and health consequences of such exposure, using the existing literature. The available data indicate that magnetic fields can interact with biological material to produce effects, although the reported effects are usually small in magnitude and often unconfirmed. The existing data base is judged to be totally inadequate for assessment of potential health and environmental consequences of magnetic fields and for the establishment of appropriate standards. Requisite studies to provide an adequate data base are outlined. PMID:598345

  15. Advanced burner test reactor preconceptual design report.

    SciTech Connect

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence 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. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  16. Estimated radiactive and shock loading of fusion reactor armor

    SciTech Connect

    Swift, D C

    2008-11-25

    Inertial confinement fusion (ICF) is of interest as a source of neutrons for proliferation-resistant and high burn-up fission reactor designs. ICF is a transient process, each implosion leading to energy release over a short period, with a continuous series of ICF operations needed to drive the fission reactor. ICF yields energy in the form of MeV-range neutrons and ions, and thermal x-rays. These radiations, particularly the thermal x-rays, can deposit a pulse of energy in the wall of the ICF chamber, inducing loading by isochoric heating (i.e. at constant volume before the material can expand) or by ablation of material from the surface. The explosion of the hot ICF system, and the compression of any fill material in the chamber, may also result in direct mechanical loading by a blast wave (decaying shock) reaching the chamber wall. The chamber wall must be able to survive the repetitive loading events for long enough for the reactor to operate economically. It is thus necessary to understand the loading induced by ICF systems in possible chamber wall designs, and to predict the response and life time of the wall. Estimates are given for the loading induced in the wall armor of the fusion chamber caused by ablative thermal radiation from the fusion plasma and by the hydrodynamic shock. Taking a version of the LIFE design as an example, the ablation pressure was estimated to be {approx}0.6 GPa with an approximately exponential decay with time constant {approx}0.6 ns. Radiation hydrodynamics simulations suggested that ablation of the W armor should be negligible.

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

    SciTech Connect

    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)

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

    SciTech Connect

    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.

  19. Articulated limiter blade for a tokamak fusion reactor

    DOEpatents

    Doll, D.W.

    1982-10-21

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  20. Articulated limiter blade for a tokamak fusion reactor

    DOEpatents

    Doll, David W.

    1985-01-01

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  1. Radiation-induced segregation in candidate fusion-reactor alloys

    SciTech Connect

    Brimhall, J.L.; Baer, D.R.; Jones, R.H.

    1981-07-01

    The effect of radiation on surface segregation of minor and impurity elements has been studied in four candidate fusion reactor alloys. Radiation induced surface segregation of phosphorus was found in both 316 type stainless steel and in Nimonic PE-16. Segregation and depletion of the other alloying elements in 316 stainless steel agreed with that reported by other investigators. Segregation of nitrogen in ferritic HT-9 was enhanced by radiation but no phosphorus segregation was detected. No significant radiation enhanced or induced segregation was observed in a Ti-6Al-4V alloy. The results indicate that radiaton enhanced grain boundary segregation could contribute to the embrittlement of 316 SS and PE-16.

  2. Articulated limiter blade for a tokomak fusion reactor

    SciTech Connect

    Doll, D. W.

    1985-07-30

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  3. High conductivity Be-Cu alloys for fusion reactors

    SciTech Connect

    Lilley, E.A.; Adachi, Takao; Ishibashi, Yoshiki

    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.

  4. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  5. A Z-Pinch Driven Fusion Reactor Concept

    NASA Astrophysics Data System (ADS)

    Derzon, Mark; Rochau, Gregory; Spielman, Rick; Slutz, Stephen; Rochau, G. E.; Peterson, R. R.; Peterson, P. F.

    1999-11-01

    Recent z-pinch target physics progress has encouraged us to consider how a power reactor could be configured based on a fast z-pinch driver. Initial cost estimates show that recyclable transmission lines (RTLs) are economically viable. Providing 'standoff' between the primary power supply and the target, which is what disposable RTLs provide, has historically been the main obstacle to the consideration of pinches as fusion drivers. We will be introducing basic reactor scaling in terms of shot rate, yield, tritium breeding and neutron flux, etc. This concept has advantages in that z-pinches provide a robust mechanical environment, as well as a chamber which does not require low-pressure pumping between shots and the wall lifetime is expected to be limited factors other than neutron damage. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy under contract DE-AC04-94AL85000.

  6. Fusion-reactor plasmas with polarized nuclei. II

    SciTech Connect

    Kulsrud, R.M.; Furth, H.P.; Valeo, E.J.; Budny, R.V.; Jassby, D.L.; Micklich, B.J.; Post, D.E.; Goldhaber, M.; Happer, W.

    1982-11-01

    New techniques of bulk polarization could be used to fuel a reactor with polarized hydrogenic atoms, so as to form a plasma of polarized nuclei. Theoretical calculations indicate that, once the nuclei of the plasma are polarized in some preferred state, they can maintain this state with a probability near 100% during their lifetime in the reactor, including possible recycling. There are a number of practical advantages to be gained from the use of polarized plasma in a fusion reactor. The nuclear reaction rates can be increased or decreased, and/or the direction of emission of the reaction products can be controlled. The D-T reaction rate can be enhanced by as much as 50%, with the reaction products emitted perpendicular to the magnetic field. Alternatively, it is possible to direct the reaction products primarily along the field, with no enhancement. In this case of the D-D reaction, the theoretical predictions are somewhat less certain. Enhancement of the reaction rate by a factor of 1.5-2.5 is to be expected. In a different polarization state, suppression of D-D reactions may be feasible - a possibility that would be of interest for a neutron-free D-He/sup 3/ reactor. A quantitative discussion of the relevant nuclear physics as well as of the various mechanisms producing depolarization is given.

  7. Prospects for the development of advanced reactors

    SciTech Connect

    Semenov, B.A.; Kupitz, J.; Cleveland, J.

    1992-12-31

    Energy supply is an important prerequisite for further socio-economic development, especially in developing countries where the per capita energy use is only a very small fraction of that in industrialized countries. Nuclear energy is an essentially unlimited energy resource with the potential to provide this energy in the form of electricity, district heat and process heat under environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide a tremendous amount of experience has been accumulated during development, licensing, construction and operation of nuclear power reactors. The experience forms a sound basis for further improvements. Nuclear programmes in many countries are addressing the development of advanced reactors which are intended to have better economics, higher reliability and improved safety in order to overcome the current concerns of nuclear power. Advanced reactors now being developed could help to meet the demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced nuclear power programmes and offers assistance to countries with an interest in exploratory or research programmes.

  8. Scientific and technological advancements in inertial fusion energy

    DOE PAGES

    Hinkel, D. E.

    2013-09-26

    Scientific advancements in inertial fusion energy (IFE) were reported on at the IAEA Fusion Energy Conference, October 2012. Results presented transect the different ways to assemble the fuel, different scenarios for igniting the fuel, and progress in IFE technologies. The achievements of the National Ignition Campaign within the USA, using the National Ignition Facility (NIF) to indirectly drive laser fusion, have found beneficial the achievements in other IFE arenas such as directly driven laser fusion and target fabrication. Moreover, the successes at NIF have pay-off to alternative scenarios such as fast ignition, shock ignition, and heavy-ion fusion as well asmore » to directly driven laser fusion. As a result, this synergy is summarized here, and future scientific studies are detailed.« less

  9. Scientific and technological advancements in inertial fusion energy

    SciTech Connect

    Hinkel, D. E.

    2013-09-26

    Scientific advancements in inertial fusion energy (IFE) were reported on at the IAEA Fusion Energy Conference, October 2012. Results presented transect the different ways to assemble the fuel, different scenarios for igniting the fuel, and progress in IFE technologies. The achievements of the National Ignition Campaign within the USA, using the National Ignition Facility (NIF) to indirectly drive laser fusion, have found beneficial the achievements in other IFE arenas such as directly driven laser fusion and target fabrication. Moreover, the successes at NIF have pay-off to alternative scenarios such as fast ignition, shock ignition, and heavy-ion fusion as well as to directly driven laser fusion. As a result, this synergy is summarized here, and future scientific studies are detailed.

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

    SciTech Connect

    Chin, B.A.; Kee, C.K.; Wilcox, R.C.; Zinkle, S.J.

    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.

  11. Exploratory studies of tokamaks as fusion test reactors

    NASA Astrophysics Data System (ADS)

    Mau, T. K.; Conn, R. W.

    1982-06-01

    Studies have been performed to explore various plasma burn scenarios for a tokamak test reactor which could follow the next generation of large tokamak experiments. Tradeoffs between an ignited burning plasma and a sub-ignited driven plasma are examined in terms of device size and performance as a fusion engineering test facility. It is found that plasma performance levels, measured by ignition margin, amplification factor Q, and fusion power output, increase with device size, more optimistic transport scaling laws, lower magnetic field ripple, and higher Β. The performance of a generally low stress ( B 0=4 T) reference device, with major radius R=4.5 m and minor radius a=1.3 m in a D-shaped ( κ=1.6) plasma has been evaluated over a wide range of operating parameters. In particular, a moderate fusion power output of 300 MW is obtained, the driven plasma having Q≅ 10, an edge ripple of 1%, and a density ranging between 1.0 and 1.5×1014 cm-3. The same device operated at a higher general level of stress ( B 0=5.3 T) is predicted to achieve ignition, but is not required for the mission of an engineering test facility and would entail greater technical risk.

  12. Advanced Small Modular Reactor Economics Status Report

    SciTech Connect

    Harrison, Thomas J.

    2014-10-01

    This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic and nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation

  13. Plasma transport control and self-sustaining fusion reactor

    SciTech Connect

    Ono, M.; Bell, R.; Choe, W.

    1997-02-01

    The possibility of a high performance/low cost fusion reactor concept which can simultaneously satisfy (1) high beta, (2) high bootstrap fraction (self-sustaining), and (3) high confinement is discussed. In CDX-U, a tokamak configuration was created and sustained solely by internally generated bootstrap currents, in which a seed current is created through a non-classical current diffusion process. Recent theoretical studies of MHD stability limits in spherical torus [e.g., the National Spherical Torus Experiment (NSTX)] produced a promising regime with stable beta of 45% and bootstrap current fraction of {ge}99%. Since the bootstrap current is generated by the pressure gradient, to satisfy the needed current profile for MHD stable high beta regimes, it is essential to develop a means to control the pressure profile. It is suggested that the most efficient approach for pressure profile control is through a creation of transport barriers (localized regions of low plasma transport) in the plasma. As a tool for creating the core transport barrier, poloidal-sheared-flow generation by ion Bernstein waves (IBW) near the wave absorption region appears to be promising. In PBX-M, application of IBW power produced a high-quality internal transport barrier where the ion energy and particle transport became neoclassical in the barrier region. The observation is consistent with the IBW-induced-poloidal-sheared-flow model. An experiment is planned on TFTR to demonstrate this concept with D-T reactor-grade plasmas. For edge transport control, a method based on electron ripple injection (ERI), driven by electron cyclotron heating (ECH), is being developed on CDX-U. It is estimated that both the IBW and ERI methods can create a transport barrier in reactor-grade plasmas (e.g., ITER) with a relatively small amount of power ({approx}10 MW {much_lt} P{sub fusion}).

  14. Introduction to D-He(3) fusion reactors

    NASA Technical Reports Server (NTRS)

    Vlases, G. C.; Steinhauer, L. C.

    1989-01-01

    A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

  15. Plant maintenance and advanced reactors, 2007

    SciTech Connect

    Agnihotri, Newal

    2007-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: A new day for energy in America; Committed to success more than ever, by Andy White, GE--Hitachi Nuclear Energy; Competitive technology for decades, by Steve Tritch, Westinghouse Electric Company; Pioneers of positive community relationship, by Exelon Nuclear; A robust design for 60-years, by Ray Ganthner, Areva; Aiming at no evacuation plants, by Kumiaki Moriya, Hitachi-GE Nuclear Energy, Ltd.; and, Desalination and hydrogen economy, by Dr. I. Khamis, International Atomic Energy Agency. Industry innovation articles in this issue are: Reactor vessel closure head project, by Jeff LeClair, Prairie Island Nuclear Generating Plant; and Submersible remote-operated vehicle, by Michael S. Rose, Entergy's Fitzpatrick Nuclear Station.

  16. Preliminary evaluation of beta-spodumene as a fusion reactor structural material

    SciTech Connect

    Kelsey, P.V. Jr.; Schmunk, R.E.; Henslee, S.P.

    1981-01-01

    Beta-spodumene was investigated as a candidate material for use in fusion reactor environments. Properties which support the use of beta-spodumene include good thermal shock resistance, a very low coefficient of thermal expansion, a low-Z composition which would result in minimum impact on the plasma, and flexibility in fabrication processes. Specimens were irradiated in the Advanced Test Reactor (ATR) to a fluence of 5.3 x 10/sup 22/ n/m/sup 2/, E > 0.1 MeV, and 4.9 x 10/sup 23/ n/m/sup 2/ thermal fluence in order to obtain a preliminary evaluation of the impact of irradiation on the material. Preliminary data indicate that the mechanical properties of beta-spodumene are little affected by irradiation. Gas production and release have also been investigated.

  17. Plant maintenance and advanced reactors issue, 2008

    SciTech Connect

    Agnihotri, Newal

    2009-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

  18. Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor.

    PubMed

    Nie, Baojie; Ni, Muyi; Jiang, Jieqiong; Wu, Yican

    2015-10-01

    As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system.

  19. Context representation and fusion: advancements and opportunities.

    PubMed

    Khattak, Asad Masood; Akbar, Noman; Aazam, Mohammad; Ali, Taqdir; Khan, Adil Mehmood; Jeon, Seokhee; Hwang, Myunggwon; Lee, Sungyoung

    2014-01-01

    The acceptance and usability of context-aware systems have given them the edge of wide use in various domains and has also attracted the attention of researchers in the area of context-aware computing. Making user context information available to such systems is the center of attention. However, there is very little emphasis given to the process of context representation and context fusion which are integral parts of context-aware systems. Context representation and fusion facilitate in recognizing the dependency/relationship of one data source on another to extract a better understanding of user context. The problem is more critical when data is emerging from heterogeneous sources of diverse nature like sensors, user profiles, and social interactions and also at different timestamps. Both the processes of context representation and fusion are followed in one way or another; however, they are not discussed explicitly for the realization of context-aware systems. In other words most of the context-aware systems underestimate the importance context representation and fusion. This research has explicitly focused on the importance of both the processes of context representation and fusion and has streamlined their existence in the overall architecture of context-aware systems' design and development. Various applications of context representation and fusion in context-aware systems are also highlighted in this research. A detailed review on both the processes is provided in this research with their applications. Future research directions (challenges) are also highlighted which needs proper attention for the purpose of achieving the goal of realizing context-aware systems. PMID:24887042

  20. Context Representation and Fusion: Advancements and Opportunities

    PubMed Central

    Khattak, Asad Masood; Akbar, Noman; Aazam, Mohammad; Ali, Taqdir; Khan, Adil Mehmood; Jeon, Seokhee; Hwang, Myunggwon; Lee, Sungyoung

    2014-01-01

    The acceptance and usability of context-aware systems have given them the edge of wide use in various domains and has also attracted the attention of researchers in the area of context-aware computing. Making user context information available to such systems is the center of attention. However, there is very little emphasis given to the process of context representation and context fusion which are integral parts of context-aware systems. Context representation and fusion facilitate in recognizing the dependency/relationship of one data source on another to extract a better understanding of user context. The problem is more critical when data is emerging from heterogeneous sources of diverse nature like sensors, user profiles, and social interactions and also at different timestamps. Both the processes of context representation and fusion are followed in one way or another; however, they are not discussed explicitly for the realization of context-aware systems. In other words most of the context-aware systems underestimate the importance context representation and fusion. This research has explicitly focused on the importance of both the processes of context representation and fusion and has streamlined their existence in the overall architecture of context-aware systems’ design and development. Various applications of context representation and fusion in context-aware systems are also highlighted in this research. A detailed review on both the processes is provided in this research with their applications. Future research directions (challenges) are also highlighted which needs proper attention for the purpose of achieving the goal of realizing context-aware systems. PMID:24887042

  1. Plant maintenance and advanced reactors issue, 2004

    SciTech Connect

    Agnihotri, Newal

    2004-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Optimism about the future of nuclear power, by Ruth G. Shaw, Duke Power Company; Licensed in three countries, by GE Energy; Enhancing public acceptance, by Westinghouse Electric Company; Standardized MOV program, by Ted Neckowicz, Exelon; Inservice testing, by Steven Unikewicz, U.S. Nuclear Regulatory Commission; Asian network for education, Fatimah Mohd Amin, Malaysian Institute for Nuclear Technology Research; and, Cooling water intake optimization, by Jeffrey M. Jones and Bert Mayer, P.E., Framatome ANP.

  2. Programmable AC power supply for simulating power transient expected in fusion reactor

    SciTech Connect

    Halimi, B.; Suh, K. Y.

    2012-07-01

    This paper focus on control engineering of the programmable AC power source which has capability to simulate power transient expected in fusion reactor. To generate the programmable power source, AC-AC power electronics converter is adopted to control the power of a set of heaters to represent the transient phenomena of heat exchangers or heat sources of a fusion reactor. The International Thermonuclear Experimental Reactor (ITER) plasma operation scenario is used as the basic reference for producing this transient power source. (authors)

  3. Astronomers Reveal Extinct Extra-Terrestrial Fusion Reactor

    NASA Astrophysics Data System (ADS)

    2004-06-01

    An international team of astronomers, studying the left-over remnants of stars like our own Sun, have found a remarkable object where the nuclear reactor that once powered it has only just shut down. This star, the hottest known white dwarf, H1504+65, seems to have been stripped of its entire outer regions during its death throes leaving behind the core that formed its power plant. Scientists from the United Kingdom, Germany and the USA focused two of NASA's space telescopes, the Chandra X-ray Observatory and the Far Ultraviolet Spectroscopic Explorer (FUSE), onto H1504+65 to probe its composition and measure its temperature. The data revealed that the stellar surface is extremely hot, 200,000 degrees, and is virtually free of hydrogen and helium, something never before observed in any star. Instead, the surface is composed mainly of carbon and oxygen, the 'ashes' of the fusion of helium in a nuclear reactor. An important question we must answer is why has this unique star lost the hydrogen and helium, which usually hide the stellar interior from our view? Professor Martin Barstow (University of Leicester) said. 'Studying the nature of the ashes of dead stars give us important clues as to how stars like the Sun live their lives and eventually die. The nuclear waste of carbon and oxygen produced in the process are essential elements for life and are eventually recycled into interstellar space to form new stars, planets and, possibly, living beings.' Professor Klaus Werner (University of Tübingen) said. 'We realized that this star has, on astronomical time scales, only very recently shut down nuclear fusion (about a hundred years ago). We clearly see the bare, now extinct reactor that once powered a bright giant star.' Dr Jeffrey Kruk (Johns Hopkins University) said: 'Astronomers have long predicted that many stars would have carbon-oxygen cores near the end of their lives, but I never expected we would actually be able to see one. This is a wonderful opportunity to

  4. Alloy development for irradiation performance in fusion reactors

    NASA Astrophysics Data System (ADS)

    Harling, O. K.; Grant, N. J.

    1980-12-01

    The development of improved structural alloys for the fusion reactor first wall application is addressed. Several new alloys were produced by rapid solidification. Emphasis in alloy design and production was placed on producing austenitic Type 316SS with fine dispersions of TiC and Al2O3 particles. Results of mechanical and microstructural tests are presented. A number of neutron irradiations were initiated on samples fabricated from alloys produced. A dual beam, heavy ion, and helium ion, irradiation was completed using several alloys and a range of temperatures, damage rates, and total doses. Modeling of irradiation phenomena was continued with emphasis on understanding the effect of recoil resolution on relatively stable second phase particles. The microstructure of several ZrB2 doped stainless steels was characterized.

  5. Method and apparatus for making uniform pellets for fusion reactors

    DOEpatents

    Budrick, Ronald G.; King, Frank T.; Martin, Alfred J.; Nolen, Jr., Robert L.; Solomon, David E.

    1977-01-01

    A method and apparatus for making uniform pellets for laser driven fusion reactors which comprises selection of a quantity of glass frit which has been accurately classified as to size within a few micrometers and contains an occluded material, such as urea, which gasifies and expands when heated. The sized particles are introduced into an apparatus which includes a heated vertical tube with temperatures ranging from 800.degree. C to 1300.degree. C. The particles are heated during the drop through the tube to molten condition wherein the occluded material gasifies to form hollow microspheres which stabilize in shape and plunge into a collecting liquid at the bottom of the tube. The apparatus includes the vertical heat resistant tube, heaters for the various zones of the tube and means for introducing the frit and collecting the formed microspheres.

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

    NASA Astrophysics Data System (ADS)

    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.

  7. Tritium pellet injector design for tokamak fusion test reactor

    SciTech Connect

    Fisher, P.W.; Baylor, L.R.; Bryan, W.E.; Combs, S.K.; Easterly, C.E.; Lunsford, R.V.; Milora, S.L.; Schuresko, D.D.; White, J.A.; Williamson, D.H.

    1985-01-01

    A tritium pellet injector (TPI) system has been designed for the Tokamak Fusion Test Reactor (TFTR) Q approx. 1 phase of operation. The injector gun utilizes a radial design with eight independent barrels and a common extruder to minimize tritium inventory. The injection line contains guide tubes with intermediate vacuum pumping stations and fast valves to minimize propellant leakage to the torus. The vacuum system is designed for tritium compatibility. The entire injector system is contained in a glove box for secondary containment protection against tritium release. Failure modes and effects have been analyzed, and structural analysis has been performed for most intense predicted earthquake conditions. Details of the design and operation of this system are presented in this paper.

  8. Laser in vessel-viewing system for nuclear fusion reactors

    NASA Astrophysics Data System (ADS)

    Bartolini, Luciano; Bordone, Andrea; Coletti, Alberto; Ferri De Collibus, Mario; Fornetti, Giorgio G.; Lupini, S.; Neri, Carlo; Poggi, Claudio; Riva, Marco; Semeraro, Luigi; Talarico, Carlo

    2000-11-01

    An amplitude modulated laser radar has been developed by ENEA (Italian Agency for New Technologies, Energy and Environment) for periodic in-vessel inspection in large fusion machines. Its overall optical design has been developed taking into account the extremely high radiation levels and operating temperatures foreseen in large European fusion machines such as JET (Joint European Torus) and ITER (International Thermo- nuclear Experimental Reactor). The viewing system is based on a transceiving optical radar using a RF modulated single mode 840 nm wavelength laser beam. The sounding beam is transmitted through a coherent optical fiber and a focusing optic to the inner part of the nuclear reactor vessel by a stainless steel probe on the tip of which a suitable scanning silica prism steers the laser beam along a linear raster spanning a -90 degree(s) to +60 degree(s) in elevation and 360 degree(s) in azimuth for a complete mapping of the vessel itself. All the electronics, including the laser source, avalanche photodiode and all the active components are located outside the bioshield, while passive components (receiving optics, transmitting collimator, fiber optics), located in the torus hall, are made of fused silica so that the overall laser radar is radiation resistant. The signal is acquired, the raster lines being synchronized with the aid of optical encoders linked to the scanning prism, thus yielding a TV like image. Preliminary results have been obtained scanning large sceneries including several real targets having different backscattering properties, colors and surface reflectivity ranging over several decades to simulate the expected dynamic range of the video signals incoming from the vessel.

  9. The Magnetic Dipole as an Attractive Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Dawson, John M.

    1997-11-01

    Stability for low β plasma confined by closed B field lines is PV^γ = C_0, P = pressure, V = flux tube volume, γ is c_p/cv = 5/3. Kesner(J. Kesner, Innovative Confinement Concepts Workshop, Mar. 3-6, 1997) proposed a levitated current ring with the plasma stabilized by this condition as an alternate fusion reactor. Such a reactor has many attractive features; at radii large compared to the ring radius, V goes like r^4; the stability condition is Pr^20/3 = C_1. If nr^4 = C_2, then interchanges keep the density constant. The temperature can drop according to Tr^8/3 = C_3. If the chamber is ten times the ring radius, the density can drop from 10^14 near the ring to 10^10 at the edge and the temperature can drop from 50 keV near the ring to 100 eV at the edge. This plasma should present no problems for a divertor. Reacting plasma near the ring will heat it, upsetting the stability relation and cause convection to carry burnt plasma out; it will cool as it expands. At the same time the convection will bring in fresh fuel from the outside which will be compressed and heated to ignition. A super conducting ring design that can float in reacting D-He^3 for 16 hours exists(J.M. Dawson, FUSION, edited by Edward Teller, Vol. 1, Magnetic Confinement, Part, Ch. 16, Academic Press, 1981).

  10. Structural materials challenges for advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  11. Advanced Burner Reactor Preliminary NEPA Data Study.

    SciTech Connect

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  14. Transient getter scheme for the Tokamak Fusion Test Reactor

    SciTech Connect

    Cecchi, J.L.; Cohen, S.A.; Sredniawski, J.J.

    1980-01-01

    The ability of the Tokamak Fusion Test Reactor (TFTR) to attain the largest fusion power gain depends critically on minimizing plasma contamination and controlling the densities of the reacting deuterium and tritium. Experiments on a number of tokamaks have demonstrated that gettering over an appreciable surface area (greater than or equal to 10%) of the vacuum vessel greatly facilitates both of these objectives. One particular problem in implementing a surface pumping system in TFTR, however, is a restriction on the maximum allowable tritium content of the getter. This restriction could require regeneration of the absorbed tritium after as few as 50 machine pulses. We have developed a scheme utilizing SAES Zr/Al getter modules which obviates the need for such frequent interruptions of machine operation by taking advantage of the pulsed operation of TFTR. With the Zr/Al getter at temperatures between 500/sup 0/C to 600/sup 0/C it is possible to achieve a quasi-steady state in the tritium loading where the quantity of tritium desorbed between pulses is equal to the quantity which is absorbed during a pulse. Since frequent thermal cycling is not required, this scheme also reduces the possibility of Zr/Al getter material fatigue.

  15. ASME Material Challenges for Advanced Reactor Concepts

    SciTech Connect

    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.

  16. Particle confinement property in the cusp-mirror field of a compact fusion reactor

    NASA Astrophysics Data System (ADS)

    Zhu, Limin; Liu, Haifeng; Wang, Xianqu

    2016-09-01

    The cusp-mirror magnetic structure in a compact fusion reactor (CFR) is investigated to understand the properties of the particle confinement for the first time. Compared with a cascade magnetic mirror device, its advanced performance is shown by means of test particle simulations. Some interesting results are obtained as follows: the adiabatic region and non-adiabatic region are found in the CFR’s magnetic configuration. In the non-adiabatic region, due to the magnetic field-free space existing, the ions are scattered stochastically and are not directly guided into the loss cone, unlike the particles around the fixed magnetic lines in the adiabatic region, which decrease the ion loss fraction. The CFR’s configuration, combining advantages of cusp-magnetic configuration and mirror-magnetic configuration, leads to confine particles longer than cascade magnetic mirror’s. This phenomenon may be relevant to the construction of advanced magnetic-confinement devices.

  17. Identification and characterization of RET fusions in advanced colorectal cancer

    PubMed Central

    Garrett, Christopher R.; Seery, Tara; Sanford, Eric M.; Balasubramanian, Sohail; Ross, Jeffrey S.; Stephens, Philip J.; Miller, Vincent A.; Ali, Siraj M.; Chiu, Vi K.

    2015-01-01

    There is an unmet clinical need for molecularly directed therapies available for metastatic colorectal cancer. Comprehensive genomic profiling has the potential to identify actionable genomic alterations in colorectal cancer. Through comprehensive genomic profiling we prospectively identified 6 RET fusion kinases, including two novel fusions of CCDC6-RET and NCOA4-RET, in metastatic colorectal cancer (CRC) patients. RET fusion kinases represent a novel class of oncogenic driver in CRC and occurred at a 0.2% frequency without concurrent driver mutations, including KRAS, NRAS, BRAF, PIK3CA or other fusion tyrosine kinases. Multiple RET kinase inhibitors were cytotoxic to RET fusion kinase positive cancer cells and not RET fusion kinase negative CRC cells. The presence of a RET fusion kinase may identify a subset of metastatic CRC patients with a high response rate to RET kinase inhibition. This is the first characterization of RET fusions in CRC patients and highlights the therapeutic significance of prospective comprehensive genomic profiling in advanced CRC. PMID:26078337

  18. Identification and characterization of RET fusions in advanced colorectal cancer.

    PubMed

    Le Rolle, Anne-France; Klempner, Samuel J; Garrett, Christopher R; Seery, Tara; Sanford, Eric M; Balasubramanian, Sohail; Ross, Jeffrey S; Stephens, Philip J; Miller, Vincent A; Ali, Siraj M; Chiu, Vi K

    2015-10-01

    There is an unmet clinical need for molecularly directed therapies available for metastatic colorectal cancer. Comprehensive genomic profiling has the potential to identify actionable genomic alterations in colorectal cancer. Through comprehensive genomic profiling we prospectively identified 6 RET fusion kinases, including two novel fusions of CCDC6-RET and NCOA4-RET, in metastatic colorectal cancer (CRC) patients. RET fusion kinases represent a novel class of oncogenic driver in CRC and occurred at a 0.2% frequency without concurrent driver mutations, including KRAS, NRAS, BRAF, PIK3CA or other fusion tyrosine kinases. Multiple RET kinase inhibitors were cytotoxic to RET fusion kinase positive cancer cells and not RET fusion kinase negative CRC cells. The presence of a RET fusion kinase may identify a subset of metastatic CRC patients with a high response rate to RET kinase inhibition. This is the first characterization of RET fusions in CRC patients and highlights the therapeutic significance of prospective comprehensive genomic profiling in advanced CRC. PMID:26078337

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

    SciTech Connect

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

    1995-01-01

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

  20. Tandem-mirror end plugs for future fusion reactors

    NASA Astrophysics Data System (ADS)

    1981-06-01

    Electrostatic fields for confining central-cell plasma are achieved by heating the electrons in end-plugs via electron-cyclotron-resonance heating. Four end-plug magnetic configurations are being developed and tested to determine which will provide the best thermal barrier between plug- and central-cell electrons in a fusion reactor: (1) the inside barrier, with its auxiliary solenoid; (2) the auxiliary-mirror-cell (A-cell) barrier, which makes use of C-shaped magnet coils; (3) the axisymmetric-cusp barrier, using circular coils; and (4) the electron-ring barrier, in which two magnetic coils are stabilized by a ring of hot electrons. Calculations of the magnetohydrodynamic (MHD) stability are being performed with respect to the magnetic curvatures of each end-plug configuration. Models for describing the behavior of plasmas with finite ion orbits are being developed to predict MHD stability. Charge-exchange pumping systems for reactors with inside, A-cell, and axisymmetric-cusp barriers have already been designed, and a pumping system for removing thermalized helium ions is being explored.

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

    NASA Astrophysics Data System (ADS)

    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).

  2. Reactor prospects of muon-catalyzed fusion of deuterium and tritium concentrated in transition metals

    SciTech Connect

    Stacey, W.M. Jr. . Fusion Research Center )

    1989-09-01

    It is conjectured that the number of fusion events catalyzed by a single muon is orders of magnitude greater for deuterium and tritium concentrated in a transition metal than in gaseous form and that the recent observation of 2.5-MeV neutrons from a D/sub 2/O electrolytic cell with palladium and titanium cathodes can thereby be interpreted in terms of cosmic muon-catalyzed deuterium-deuterium fusion. This suggests a new fusion reactor reactor consisting of deuterium and tritium concentrated in transition metal fuel elements in a fusion core that surrounds an accelerator-produced muon source. The feasibility of net energy production in such a reactor is established in terms of requirements on the number of fusion events catalyzed per muon. The technological implications for a power reactor based on this concept are examined. The potential of such a concept as a neutron source for materials testing and tritium and plutonium production is briefly discussed.

  3. Environmental and economic assessments of magnetic and inertial fusion energy reactors

    NASA Astrophysics Data System (ADS)

    Yamazaki, K.; Oishi, T.; Mori, K.

    2011-10-01

    Global warming due to rapid greenhouse gas (GHG) emissions is one of the present-day crucial problems, and fusion reactors are expected to be abundant electric power generation systems to reduce human GHG emission amounts. To search for an environmental-friendly and economical fusion reactor system, comparative system studies have been done for several magnetic fusion energy reactors, and have been extended to include inertial fusion energy reactors. We clarify new scaling formulae for the cost of electricity and GHG emission rate with respect to key design parameters, which might be helpful in making a strategy for fusion research development. Comparisons with other conventional electric power generation systems are carried out taking into account the introduction of GHG taxes and the application of the carbon dioxide capture and storage system to fossil power generators.

  4. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  5. Plasma transport control and self-sustaining fusion reactor

    SciTech Connect

    Ono, M.; Peng, Yueng Kay Martin

    1997-01-01

    The possibility of a high-performance/low-cost fusion reactor concept which can simultaneously satisfy (1) high beta, (2) high bootstrap fractio (self-sustaining) and (3) high confinement is discussed. In CDX-U, a tokamak configuration was created and sustained solely by internally generated bootstrap currents, in which a seed current is created through a nonclassical current diffusion process. Recent theoretical studies of MHD stability limits in spherical tori [e.g. the National Spherical Torus Experiment (NSTX)] produced a promising regime with stable beta of 45% and bootstrap current fraction of > 99%. Since the bootstrap current is generated by the pressure gradient, to satisfy the needed current profile for MHD stable high beta regimes, it is essential to develop a means to control the pressure profile. It is suggested that the most efficient approach for pressure profile control is through the creation of transport barriers (localized regions of low plasma transport) in the plasma. As a tool for creating the core transport barrier, poloidal-sheared-flow generation by ion Bernstein waves (IBW) near the wave absorption region appears to be promising. In PBX-M, application of IBW power produced a high-quality internal transport barrier where the ion energy and particle transport became neoclassical in the barrier region. The observation is consistent with the IBW-inducedpoloidal- sheared-flow model. An experiment is planned on TFTR to demonstrate this concept with D T reactor-grade plasmas. For edge transport control, a method based on electron ripple injection (ERI), driven by electron cyclotron heating (ECH), is being developed on CDX-U. It is estimated that both the IBW and ERI methods can create a transport barrier in reactor-grade plasmas (e.g. ITER) with a relatively small amount of power ( 10 MW Pfusion).

  6. Plasma transport control and self-sustaining fusion reactor

    NASA Astrophysics Data System (ADS)

    Ono, M.; Bell, R.; Choe, W.; Chang, C. S.; Forest, C. B.; Goldston, R.; Hwang, Y. S.; Jardin, S. C.; Kaita, R.; Kaye, S.; Kessel, C. E.; Kugel, H.; LeBlanc, B.; Manickam, J.; Menard, J. E.; Munsat, T.; Okabayashi, M.; Peng, M.; Sesnic, S.; Tighe, W.

    1997-05-01

    The possibility of a high-performance/low-cost fusion reactor concept which can simultaneously satisfy (1) high beta, (2) high bootstrap fractio (self-sustaining) and (3) high confinement is discussed. In CDX-U, a tokamak configuration was created and sustained solely by internally generated bootstrap currents, in which a 'seed' current is created through a nonclassical current diffusion process. Recent theoretical studies of MHD stability limits in spherical tori [e.g. the National Spherical Torus Experiment (NSTX)] produced a promising regime with stable beta of 45% and bootstrap current fraction of $\\ge$ 99%reactor-grade plasmas. For edge transport control, a method based on electron ripple injection (ERI), driven by electron cyclotron heating (ECH), is being developed on CDX-U. It is estimated that both the IBW and ERI methods can create a transport barrier in reactor-grade plasmas (e.g. ITER) with a relatively small amount of power $(\\approx 10 MW \\ll P_{<span class=fusion})$" SRC="http://www.iop.org/0741-3335/39/5A/033/img2

  7. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    SciTech Connect

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

    This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

  8. Corrosion of spent Advanced Test Reactor fuel

    SciTech Connect

    Lundberg, L.B.; Croson, M.L.

    1994-11-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented.

  9. Advanced ceramic cladding for water reactor fuel

    SciTech Connect

    Feinroth, H.

    2000-07-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

  10. Advanced nuclear reactor public opinion project

    SciTech Connect

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  11. Advanced Small Modular Reactor Economics Model Development

    SciTech Connect

    Harrison, Thomas J.

    2014-10-01

    The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the

  12. LBB application in the US operating and advanced reactors

    SciTech Connect

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  13. Development of ferritic steels for fusion reactor applications

    SciTech Connect

    Klueh, R.L.; Maziasz, P.J.; Corwin, W.R.

    1988-08-01

    Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for future fusion reactors. However, irradiation of such steels in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste-disposal problems. Such problems could be reduced by replacing the elements in the steels (i.e., Mo, Nb, Ni, N, and Cu) that lead to long-lived radioactive isotopes. We have proposed the development of ferritic steels analogous to conventional Cr-Mo steels, which contain molybdenum and niobium. It is proposed that molybdenum be replaced by tungsten and niobium be replaced by tantalum. Eight experimental steels were produced. Chromium concentrations of 2.25, 5, 9, and 12% were used (all concentrations are in wt %). Steels with these chromium compositions, each containing 2% W and 0.25% V, were produced. To determine the effect of tungsten and vanadium, 2.25 Cr steels were produced with 2% W and no vanadium and with 0.25% V and O and 1% W. A 9Cr steel containing 2% W, 0.25 V, and 0.07% Ta was also studied. For all alloys, carbon was maintained at 0.1%. Tempering studies on the normalized steels indicated that the tempering behavior of the new Cr-W steels was similar to that of the analogous Cr-Mo steels. Microscopy studies indicated that 2% tungsten was required in the 2.25 Cr steels to produce 100% bainite in 15.9-mm-thick plate during normalization. The 5Cr and 9Cr steels were 100% martensite, but the 12 Cr steel contained about 75% martensite with the balance delta-ferrite. 33 refs., 35 figs., 5 tabs.

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

    SciTech Connect

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

    1994-07-01

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

  15. Overview of Indian activities on fusion reactor materials

    NASA Astrophysics Data System (ADS)

    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

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

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

    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.

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

    SciTech Connect

    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.

  19. Scale length study in TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Hiroe, S.; Goldston, R.J.; Bitter, M.; Bush, C.E.; Efthimion, P.C.; Grek, B.; Johnson, D.W.; Murakami, M.; Schivell, J.; Towner, H.H.

    1988-12-01

    The scale lengths of the electron density (L/sub n//sub e/), temperature (L/sub T//sub e/), and pressure (L/sub p//sub e/) gradients were investigated during the 1985 operating period of the Tokamak Fusion Test Reactor (TFTR) for gas-fueled plasmas with neutral beam injection and movable limiter. Although the global energy confinement time degrades as the heating power increases or the plasma current decreases, the radial profiles of the scale lengths (L/sub T//sub e/ and L/sup p//sub e/) remain unchanged. Especially, the electron pressure profile is constrained not to change. This trend appears to hold over a fairly wide range of TFTR operational regimes. The radial profiles of L/sub n//sub e/ and /eta//sub e/ (= L/sub n//sub e//L/sub T//sub e/) also appear to remain unchanged, although the uncertainties of the experimental data for these quantities are greater than those for L/sub T//sub e/ and L/sub p//sub e/. The experimental parameters are used to evaluate theoretical predictions of the electron thermal diffusivity, and the results are compared with the empirical thermal diffusivity. 34 refs., 18 figs., 2 tabs.

  20. Different Mechanisms for Establishing Liquid Walls in Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Hançerlioğullari, Aybaba; Cini, Mesut

    2013-04-01

    The APEX study is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around a fusion plasma. In this study the modeling of APEX hybrid reactor produced by using ARIES-RS hybrid reactor technology, was performed by using the Monte Carlo code and ENF/B-V-VI nuclear data. The most important feature of APEX hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity, good power transformation productivity the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. Around the fusion chamber, molten salt Li2BeF4 and natural lithium were used as cooling materials. The result of the study indicated that fissile material production UF4 and ThF4 heavy metal salt increased nearly at the same percentage.

  1. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

    NASA Astrophysics Data System (ADS)

    Hançerliogulları, Aybaba; Cini, Mesut

    2013-10-01

    In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

  2. Compact Fusion Advanced Rankine (CFARII) power cycle---Operating regimes

    SciTech Connect

    Logan, B.G.

    1991-09-30

    Performance (cost/kWe and efficiency) of generic Compact Fusion Advanced Rankine (CFARII) power conversion is investigated for various working fluids, operating temperatures and pressures, and thermal power levels. A general conclusion is that good CFARII performance is found for a remarkably broad range of materials, temperatures, pressures and power levels, which gives considerable flexibility to future design studies which may apply CFARII energy conversion to specific fusion energy sources such as ICF, MICF, and Mini-PACER. 5 refs, 7 figs., 2 tabs.

  3. Applying design principles to fusion reactor configurations for propulsion in space

    NASA Technical Reports Server (NTRS)

    Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

    1993-01-01

    The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. We applied three design principles (DP's) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS).

  4. Advances in U.S. Heavy Ion Fusion Science

    SciTech Connect

    Barnard, JJ; Logan, B.G.; Barnard, J.J.; Bieniosek, F.M.; Cohen, R.H.; Coleman, J.E.; Davidson, R.C.; Efthimion, P.C.; Friedman, A.; Gilson, E.P.; Grisham, L.R.; Grote, D.P.; Henestroza, E.; Kaganovich, I.D.; Kireeff-Covo, M.; Lee, E.P.; Leitner, M.A.; Lund, S.M.; Molvik, A.W.; Ni, P.; Perkins, L.J.; Qin, H.; Roy, P.K.; Sefkow, A.B.; Seidl, P.A.; Startsev, E.A.; Waldron, W.L.

    2007-09-03

    During the past two years, the US heavy ion fusion science program has made significant experimental and theoretical progress in simultaneous transverse and longitudinal beam compression, ion-beam-driven warm dense matter targets, high-brightness beam transport, advanced theory and numerical simulations, and heavy ion target physics for fusion. First experiments combining radial and longitudinal compression {pi} of intense ion beams propagating through background plasma resulted in on-axis beam densities increased by 700X at the focal plane. With further improvements planned in 2008, these results enable initial ion beam target experiments in warm dense matter to begin next year. They are assessing how these new techniques apply to higher-gain direct-drive targets for inertial fusion energy.

  5. Advances in U.S. Heavy Ion Fusion Science

    SciTech Connect

    Logan, B.G.; Barnard, J.J.; Bieniosek, F.M.; Cohen, R.H.; Coleman, J.E.; Davidson, R.C.; Efthimion, P.C.; Friedman, A.; Gilson, E.P.; Grisham, L.R.; Grote, D.P.; Henestroza, E.; Kaganovich, I.D.; Kireeff-Covo, M.; Lee, E.P.; Leitner, M.A.; Lund, S.M.; Molvik, A.W.; Ni, P.; Perkins, L. J.; Qin, H.; Roy, P.K.; Sefkow, A.B.; Seidl, P.A.; Startsev, E.A.; Waldron, W.L.

    2007-09-01

    During the past two years, the US heavy ion fusion science program has made significant experimental and theoretical progress in simultaneous transverse and longitudinal beam compression, ion-beam-driven warm dense matter targets, high-brightness beam transport, advanced theory and numerical simulations, and heavy ion target physics for fusion. First experiments combining radial and longitudinal compression {pi} of intense ion beams propagating through background plasma resulted in on-axis beam densities increased by 700X at the focal plane. With further improvements planned in 2008, these results enable initial ion beam target experiments in warm dense matter to begin next year. They are assessing how these new techniques apply to higher-gain direct-drive targets for inertial fusion energy.

  6. Development of a system model for advanced small modular reactors.

    SciTech Connect

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). 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 (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  7. ANDES Measurements for Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Plompen, A. J. M.; Hambsch, F.-J.; Kopecky, S.; Nyman, M.; Rouki, C.; Salvador Castiñeira, P.; Schillebeeckx, P.; Belloni, F.; Berthoumieux, E.; Gunsing, F.; Lampoudis, C.; Calviani, M.; Guerrero, C.; Cano-Ott, D.; Gonzalez Romero, E.; Aïche, M.; Jurado, B.; Mathieu, L.; Derckx, X.; Farget, F.; Rodrigues Tajes, C.; Bacquias, A.; Dessagne, Ph.; Kerveno, M.; Borcea, C.; Negret, A.; Colonna, N.; Goncalves, I.; Penttilä, H.; Rinta-Antila, S.; Kolhinen, V. S.; Jokinen, A.

    2014-05-01

    A significant number of new measurements was undertaken by the ANDES “Measurements for advanced reactor systems” initiative. These new measurements include neutron inelastic scattering from 23Na, Mo, Zr, and 238U, neutron capture cross sections of 238U, 241Am, neutron induced fission cross sections of 240Pu, 242Pu, 241Am, 243Am and 245Cm, and measurements that explore the limits of the surrogate technique. The latter study the feasibility of inferring neutron capture cross sections for Cm isotopes, the neutron-induced fission cross section of 238Pu and fission yields and fission probabilities through full Z and A identification in inverse kinematics for isotopes of Pu, Am, Cm and Cf. Finally, four isotopes are studied which are important to improve predictions for delayed neutron precursors and decay heat by total absorption gamma-ray spectrometry (88Br, 94Rb, 95Rb, 137I). The measurements which are performed at state-of-the-art European facilities have the ambition to achieve the lowest possible uncertainty, and to come as close as is reasonably achievable to the target uncertainties established by sensitivity studies. An overview is presented of the activities and achievements, leaving detailed expositions to the various parties contributing to the conference.

  8. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

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

    SciTech Connect

    Jones, R.H. ); Lucas, G.E. )

    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.

  10. Mirror Advanced Reactor Study (MARS): executive summary and overview

    SciTech Connect

    Logan, B.G.; Perkins, L.J.; Gordon, J.D.

    1984-07-01

    Two self-consistent MARS configurations are discussed - a 1200-MWe commercial electricity-generating plant and a synguels-generating plant that produces hydrogen with an energy equivalent to 26,000 barrels of oil per day. The MARS machine emphasizes the attractive features of the tandem mirror concept, including steady-state operation, a small-diameter high-beta plasma, a linear central cell with simple low-maintenance blankets, low first-wall heat fluxes (<10 W/cm/sup 2/), no driven plasma currents or associated disruptions, natural halo impurity diversion, and direct conversion of end-loss charged-particle power. The MARS electric plant produces 2600 MW of fusion power in a 130-m-long central cell. Advanced tandem-mirror plasma-engineering concepts, a high-efficiency liquid lithium-lead (Li/sub 17/Pb/sub 83/) blanket, and efficient direct electrical conversion of end loss power combine to produce a high net plant efficiency of 36%. With a total capital cost of $2.9 billion (constant 1983 dollars), the MARS electric plant produces busbar electricity at approx. 7 cents/kW-hour. The MARS synfuels plant produces 3500 MW of fusion power in a 150-m-long central cell. A helium-gas-cooled silicon carbide pebble-bed blanket provides high-temperature (1000/sup 0/C) heat to a thermochemical water-splitting cycle and the resulting hydrogen is catalytically converted to methanol for distribution. With a total capital cost of $3.6 billion (constant 1983 dollars), the synfuels plant produces methanol fuel at about $1.7/gal. The major features of the MARS reactor include sloshing-ion thermal barrier plugs for efficient plasma confinement, a high efficiency blanket, high-field (24-T) choke cells, drift pumping for trapped plasma species, quasi-optical electron-cyclotron resonant heating (ECRH) systems, and a component gridless direct converter.

  11. Cladding and Duct Materials for Advanced Nuclear Recycle Reactors

    SciTech Connect

    Allen, Todd R.; Busby, Jeremy T; Klueh, Ronald L; Maloy, S; Toloczko, M

    2008-01-01

    The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP s advanced nuclear recycle reactors program.

  12. Cladding and duct materials for advanced nuclear recycle reactors

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

    2008-01-01

    The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

  13. Aerosol Resuspension Model for MELCOR for Fusion and Very High Temperature Reactor Applications

    SciTech Connect

    B.J. Merrill

    2011-01-01

    Dust is generated in fusion reactors from plasma erosion of plasma facing components within the reactor’s vacuum vessel (VV) during reactor operation. This dust collects in cooler regions on interior surfaces of the VV. Because this dust can be radioactive, toxic, and/or chemically reactive, it poses a safety concern, especially if mobilized by the process of resuspension during an accident and then transported as an aerosol though out the reactor confinement building, and possibly released to the environment. A computer code used at the Idaho National Laboratory (INL) to model aerosol transport for safety consequence analysis is the MELCOR code. A primary reason for selecting MELCOR for this application is its aerosol transport capabilities. The INL Fusion Safety Program (FSP) organization has made fusion specific modifications to MELCOR. Recent modifications include the implementation of aerosol resuspension models in MELCOR 1.8.5 for Fusion. This paper presents the resuspension models adopted and the initial benchmarking of these models.

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

    SciTech Connect

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

    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.

  15. HYLIFE-II inertial confinement fusion reactor design

    NASA Astrophysics Data System (ADS)

    Moir, R. W.

    1990-12-01

    The HYLIFE-2 inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-1 used liquid lithium. HYLIFE 2 avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li2, BeF4) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-1. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-1, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW times h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost.

  16. HYLIFE-II inertial confinement fusion reactor design

    SciTech Connect

    Moir, R.W.

    1990-12-14

    The HYLIFE-2 inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-1 used liquid lithium. HYLIFE 2 avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li{sub 2}BeF{sub 4}) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-1. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-1, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW{center dot}h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost. 15 refs., 9 figs., 3 tabs.

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

    SciTech Connect

    Not Available

    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.

  18. OPERATION OF FUSION REACTORS IN ONE ATMOSPHERE OF AIR INSTEAD OF VACUUM SYSTEMS

    SciTech Connect

    Roth, J. Reece

    2009-07-26

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Wang, Shijia Wang, Shaojie

    2015-04-15

    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.

  1. A Subcritical, Gas-Cooled Fast Transmutation Reactor with a Fusion Neutron Source

    SciTech Connect

    Stacey, W.M.; Beavers, V.L.; Casino, W.A.; Cheatham, J.R.; Friis, Z.W.; Green, R.D.; Hamilton, W.R.; Haufler, K.W.; Hutchinson, J.D.; Lackey, W.J.; Lorio, R.A.; Maddox, J.W.; Mandrekas, J.; Manzoor, A.A.; Noelke, C.A.; Oliveira, C. de; Park, M.; Tedder, D.W.; Terry, M.R.; Hoffman, E.A.

    2005-05-15

    A design is presented for a subcritical, He-cooled fast reactor, driven by a tokamak D-T fusion neutron source, for the transmutation of spent nuclear fuel (SNF). The reactor is fueled with coated transuranic (TRU) particles and is intended for the deep-burn (>90%) transmutation of the TRUs in SNF without reprocessing of the coated fuel particles. The reactor design is based on the materials, fuel, and separations technologies under near-term development in the U.S. Department of Energy (DOE) Nuclear Energy Program and on the plasma physics and fusion technologies under near-term development in the DOE Fusion Energy Sciences Program, with the objective of intermediate-term ({approx}2040) deployment. The physical and performance characteristics and research and development requirements of such a reactor are described.

  2. Applying design principles to fusion reactor configurations for propulsion in space

    NASA Technical Reports Server (NTRS)

    Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

    1993-01-01

    We applied three design principles (DPs) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: (1) provide maximum direct access to space for waste radiation, (2) operate components as passive radiators to minimize cooling-system mass, and (3) optimize the plasma fuel, fuel mix, and temperature for best specific Jet power. The three candidate-terrestrial-fusion-reactor configurations are: (1) the thermal-barrier-tandem-mirror (TBTM), (2) field-reversed-mirror (FRM), and (3) levitated-dipole-field (LDF). The resulting three candidate-space-fusion-propulsion systems have their initial-mass-to-LEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System.

  3. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    NASA Astrophysics Data System (ADS)

    Indah Rosidah, M.; Suud, Zaki; Yazid, Putranto Ilham

    2015-09-01

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  4. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    SciTech Connect

    Indah Rosidah, M. Suud, Zaki; Yazid, Putranto Ilham

    2015-09-30

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  5. Conceptual study of fusion-driven transmutation reactor with ITER physics and engineering constraints

    NASA Astrophysics Data System (ADS)

    Hong, Bong

    2011-10-01

    A conceptual study of fusion-driven transmutation reactor was performed based on ITER physics and engineering constraints. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

  6. Shielding considerations for advanced space nuclear reactor systems

    SciTech Connect

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  7. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

    DOE PAGES

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; et al

    2016-09-14

    A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, whichmore » we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less

  8. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

    NASA Astrophysics Data System (ADS)

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

    2016-12-01

    A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

  9. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    SciTech Connect

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.

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

    SciTech Connect

    Rule, K.; Scott, J.; Larson, S.

    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.

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

    SciTech Connect

    Not Available

    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. Lithium Ceramic Blankets for Russian Fusion Reactors and Influence of Breeding Operation Mode on Parameters of Reactor Tritium Systems

    SciTech Connect

    Kapyshev, Victor K.; Chernetsov, Mikhail Yu.; Zhevotov, Sergej I.; Kersnovskij, Alexandr Yu.; Kolbasov, Boris N.; Kovalenko, Victor G.; Paltusov, Nikolaj P.; Sernyaev, Georgeij A.; Sterebkov, Juri S.; Zyryanov, Alexej P.

    2005-07-15

    Russian controlled fusion program supposes development of a DEMO reactor design and participation in ITER Project. A solid breeder blanket of DEMO contains a ceramic lithium orthosilicate breeder and a beryllium multiplier. Test modules of the blanket are developed within the scope of ITER activities. Experimental models of module tritium breeding zones (TBZ), materials and fabrication technology of the TBZ, tritium reactor systems to analyse and process gas released from lithium ceramics are being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ have been designed, manufactured and tested in IVV-2M nuclear reactor. Initial results of the in-pile experiments and outcome of lithium ceramics irradiation in a water-graphite nuclear reactor are considered to be a data base for development of the test modules and initial requirements for DEMO tritium system design. Influence of the tritium release parameters and hydrogen concentration in a purge gas on parameters of reactor system are discussed.

  13. Nuclear fuels for low-beta fusion reactors: Lithium resources revisited

    SciTech Connect

    Eckhartt, D.

    1995-12-01

    In searching to attain optimum conditions for the controlled release of nuclear energy by fusion processes, the stationary confinement of low-pressure ring-shaped plasmas by strong magnetic fields is now regarded as the most promising approach. The author considers a number of fuel combinations that could be operated in such low-beta reactor systems and looks upon the relevant fuel reserves. The {open_quotes}classical{close_quotes} D-T-Li cycle will be used as a standard and is extensively discussed therefore. It could supply most of mankind`s future long-term power needs - but only on condition that the required lithium fuel can be extracted from seawater at reasonable expenses. The estimated land-bound lithium reserves are too small to that end, they will last for about 500 years at most, depending on forecasts of future energy consumption and on assumptions about exploitable resources. Recovery of lithium from seawater would extend the possible range by a factor of 300 or so, provided that extraction technologies which are at present available in the laboratory, could be extended to a very large and industrial scale. Deuterium is abundant on earth but D-D fusion is difficult, if not impossible, to be achieved in the low-beta systems presently investigated for D-T fusion. The same arguments apply to so-called {open_quotes}advanced{close_quotes} concepts, such as the D-{sup 3}He and the D-{sup 6}Li cycles. 37 refs., 1 fig., 1 tab.

  14. Progress on Gyrotrons for ITER and Future Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Thumm, Manfred K.

    2009-11-01

    The prototype of the Japan 170 GHz ITER gyrotron holds the energy and efficiency world record of 2.88 GJ (0.8 MW, 3600 s, 57%) with 55% efficiency at 1 MW, 800 s, whereas the Russian 170 GHz ITER prototype tube achieved 0.83 MW with a pulse duration of 203 s at 48% efficiency and 1 MW at 116 s and 52%. The record parameters of the European megawatt-class 140 GHz gyrotron for the Stellarator Wendelstein W7-X are: 0.92 MW output power at 1800 s pulse duration, almost 45% efficiency and 97.5% Gaussian mode purity. All these gyrotrons employ a cylindrical cavity, a quasi-optical output coupler, a synthetic diamond window and a single-stage depressed collector (SDC) for energy recovery. In coaxial cavities the existence of the longitudinally corrugated inner conductor reduces the problems of mode competition and limiting current, thus allowing one to use even higher order modes with lower Ohmic attenuation than in cylindrical cavities. Synthetic diamond windows with a transmission capability of 2 MW, continuous wave (CW) are feasible. In order to keep the number of the required gyrotrons and magnets as low as possible, to reduce the costs of the ITER 26 MW, 170 GHz ECRH system and to allow compact upper launchers for plasma stabilization, 2 MW mm-wave power per gyrotron tube is desirable. The FZK pre-prototype tube for an EU 170 GHz, 2 MW ITER gyrotron has achieved 1.8 MW at 28% efficiency (without depressed collector). Design studies for a 4 MW 170 GHz coaxial-cavity gyrotron with two synthetic diamond output windows and two 2 MW mm-wave output beams for future fusion reactors are currently being performed at FZK. The availability of sources with fast frequency tunability (several GHz s-1, tuning in 1.5-2.5% steps for about ten different frequencies) would permit the use of a simple, fixed, non-steerable mirror antenna for local current drive (ECCD) experiments and plasma stabilization. GYCOM in Russia develops in collaboration with IPP Garching and FZK an industrial

  15. Designing a Gas Test Loop for the Advanced Test Reactor

    SciTech Connect

    James R. Parry

    2005-11-01

    The Generation IV Reactor Program and the Advanced Fuel Cycle Initiative are investigating some new reactor concepts which require extensive materials and fuels testing in a fast neutron spectrum. The capability to test materials and fuels in a fast neutron flux in the United States is very limited to non-existent. It has been proposed to install a gas test loop (GTL) in one of the lobes of the Advanced Test Reactor (ATR) at the Idaho National Laboratory and harden the spectrum to provide some fast neutron flux testing capabilities in the United States. This paper describes the neutronics investigation into the design of the GTL for the ATR.

  16. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-08-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  17. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  18. Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

  19. Fusion

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

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

    2015-12-15

    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.

  2. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, George P.

    1988-01-01

    A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

  3. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1987-02-20

    A high-power-density-laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems. 25 figs.

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

    SciTech Connect

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

    1986-04-01

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

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

    SciTech Connect

    Not Available

    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.

  6. Coatings and claddings for the reduction of plasma contamination and surface erosion in fusion reactors

    SciTech Connect

    Kaminsky, M.

    1980-01-01

    For the successful operation of plasma devices and future fusion reactors it is necessary to control plasma impurity release and surface erosion. Effective methods to obtain such controls include the application of protective coatings to, and the use of clad materials for, certain first wall components. Major features of the development programs for coatings and claddings for fusion applications will be described together with an outline of the testing program. A discussion of some pertinent test results will be included.

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

    NASA Astrophysics Data System (ADS)

    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.

  8. Advanced light water reactor requirements document: Chapter 3, Reactor coolant system and reactor non-safety auxiliary systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Coolant System and the Reactor Non-safety Auxiliary Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the reactor coolant system and reactor non-safety auxiliary systems for Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Non-safety auxiliaries include systems which are required for normal operation of the plant but are not required to operate for accident mitigation or to bring the plant to a safe shutdown condition. For PWRs, the reactor coolant system, steam generator system, chemical and volume control system and boron recycle system are included. For BWRs, the reactor coolant system and reactor water cleanup system are included. The chapter also includes requirements for the above systems which are common to BWRs and PWRs and requirements for process sampling for BWRs and PWRs.

  9. X-ray-ablated plumes in inertial confinement fusion reactors

    NASA Astrophysics Data System (ADS)

    Scott, John Mitchell

    Modeling of inertial confinement fusion (ICF) target chamber phenomena presents researchers with various technical problems requiring creative solutions. In particular, the wide ranging physical and time scales of the problem give special difficulty when modeling one shot cycle of an ICF target chamber. Ultimately, the goal of the modeling effort is a unified model beginning with target injection and ending with condensation of the vaporized debris. The work here develops a combined gas dynamics/X-ray ablation model used to predict the response of materials to X-ray emissions from ICF targets. This model in combination with experiments performed at the Nova facility at Lawrence Livermore National Laboratory (LLNL) aided in the design effort for the first wall of the National Ignition Facility (NIF). The phenomena inside an ICF target chamber include the fusion bum of a D-T fuel capsule enclosed in a hohlraum leading to the emission of neutrons, debris, and X rays. The X rays emitted from the target deposit on target facing surfaces, heating and vaporizing the surface layers of the material. The vapor plume generated will travel through the chamber and deposit on various other surfaces. For the NIF and other ICF laser facilities, modeling of these X-ray ablated plumes is important to ascertain the performance of the first wall surface of the target chamber. The first wall must be designed to minimize contamination to laser optics that interface with the target chamber. For this work, experiments were performed to assess the performance of materials at X-ray fluences expected at the NIF first wall. These experiments included long-term exposure of potential target chamber materials, the X-ray response of stainless steel, and a louvered geometry experiment to aid in the assessment of the geometrical design and material selection of the first wall. The results of these experiments show that boron carbide and stainless steel will both perform adequately during facility

  10. SAFIRE: A systems analysis code for ICF (inertial confinement fusion) reactor economics

    SciTech Connect

    McCarville, T.J.; Meier, W.R.; Carson, C.F.; Glasgow, B.B.

    1987-01-12

    The SAFIRE (Systems Analysis for ICF Reactor Economics) code incorporates analytical models for scaling the cost and performance of several inertial confinement fusion reactor concepts for electric power. The code allows us to vary design parameters (e.g., driver energy, chamber pulse rate, net electric power) and evaluate the resulting change in capital cost of power plant and the busbar cost of electricity. The SAFIRE code can be used to identify the most attractive operating space and to identify those design parameters with the greatest leverage for improving the economics of inertial confinement fusion electric power plants.

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

    SciTech Connect

    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.

  12. Development of the cascade inertial-confinement-fusion reactor

    SciTech Connect

    Pitts, J.H.

    1985-04-15

    Cascade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670/sup 0/K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.

  13. Development of the cascade inertial-confinement-fusion reactor

    SciTech Connect

    Pitts, J.H.

    1985-07-01

    Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.

  14. Advanced Reactors Thermal Energy Transport for Process Industries

    SciTech Connect

    P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

  15. Advances in data representation for hard/soft information fusion

    NASA Astrophysics Data System (ADS)

    Rimland, Jeffrey C.; Coughlin, Dan; Hall, David L.; Graham, Jacob L.

    2012-06-01

    Information fusion is becoming increasingly human-centric. While past systems typically relegated humans to the role of analyzing a finished fusion product, current systems are exploring the role of humans as integral elements in a modular and extensible distributed framework where many tasks can be accomplished by either human or machine performers. For example, "participatory sensing" campaigns give humans the role of "soft sensors" by uploading their direct observations or as "soft sensor platforms" by using mobile devices to record human-annotated, GPS-encoded high quality photographs, video, or audio. Additionally, the role of "human-in-the-loop", in which individuals or teams using advanced human computer interface (HCI) tools such as stereoscopic 3D visualization, haptic interfaces, or aural "sonification" interfaces can help to effectively engage the innate human capability to perform pattern matching, anomaly identification, and semantic-based contextual reasoning to interpret an evolving situation. The Pennsylvania State University is participating in a Multi-disciplinary University Research Initiative (MURI) program funded by the U.S. Army Research Office to investigate fusion of hard and soft data in counterinsurgency (COIN) situations. In addition to the importance of this research for Intelligence Preparation of the Battlefield (IPB), many of the same challenges and techniques apply to health and medical informatics, crisis management, crowd-sourced "citizen science", and monitoring environmental concerns. One of the key challenges that we have encountered is the development of data formats, protocols, and methodologies to establish an information architecture and framework for the effective capture, representation, transmission, and storage of the vastly heterogeneous data and accompanying metadata -- including capabilities and characteristics of human observers, uncertainty of human observations, "soft" contextual data, and information pedigree

  16. Plasma Heating and Current Drive for Fusion Reactors

    NASA Astrophysics Data System (ADS)

    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. )

  17. Design considerations of the irradiation test vehicle for the advanced test reactor

    SciTech Connect

    Tsai, H.; Gomes, I.C.; Smith, D.L.

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  18. Advanced development of immobilized enzyme reactors

    NASA Technical Reports Server (NTRS)

    Jolly, Clifford D.; Schussel, Leonard J.; Carter, Layne

    1991-01-01

    Fixed-bed reactors have been used at NASA-Marshall to purify wastewater generated by an end-use equipment facility, on the basis of a combination of multifiltration unibeds and enzyme unibeds. The enzyme beds were found to effectively remove such targeted organics as urea, alcohols, and aldehydes, down to levels lying below detection limits. The enzyme beds were also found to remove organic contaminants not specifically targeted.

  19. Utilization of Heavy Metal Molten Salts in the ARIES-RS Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Übeyli, Mustafa; Yapıcı, Hüseyin

    2008-09-01

    ARIES-RS is one of the major magnetic fusion energy reactor designs that uses a blanket having vanadium alloy structure cooled by lithium [1, 2]. It is a deuterium-tritium (DT) fusion driven reactor, having a fusion power of 2170 MW [1, 2]. This study presents the neutronic analysis of the ARIES-RS fusion reactor using heavy metal molten salts in which Li2BeF4 as the main constituent was mixed with increased mole fractions of heavy metal salt (ThF4 or UF4) starting by 2 mol.% up to 12 mol.%. Neutron transport calculations were carried out with the help of the SCALE 4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S 8- P 3 approximation. According to the numerical results, tritium self-sufficiency was attained for the coolants, Flibe with 2% UF4 or ThF4 and 4% UF4. In addition, higher energy multiplication values were found for the salt with UF4 compared to that with ThF4. Furthermore, significant amount of high quality nuclear fuel was produced to be used in external reactors.

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

    SciTech Connect

    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.

  1. The development of a universal diagnostic probe system for Tokamak fusion test reactor

    NASA Technical Reports Server (NTRS)

    Mastronardi, R.; Cabral, R.; Manos, D.

    1982-01-01

    The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Not Available

    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.

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

    SciTech Connect

    Not Available

    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.

  7. Reduction in TFTR (Tokamak Fusion Test Reactor) fusion reaction rate by unbalanced beam injection and rotation

    SciTech Connect

    Hendel, H.W.; Jassby, D.L.; Bitter, M.L.; Taylor, G.

    1987-06-01

    In TFTR plasmas at low to moderate density, the highest fusion energy gain Q/sub dd/ (D-D fusion power/injected power P/sub b/) is obtained with nearly balanced co- and counter-injection of neutral beams. For a given beam power, significantly unbalanced injection reduces Q/sub dd/ because the accompanying plasma rotation reduces the beam-target fusion reactivity, the fast-ion slowing-down time, and the beam-beam reaction rate, while and decrease from their maximum values. 9 refs., 3 figs., 1 tab.

  8. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    NASA Astrophysics Data System (ADS)

    Andrienko, O. S.; Egorov, N. B.; Zherin, I. I.; Indyk, D. V.

    2016-01-01

    The paper describes a spectrophotometric procedure for fast reactor advanced coolant manufacture control. The molar absorption coefficient of dimethyllead dibromide with dithizone was defined as equal to 68864 ± 795 l·mole-1·cm-1, limit of detection as equal to 0.583 · 10-6 g/ml. The spectrophotometric procedure application range was found to be equal to 37.88 - 196.3 g. of dimethyllead dibromide in the sample. The procedure was used within the framework of the development of the method of synthesis of the advanced coolant for fast reactors.

  9. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1983-09-29

    The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

  10. The First Decommissioning of a Fusion Reactor Fueled by Deuterium-Tritium

    SciTech Connect

    Charles A. Gentile; Erik Perry; Keith Rule; Michael Williams; Robert Parsells; Michael Viola; James Chrzanowski

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Plasma Physics Laboratory of Princeton University (PPPL) was the first fusion reactor fueled by a mixture of deuterium and tritium (D-T) to be decommissioned in the world. The decommissioning was performed over a period of three years and was completed safely, on schedule, and under budget. Provided is an overview of the project and detail of various factors which led to the success of the project. Discussion will cover management of the project, engineering planning before the project started and during the field work as it was being performed, training of workers in the field, the novel adaptation of tools from other industry, and the development of an innovative process for the use of diamond wire to segment the activated/contaminated vacuum vessel. The success of the TFTR decommissioning provides a viable model for the decommissioning of D-T burning fusion devices in the future.

  11. Advanced Test Reactor Capabilities and Future Irradiation Plans

    SciTech Connect

    Frances M. Marshall

    2006-10-01

    The Advanced Test Reactor (ATR), located at the Idaho National Laboratory (INL), is one of the most versatile operating research reactors in the Untied States. The ATR has a long history of supporting reactor fuel and material research for the US government and other test sponsors. The INL is owned by the US Department of Energy (DOE) and currently operated by Battelle Energy Alliance (BEA). The ATR is the third generation of test reactors built at the Test Reactor Area, now named the Reactor Technology Complex (RTC), whose mission is to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The current experiments in the ATR are for a variety of customers--US DOE, foreign governments and private researchers, and commercial companies that need neutrons. The ATR has several unique features that enable the reactor to perform diverse simultaneous tests for multiple test sponsors. The ATR has been operating since 1967, and is expected to continue operating for several more decades. The remainder of this paper discusses the ATR design features, testing options, previous experiment programs, future plans for the ATR capabilities and experiments, and some introduction to the INL and DOE's expectations for nuclear research in the future.

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

    SciTech Connect

    Garner, F.A.; Greenwood, L.R.; 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.

  13. Volatility from copper and tungsten alloys for fusion reactor applications

    SciTech Connect

    Smolik, G.R.; Neilson, R.M. Jr.; Piet, S.J. )

    1989-01-01

    Accident scenarios for fusion power plants present the potential for release and transport of activated constituents volatilized from first wall and structural materials. The extent of possible mobilization and transport of these activated species, many of which are oxidation driven'', is being addressed by the Fusion Safety Program at the Idaho National Engineering Laboratory (INEL). This report presents experimental measurements of volatilization from a copper alloy in air and steam and from a tungsten alloy in air. The major elements released included zinc from the copper alloy and rhenium and tungsten from the tungsten alloy. Volatilization rates of several constituents of these alloys over temperatures ranging from 400 to 1200{degree}C are presented. These values represent release rates recommended for use in accident assessment calculations. 8 refs., 3 figs., 5 tabs.

  14. Advanced sodium fast reactor accident source terms :

    SciTech Connect

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  15. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2004-10-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  16. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    SciTech Connect

    Grover, S.B.

    2004-10-06

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  17. Foundational development of an advanced nuclear reactor integrated safety code.

    SciTech Connect

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  18. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    SciTech Connect

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  19. A wall-crawling robot for reactor vessel inspection in advanced reactors

    SciTech Connect

    Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

    1994-06-01

    A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected.

  20. Health Monitoring to Support Advanced Small Modular Reactors

    SciTech Connect

    Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (aSMRs) are based on advanced reactor concepts, some of which were promoted by the Generation IV International Forum, and are being considered for diverse missions including desalination of water, production of hydrogen, etc. While the existing fleet of commercial nuclear reactors provides baseload electricity, it is conceivable that aSMRs could be implemented for both baseload and load following applications. The effect of diverse operating missions and unit modularity on plant operations and maintenance (O&M) is not fully understood and limiting these costs will be essential to successful deployment of aSMRs. Integrated health monitoring concepts are proposed to support the safe and affordable operation of aSMRs over their lifetime by enabling management of significant in-vessel and in-containment active and passive components.

  1. Advanced Neutron Source Reactor thermal analysis of fuel plate defects

    SciTech Connect

    Giles, G.E.

    1995-08-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor designed to provide the highest continuous neutron beam intensity of any reactor in the world. The present technology for determining safe operations were developed for the High Flux Isotope Reactor (HFIR). These techniques are conservative and provide confidence in the safe operation of HFIR. However, the more intense requirements of ANSR necessitate the development of more accurate, but still conservative, techniques. This report details the development of a Local Analysis Technique (LAT) that provides an appropriate approach. Application of the LAT to two ANSR core designs are presented. New theories of the thermal and nuclear behavior of the U{sub 3}Si{sub 2} fuel are utilized. The implications of lower fuel enrichment and of modifying the inspection procedures are also discussed. Development of the computer codes that enable the automate execution of the LAT is included.

  2. Astronaut Kevin Chilton works with advanced cell reactor

    NASA Technical Reports Server (NTRS)

    1994-01-01

    Astronaut Kevin P. Chilton, pilot, works with an advanced cell reactor, which incorporated the first ever videomicroscope, on the Space Tissue Loss (STL-B) experiment on the Space Shuttle Endeavour's middeck. This experiment studied cell growth during the STS-59 mission.

  3. Neutron Cross Section Covariances: Recent Workshop and Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Oblozinsky, Pavel

    2008-10-01

    The recent Workshop on Neutron Cross Section Covariances, organized by BNL and attended by more than 50 scientists, responded to demands of many user groups, including advanced reactor systems, for uncertainty and correlation information. These demands can be explained by considerable progress in advanced neutronics simulation that probe covariances and their impact on design and operational margins of nuclear systems. The Workshop addressed evaluation methodology, recent evaluations as well as user's perspective, marking era of revival of covariance development that started some two years ago. We illustrate urgent demand for covariances in the case of advanced reactor systems, including fast actinide burner under GNEP, new generation of power reactors, Gen-IV, and reactors under AFCI. A common feature of many of these systems is presence of large amount of minor actinides and fission products that require improved nuclear data. Advanced simulation codes rely on quality input, to be obtained by adjusting the data library, such as the new ENDF/B-VII.0, by considering integral experiments as currently pursued by GNEP. To this end the nuclear data community is developing covariances for formidable amount of 112 materials (isotopes).

  4. Inertial fusion energy power reactor fuel recovery system

    SciTech Connect

    Gentile, C. A.; Kozub, T.; Langish, S. W.; Ciebiera, L. P.; Nobile, A.; Wermer, J.; Sessions, K.

    2008-07-15

    A conceptual design is proposed to support the recovery of un-expended fuel, ash, and associated post-detonation products resident in plasma exhaust from a {approx}2 GWIFE direct drive power reactor. The design includes systems for the safe and efficient collection, processing, and purification of plasma exhaust fuel components. The system has been conceptually designed and sized such that tritium bred within blankets, lining the reactor target chamber, can also be collected, processed, and introduced into the fuel cycle. The system will nominally be sized to process {approx}2 kg of tritium per day and is designed to link directly to the target chamber vacuum pumping system. An effort to model the fuel recovery system (FRS) using the Aspen Plus engineering code has commenced. The system design supports processing effluent gases from the reactor directly from the exhaust of the vacuum pumping system or in batch mode, via a buffer vessel in the Receiving and Analysis System. Emphasis is on nuclear safety, reliability, and redundancy as to maximize availability. The primary goal of the fuel recovery system design is to economically recycle components of direct drive IFE fuel. The FRS design is presented as a facility sub-system in the context of supporting the larger goal of producing safe and economical IFE power. (authors)

  5. Advanced high-temperature, high-pressure transport reactor gasification

    SciTech Connect

    Swanson, M.L.

    1999-07-01

    The mission of the U.S. Department of Energy's (DOE's) Federal Energy Technology Center Office of Power Systems Product Management is to foster the development and deployment of advanced, clean, and affordable fossil-based (coal) power systems. These advanced power systems include the development and demonstration of gasification-based advanced power systems. These systems are integral parts of the Vision 21 Program for the co-production of power and chemicals which is being developed at DOE. DOE has been developing advanced gasification systems which lower the capital and operating cost of producing syngas for electricity or chemicals production. A transport reactor gasifier has shown potential to be a low-cost syngas producer as compared to other gasification systems because of its high throughput. This work directly supports the Power Systems Development Facility (PSDF) utilizing the Kellogg, Brown and Root (KBR) transport reactor located at the Southern Company Services (SCS) Wilsonville, Alabama, site. Over 1000 hours of operation on three different fuels in the pilot-scale transport reactor development unit (TRDU) has been completed to date. The Energy and Environmental Research Center (EERC) has established an extensive database on the operation of various fuels in a transport reactor gasifier. This database will be useful in determining the effectiveness of design changes on a transport reactor gasifier. It has been demonstrated that corrected fuel gas heating values ranging between 105 to 130 Btu/scf can be achieved. Factors that affect the TRDU product gas quality appear to be circulation rate, coal type, temperature, and air:coal and steam:coal ratios. Future plans are to modify the transport reactor mixing zone and J-leg loop seal to increase backmixing, thereby increasing solids residence time and gasifier performance. Enriched air- and oxygen-blown gasification tests, especially on widely available low-cost fuels such as petroleum coke, will also be

  6. Large orbit magnetic confinement systems for advanced fusion fuels

    SciTech Connect

    Rostoker, N.

    1992-01-01

    The objective of the grant/contract was to illuminate the problem of magnetic confinement for plasmas where the majority of ions have large gyro-orbits and do not obey adiabatic particle dynamics. The electrons are adiabatic. We considered a class of equilibria where large orbit ions dominate; the equilibria are rigorous solutions of the Vlasov/Maxwell equations. For a simple cse -- the infinitely long, low beta, rotating plasma a complete stability analysis was carried out. This problem was the basis of the first paper on finite Larmor radius stabilization. In that paper an expansion in {var epsilon} = {bar {alpha}}{sub i}/r{sub o} was carried out to the first significant order beyond MHD. In this report the same problem is solved to all orders in {var epsilon}. While this case is of limited applicability to experiments it is rigorous and without approximations, so that it can be used to verify approximations to be developed for more complex and useful cases. The application of the results to date to small fusion reactors is described in the second paper which was written after the termination of the contract, but is based in part on material developed during the contract period.

  7. Radiation-induced electrical breakdown of helium in fusion reactor superconducting magnet systems

    SciTech Connect

    Perkins, L.J.

    1983-12-02

    A comprehensive theoretical study has been performed on the reduction of the electrical breakdown potential of liquid and gaseous helium under neutron and gamma radiation. Extension of the conventional Townsend breakdown theory indicates that radiation fields at the superconducting magnets of a typical fusion reactor are potentially capable of significantly reducing currently established (i.e., unirradiated) helium breakdown voltages. Emphasis is given to the implications of these results including future deployment choices of magnet cryogenic methods (e.g., pool-boiling versus forced-flow), the possible impact on magnet shielding requirements and the analogous situation for radiation-induced electrical breakdown in fusion RF transmission systems.

  8. Metal fire implications for advanced reactors. Part 1, literature review.

    SciTech Connect

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.

    2007-10-01

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Technical Reports Server (NTRS)

    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.

  11. Synfuels from fusion: producing hydrogen with the tandem mirror reactor and thermochemical cycles

    SciTech Connect

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

    1981-01-21

    This report examines, for technical merit, the combination of a fusion reactor driver and a thermochemical plant as a means for producing synthetic fuel in the basic form of hydrogen. We studied: (1) one reactor type - the Tandem Mirror Reactor - wishing to use to advantage its simple central cell geometry and its direct electrical output; (2) two reactor blanket module types - a liquid metal cauldron design and a flowing Li/sub 2/O solid microsphere pellet design so as to compare the technology, the thermal-hydraulics, neutronics and tritium control in a high-temperature operating mode (approx. 1200 K); (3) three thermochemical cycles - processes in which water is used as a feedstock along with a high-temperature heat source to produce H/sub 2/ and O/sub 2/.

  12. Preliminary study of fusion reactor: Solution of Grad Shapranov equation

    NASA Astrophysics Data System (ADS)

    Setiawan, Y.; Fermi, N.; Su'ud, Z.

    2012-06-01

    Nuclear fussion is prospective energy sources for the future due to the abundance of the fuel and can be categorized and clean energy sources. The problem is how to contain very hot plasma of temperature few hundreed million degrees safety and reliably. Tokamax type fussion reactors is considered as the most prospective concept. To analyze the plasma confining process and its movement Grad-Shavranov equation must be solved. This paper discuss about solution of Grad-Shavranov equation using Whittaker function. The formulation is then applied to the ITER design and example.

  13. The Consortium for Advanced Simulation of Light Water Reactors

    SciTech Connect

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  14. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect

    Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

    2012-09-15

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  15. Polarized Nuclei in a Simple Mirror Fusion Reactor

    NASA Technical Reports Server (NTRS)

    Noever, David A.

    1995-01-01

    The possibility of enhancing the ratio of output to input power Q in a simple mirror machine by polarizing Deuterium-Tritium (D- T) nuclei is evaluated. Taking the Livermore mirror reference design mirror ratio of 6.54, the expected sin(sup 2) upsilon angular distribution of fusion decay products reduces immediate losses of alpha particles to the loss cone by 7.6% and alpha-ion scattering losses by approx. 50%. Based on these findings, alpha- particle confinement times for a polarized plasma should therefore be 1.11 times greater than for isotropic nuclei. Coupling this enhanced alpha-particle heating with the expected greater than 50% D- T reaction cross section, a corresponding power ratio for polarized nuclei, Q(sub polarized), is found to be 1.63 times greater than the classical unpolarized value Q(sub classical). The effects of this increase in Q are assessed for the simple mirror.

  16. On Enhancing Risk Monitors for Advanced Small Modular Reactors

    SciTech Connect

    Coble, Jamie B.; Coles, Garill A.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (AdvSMRs) can contribute to safe, sustainable, and carbon-neutral energy production. However, the economics of AdvSMRs suffer from the loss of economy-of-scale for both construction and operation. The controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance (O&M) costs. These expenses could potentially be managed through optimized scheduling of O&M activities for components, reactor modules, power blocks, and the full plant. Accurate, real-time risk assessment with integrated health monitoring of key active components can support scheduling of both online and offline inspection and maintenance activities.

  17. Recent advances and issues in development of silicon carbide composites for fusion applications

    NASA Astrophysics Data System (ADS)

    Nozawa, T.; Hinoki, T.; Hasegawa, A.; Kohyama, A.; Katoh, Y.; Snead, L. L.; Henager, C. H., Jr.; Hegeman, J. B. J.

    2009-04-01

    Radiation-resistant advanced silicon carbide (SiC/SiC) composites have been developed as a promising candidate of the high-temperature operating advanced fusion reactor. With the completion of the 'proof-of-principle' phase in development of 'nuclear-grade' SiC/SiC composites, the R&D on SiC/SiC composites is shifting toward the more pragmatic phase, i.e., industrialization of component manufactures and data-basing. In this paper, recent advances and issues in (1) development of component fabrication technology including joining and functional coating, e.g., a tungsten overcoat as a plasma facing barrier, (2) recent updates in characterization of non-irradiated properties, e.g., strength anisotropy and chemical compatibility with solid lithium-based ceramics and lead-lithium liquid metal breeders, and (3) irradiation effects are specifically reviewed. Importantly high-temperature neutron irradiation effects on microstructural evolution, thermal and electrical conductivities and mechanical properties including the fiber/matrix interfacial strength are specified under various irradiation conditions, indicating seemingly very minor influence on the composite performance in the design temperature range.

  18. Recent advances and issues in development of silicon carbide composites for fusion applications

    SciTech Connect

    Nozawa, Takashi; Hinoki, Tatsuya; Hasegawa, Akira; Kohyama, Akira; Katoh, Yutai; Snead, Lance Lewis; HenagerJr., Charles H.; Hegeman, Hans

    2009-01-01

    Radiation-resistant advanced silicon carbide (SiC/SiC) composites have been developed as a promising candidate of the high-temperature operating advanced fusion reactor. With the completion of the 'proof-of-principle' phase in development of 'nuclear-grade' SiC/SiC composites, the R&D on SiC/SiC composites is shifting toward the more pragmatic phase, i.e., industrialization of component manufactures and data-basing. In this paper, recent advances and issues in (1) development of component fabrication technology including joining and functional coating, e.g., a tungsten overcoat as a plasma facing barrier, (2) recent updates in characterization of non-irradiated properties, e.g., strength anisotropy and chemical compatibility with solid lithium-based ceramics and lead-lithium liquid metal breeders, and (3) irradiation effects are specifically reviewed. Importantly high-temperature neutron irradiation effects on microstructural evolution, thermal and electrical conductivities and mechanical properties including the fiber/matrix interfacial strength are specified under various irradiation conditions, indicating seemingly very minor influence on the composite performance in the design temperature range.

  19. Recent advances and issues in development of silicon carbide composites for fusion applications

    SciTech Connect

    Nozawa, T.; Hinoki, Tetsuya; Hasegawa, Akira; Kohyama, Akira; Katoh, Yutai; Snead, Lance L.; Henager, Charles H.; Hegeman, Hans

    2009-04-30

    Radiation-resistant advanced silicon carbide composites (SiC/SiC) have been developed as a promising candidate of the high-temperature operating advanced fusion DEMO reactor. With the completion of the “proof-of-principle” phase in development of “nuclear-grade” SiC/SiC, the R&D on SiC/SiC is shifting toward the more pragmatic phase, i.e., industrialization of component manufactures and data-basing. In this paper, recent advances and issues in 1) development of component fabrication technology including joining and functional coating, e.g., a tungsten overcoat as a plasma facing barrier, 2) recent updates in characterization of non-irradiated properties, e.g., strength anisotropy and chemical compatibility with solid lithium-based ceramics and lead-lithium liquid metal breeders, and 3) irradiation effects were specifically reviewed. Importantly high-temperature neutron irradiation effects on microstructural evolution, thermal and electrical conductivities and mechanical properties including the fiber/matrix interfacial strength were specified under various irradiation conditions, indicating seemingly very minor influence on the composite performance in the design temperature range.

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

    SciTech Connect

    Tsai, H.; Kazakov, V.A.; Chakin, V.P.

    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.

  1. Testing of Gas Reactor Fuel and Materials in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2006-10-01

    The recent growth in interest for high temperature gas reactors has resulted in an increased need for materials and fuel testing for this type of reactor. The Advanced Test Reactor (ATR), located at the US Department of Energy’s Idaho National Laboratory, has long been involved in testing gas reactor fuel and materials, and has facilities and capabilities to provide the right environment for gas reactor irradiation experiments. These capabilities include both passive sealed capsule experiments, and instrumented/actively controlled experiments. The instrumented/actively controlled experiments typically contain thermocouples and control the irradiation temperature, but on-line measurements and controls for pressure and gas environment have also been performed in past irradiations. The ATR has an existing automated gas temperature control system that can maintain temperature in an irradiation experiment within very tight bounds, and has developed an on-line fission product monitoring system that is especially well suited for testing gas reactor particle fuel. The ATR’s control system, which consists primarily of vertical cylinders used to rotate neutron poisons/reflectors toward or away from the reactor core, provides a constant vertical flux profile over the duration of each operating cycle. This constant chopped cosine shaped axial flux profile, with a relatively flat peak at the vertical centre of the core, is more desirable for experiments than a constantly moving axial flux peak resulting from a control system of axially positioned control components which are vertically withdrawn from the core.

  2. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage

  3. Study of a water-cooled convective divertor prototype for the DEMO fusion reactor

    NASA Astrophysics Data System (ADS)

    Di Maio, P.; Oliveri, E.; Vella, G.

    2000-04-01

    The plasma facing components of a fusion power reactor have a large impact on the overall plant design, its performance and availability and on the cost of electricity. The present work concerns a study of feasibility for a water-cooled prototype of the convective divertor component of the DEMO fusion reactor. The study has been carried out in two steps. In the first one thermal-hydraulic and neutronic parametric analyses have been performed to find out the prototype optimized configuration. In the second step thermo-mechanical analyses have been carried out on the obtained configuration to investigate the potential and limits of the proposed prototype, with a particular reference to the maximum heat flux it can undergo without incoming both in critical heat flux and in mechanical stress limits. The results show that the proposed divertor prototype is able to safely withstand peak heat fluxes of 9 MW/m2.

  4. Status of R&D activities on materials for fusion power reactors

    NASA Astrophysics Data System (ADS)

    Baluc, N.; Abe, K.; Boutard, J. L.; Chernov, V. M.; Diegele, E.; Jitsukawa, S.; Kimura, A.; Klueh, R. L.; Kohyama, A.; Kurtz, R. J.; Lässer, R.; Matsui, H.; Möslang, A.; Muroga, T.; Odette, G. R.; Tran, M. Q.; van der Schaaf, B.; Wu, Y.; Yu, J.; Zinkle, S. J.

    2007-10-01

    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, the People's Republic of China, Russia and the USA. They relate to the development of new high temperature, radiation resistant materials, the development of coatings that will act as erosion, corrosion, permeation and/or electrical/MHD barriers, characterization of 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.

  5. Tritium management in a fusion reactor--safety, handling and economical issues--

    SciTech Connect

    Tanabe, Tetsuo

    2009-02-19

    In order to establish a D-T fusion reactor as an energy source, it is not enough to have a DT burning plasma, and economical conversion of fusion energy to electricity and/or heat, a large enough margin of tritium breeding and tritium safety must be simultaneously achieved. In particular, handling of huge amount of tritium needs significant efforts to ensure that the radiation dose of radiological workers and of the public is below the limits specified by the International Commission on Radiological Protection (ICRP). In this paper, after the introduction of tritium as a fuel of DT reactors and as a radioisotope of hydrogen, tritium safety issues in fuel cycle and blanket systems are summarized. In particular, in-vessel tritium inventory, the most important and uncertain tritium safety issue, is discussed in detail.

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

    NASA Astrophysics Data System (ADS)

    Bosia, G.; Helou, W.; Goniche, M.; Hillairet, 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.

  7. Study of a Water-Cooled Convective Divertor Prototype for the DEMO Fusion Reactor

    SciTech Connect

    P. Di Maio; E. Oliveri; G. Vella

    2000-12-31

    The plasma facing components of a fusion power reactor have a large impact on the overall plant design, its performance and availability and on the cost of electricity. The present work concerns a study of feasibility for a water-cooled prototype of the convective divertor component of the DEMO fusion reactor. The study has been carried out in two steps. In the first one thermal-hydraulic and neutronic parametric analyses have been performed to find out the prototype optimized configuration. In the second step thermo-mechanical analyses have been carried out on the obtained configuration to investigate the potential and limits of the proposed prototype, with a particular reference to the maximum heat flux it can undergo without incoming both in critical heat flux and in mechanical stress limits. the results show that the proposed divertor prototype is able to safely withstand peak heat fluxes of 9 MW/m{sup 2}.

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

    SciTech Connect

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

    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.

  9. Neutronic analysis of alternative structural materials for fusion reactor blankets

    NASA Astrophysics Data System (ADS)

    Santos, Raul dos

    1988-07-01

    The neutronic performance of the International Tokamak Reactor (INTOR) blanket was studied when several alternative structural materials were used instead of the INTOR reference structural material, type 316 stainless steel. The alternative structural materials included: ferritic-, vanadium-, titanium-, long range ordered-, manganese austenitic-, and nimonic-alloys. All were treated both with and without a first-wall coating of beryllium or graphite. The tritium breeding ratio, the nuclear heating, and the gas (hydrogen and helium) production rates in the structural materials were calculated for the possible combinations of structural material and first-wall coating. These parameters were compared with those obtained by using SS-316. The nimonic alloy was the only one with worse neutronic performance than the SS-316.

  10. DOE/NE robotics for advanced reactors

    SciTech Connect

    Not Available

    1991-01-01

    This document details activities during this reporting period. The Michigan group has developed, built, and tested a general purpose interface circuit for DC motors and encoders. This interface is based on an advanced microchip, the HCTL 1100 manufactured by Hewlett Packard. The HCTL 1100 can be programmed by a host computer in real-time, allowing sophisticated motion control for DC motors. At the University of Florida, work on modeling the details of the seismic isolators and the jack mechanism has been completed. A separate 3D solid view of the seismic isolator floor, with the full set of isolators shown in detail, has been constructed within IGRIP. ORNL led the robotics team at the ALMR review meeting. Discussions were held with General Electric (GE) engineers and contractors on the robotic needs for the ALMR program. The Tennessee group has completed geometric modeling of the Andros Mark VI mobile platform with two fixed tracks and for articulated tracks, the give degree-of-freedom manipulator and its end-effector, and two cameras. A graphical control of panel was developed which allow the user to operate the simulated robot. The University of Texas team visited ORNL to complete the implementation of computed-torque controller on the CESARm manipulator. This controller was previously developed and computer simulations were carried out specifically for the CESARm robot.

  11. Summary of TFTR (Tokamak Fusion Test Reactor) diagnostics, including JET (Joint European Torus) and JT-60

    SciTech Connect

    Hill, K.W.; Young, K.M.; Johnson, L.C.

    1990-05-01

    The diagnostic instrumentation on TFTR (Tokamak Fusion Test Reactor) and the specific properties of each diagnostic, i.e., number of channels, time resolution, wavelength range, etc., are summarized in tables, grouped according to the plasma parameter measured. For comparison, the equivalent diagnostic capabilities of JET (Joint European Torus) and the Japanese large tokamak, JT-60, as of late 1987 are also listed in the tables. Extensive references are given to publications on each instrument.

  12. A three-bar model for ratcheting of fusion reactor first wall

    SciTech Connect

    Wolters, J.; Majumdar, S.

    1994-12-01

    First wall structures of fusion reactors are subjected to cyclic bending stresses caused by inhomogeneous temperature distribution during plasma burn cycles and by electromagnetically induced impact loads during plasma disruptions. Such a combination of loading can potentially lead to ratcheting or incremental accumulation of plastic strain with cycles. An elastic-plastic three-bar model is developed to investigate the ratcheting behavior of the first wall.

  13. Advances in process intensification through multifunctional reactor engineering

    SciTech Connect

    O'Hern, T. J.

    2012-03-01

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes critical to process intensification and implementation in commercial applications. Physics of the heat and mass transfer and chemical kinetics and how these processes are ultimately scaled were investigated. Specifically, we progressed the knowledge and tools required to scale a multifunctional reactor for acid-catalyzed C4 paraffin/olefin alkylation to industrial dimensions. Understanding such process intensification strategies is crucial to improving the energy efficiency and profitability of multifunctional reactors, resulting in a projected energy savings of 100 trillion BTU/yr by 2020 and a substantial reduction in the accompanying emissions.

  14. The DOE advanced gas reactor fuel development and qualification program

    NASA Astrophysics Data System (ADS)

    Petti, David; Maki, John; Hunn, John; Pappano, Pete; Barnes, Charles; Saurwein, John; Nagley, Scott; Kendall, Jim; Hobbins, Richard

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular high-temperature gas-cooled reactors (HTGRs) enable an efficient and cost-effective integration of the reactor system with non-electricity-generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The U.S. Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. An overview of the program and recent progress is presented.

  15. An advanced carbon reactor subsystem for carbon dioxide reduction

    NASA Technical Reports Server (NTRS)

    Noyes, Gary P.; Cusick, Robert J.

    1986-01-01

    An evaluation is presented of the development status of an advanced carbon-reactor subsystem (ACRS) for the production of water and dense, solid carbon from CO2 and hydrogen, as required in physiochemical air revitalization systems for long-duration manned space missions. The ACRS consists of a Sabatier Methanation Reactor (SMR) that reduces CO2 with hydrogen to form methane and water, a gas-liquid separator to remove product water from the methane, and a Carbon Formation Reactor (CFR) to pyrolize methane to carbon and hydrogen; the carbon is recycled to the SMR, while the produce carbon is periodically removed from the CFR. A preprototype ACRS under development for the NASA Space Station is described.

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

    SciTech Connect

    Chang, C.S. |; Zweben, S.J.; Schivell, J.; Budny, R.; Scott, S.

    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}.

  17. Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

    SciTech Connect

    Moe, Wayne Leland

    2015-05-01

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

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

    SciTech Connect

    Natesan, K.; Rink, D.L.

    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.

  19. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect

    David Petti

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

  20. Current status of the advanced high temperature reactor

    SciTech Connect

    Holcomb, D. E.; Iias, D.; Quails, A. L.; Peretz, F. J.; Varma, V. K.; Bradley, E. C.; Cisneros, A. T.

    2012-07-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central station type [1500 MW(e)] Fluoride salt-cooled High-temperature Reactor (FHR) that is currently under development by Oak Ridge National Laboratory for the U. S. Dept. of Energy, Office of Nuclear Energy's Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR design option exploration is a multidisciplinary design effort that combines core neutronic and fuel configuration evaluation with structural, thermal, and hydraulic analysis to produce a reactor and vessel concept and place it within a power generation station. The AHTR design remains at the notional level of maturity, as key technologies require further development and a logically complete integrated design has not been finalized. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. (authors)

  1. Current Status of the Advanced High Temperature Reactor

    SciTech Connect

    Holcomb, David Eugene; Ilas, Dan; Qualls, A L; Peretz, Fred J; Varma, Venugopal Koikal; Bradley, Eric Craig; Cisneros, Anselmo T.

    2012-01-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central station type [1500 MW(e)] Fluoride salt-cooled High-temperature Reactor (FHR) that is currently under development by Oak Ridge National Laboratory for the U. S. Department of Energy, Office of Nuclear Energy's Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR design option exploration is a multidisciplinary design effort that combines core neutronic and fuel configuration evaluation with structural, thermal, and hydraulic analysis to produce a reactor and vessel concept and place it within a power generation station. The AHTR design remains at the notional level of maturity, as key technologies require further development and a logically complete integrated design has not been finalized. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated.

  2. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect

    Britton Jr, Charles L; Roberts, Michael; Bull, Nora D; Holcomb, David Eugene; Wood, Richard Thomas

    2012-10-01

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  3. Development of advanced strain diagnostic techniques for reactor environments.

    SciTech Connect

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  4. Temperature controlled material irradiation in the advanced test reactor

    SciTech Connect

    Furstenau, R.V.; Ingrahm, F.W.

    1995-12-31

    The Advanced Test Reactor (ATR) is located at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho, USA and is owned and regulated by the U.S. Department of Energy (US DOE). The ATR is operated for the US DOE by Lockheed Martin Idaho Technologies. In recent years, prime irradiation space in the ATR has been made available for use by customers having irradiation service needs in addition to the reactor`s principal user, the U.S. Naval Nuclear Propulsion Program. To enhance the reactor`s capabilities, the US DOE has initiated the development of an Irradiation Test Vehicle (ITV) capable of providing neutron spectral tailoring and temperature control for up to 28 experiments. The ATR-ITV will have the flexibility to simultaneously support a variety of experiments requiring fast, thermal or mixed spectrum neutron environments. Temperature control is accomplished by varying the thermal conductivity across a gas gap established between the experiment specimen capsule wall and the experiment `in-pile tube (IPT)` inside diameter. Thermal conductivity is adjusted by alternating the control gas mixture ratio of two gases with different thermal conductivities.

  5. Advanced Test Reactor - A National Scientific User Facility

    SciTech Connect

    Clifford J. Stanley

    2008-05-01

    The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and energy can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ~1.0 x1015 n/cm2-sec with a maximum fast flux of ~5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

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

    NASA Technical Reports Server (NTRS)

    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.

  7. Advanced Test Reactor National Scientific User Facility Partnerships

    SciTech Connect

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  8. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect

    Not Available

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  9. Code qualification of structural materials for AFCI advanced recycling reactors.

    SciTech Connect

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L.

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the

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

    SciTech Connect

    Not Available

    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.

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

    SciTech Connect

    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.

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

    SciTech Connect

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

    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.

  13. Advances in Process Intensification through Multifunctional Reactor Engineering

    SciTech Connect

    O'Hern, Timothy; Evans, Lindsay; Miller, Jim; Cooper, Marcia; Torczynski, John; Pena, Donovan; Gill, Walt; Groten, Will; Judzis, Arvids; Foley, Richard; Smith, Larry; Cross, Will; Vogt, T.

    2011-06-27

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in

  14. Advances in Process Intensification through Multifunctional Reactor Engineering

    SciTech Connect

    O'Hern, Timothy; Evans, Lindsay; Miller, Jim; Cooper, Marcia; Torczynski, John; Pena, Donovan; Gill, Walt

    2011-02-01

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in

  15. [Recent Advances of Biomechanical Studies on Cervical Fusion and Non-fusion Surgery].

    PubMed

    Liao, Zhenhua; Liu, Weiqiang

    2016-02-01

    This article reviews the progress of biomechanical studies on anterior cervical fusion and non-fusion surgery in recent years. The similarities and differences between animal and human cervical spines as well as the major three biomechanical test methods are introduced. Major progresses of biomechanical evaluation in anterior cervical fusion and non-fusion devices, hybrid surgery, coupled motion and biomechanical parameters, such as the instant center of rotation, are classified and summarized. Future development of loading method, multilevel hybrid surgery and coupling character are also discussed. PMID:27382760

  16. The Advanced Test Reactor National Scientific User Facility

    SciTech Connect

    Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

    2011-08-01

    In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those researchers with the best ideas access to the most advanced test capability, regardless of the proposer’s physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these university capabilities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user’s technical needs. The current NSUF partners are

  17. Advanced pressurized water reactor for improved resource utilization, part II - composite advanced PWR concept

    SciTech Connect

    Turner, S.E.; Gurley, M.K.; Kirby, K.D.; Mitchell, W III

    1981-09-15

    This report evaluates the enhanced resource utilization in an advanced pressurized water reactor (PWR) concept using a composite of selected improvements identified in a companion study. The selected improvements were in the areas of reduced loss of neutrons to control poisons, reduced loss of neutrons in leakage from the core, and improved blanket/reflector concepts. These improvements were incorporated into a single composite advanced PWR. A preliminary assessment of resource requirements and costs and impact on safety are presented.

  18. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    SciTech Connect

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  19. Prognostics Health Management for Advanced Small Modular Reactor Passive Components

    SciTech Connect

    Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-10-18

    In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

  20. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    SciTech Connect

    Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley; Steven Taylor

    2012-08-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  1. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    SciTech Connect

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; K. Condie

    2011-06-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility (NSUF) in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  2. Preliminary safety evaluation of the advanced burner test reactor.

    SciTech Connect

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  3. The search for advanced remote technology in fast reactor reprocessing

    SciTech Connect

    Burch, W.D.; Herndon, J.N.; Stradley, J.G. )

    1990-01-01

    Research and development in fast reactor reprocessing has been under way [approximately] 20 yr in several countries. During the past decade, France and the United Kingdom have developed active programs in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the Experimental Breeder Reactor II (EBR-II) facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. The Federal Republic of Germany (FRG) and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper focuses on the search for improved facility concepts and better maintenance systems in the CFRP, and, in turn, on how developments at ORNL have influenced the technology elsewhere.

  4. Development and Assessment of Advanced Reactor Core Protection System

    NASA Astrophysics Data System (ADS)

    in, Wang-Kee; Park, Young-Ho; Baeg, Seung-Yeob

    An advanced core protection system for a pressurized water reactor, Reactor Core Protection System(RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%.

  5. Design of the reactor vessel inspection robot for the advanced liquid metal reactor

    SciTech Connect

    Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

    1994-06-01

    A consortium of four universities and Oak Ridge National Laboratory designed a prototype wall-crawling robot to perform weld inspection in an advanced nuclear reactor. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a mock non-hostile environment and shown to perform as expected, as detailed in this report.

  6. High-power-density approaches to magnetic fusion energy: Problems and promise of compact Reversed-Field Pinch Reactors (CRFPR)

    NASA Astrophysics Data System (ADS)

    Hagenson, R. L.; Krakowski, R. A.; Dreicer, H.

    If the cost assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves optimistic, approaches that promise considerably increased system power density and reduced mass utilization are 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 of this approach is particularly attractive from the view point of low field resistive coils operating with Ohmic losses that are small relative to the fusion power. The RFP, 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 is used for a subsequent conceptual engineering design of the Compact RFP Reactor.

  7. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOEpatents

    Bers, Abraham

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

  8. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOEpatents

    Fisch, Nathaniel J.

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

  9. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  10. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on October 4, 2011, Room...

  11. 77 FR 59678 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-28

    ... October 17, 2011, (76 FR 64126-64127). Detailed meeting agendas and meeting transcripts are available on... COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor...

  12. 77 FR 76089 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-26

    ... October 18, 2012, (77 FR 64146-64147). Detailed meeting agendas and meeting transcripts are available on... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor...

  13. 78 FR 37595 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-21

    ... participation in ACRS meetings were published in the Federal Register on October 18, 2012, (77 FR 64146- 64147... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor...

  14. ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  15. Joining of Oxide Dispersion Strengthened Steels for Advanced Reactors

    NASA Astrophysics Data System (ADS)

    Baker, B. W.; Brewer, L. N.

    2014-12-01

    The design, manufacture, and experimental analysis of structural materials capable of operation in the high temperatures, corrosive environments, and radiation damage spectra of future reactor designs remain one of the key pacing items for advanced reactor designs. The most promising candidate structural materials are vanadium-based refractory alloys, silicon carbide composites and oxide dispersion strengthened steels. Of these, oxide dispersion strengthened steels are a likely near-term candidate to meet required demands. This paper reviews different variants of oxide dispersion strengthened steels and discusses their capability with regard to high-temperature strength, corrosion resistance, and radiation damage resistance. Additionally, joining of oxide dispersion strengthened steels, which has been cited as a limiting factor preventing their use, is addressed and reviewed. Specifically, friction stir welding of these steels is reviewed as a promising joining method for oxide dispersion strengthened steels.

  16. Advanced Space Nuclear Reactors from Fiction to Reality

    NASA Astrophysics Data System (ADS)

    Popa-Simil, L.

    The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

  17. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect

    Michael Swanson; Daniel Laudal

    2008-03-31

    The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the KBR transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 2800 hours of operation on 11 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air-blown and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 95% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher

  18. Advanced thermionic reactors for surface nuclear power applications

    NASA Astrophysics Data System (ADS)

    Parlos, Alexander G.; Kent, Karl; Peddicord, Kenneth L.; Khan, Ehsan U.

    1991-09-01

    A preliminary feasibility study on a new concept for a highly compact space reactor power system is presented, consisting of in-core thermionic fuel elements and in-core heat pipes for passive core cooling. The reference fuel considered in this study is uranium carbide. The calculations reported include a neutronic design analysis using a 2D neutron transport model, as well as a simplified 1D thermal analysis of the reactor core, using a preliminary thermal sizing of the in-core heat pipes. Initial results indicate that the proposed core design is thermally and neutronically feasible, with a maximum steady-state fuel temperature below 2000 K. Alternate advanced fuels, such as various oxides of Am-242, result in exceedingly high fuel centerline temperatures because of the associated low thermal conductivities.

  19. A Novel Approach to Material Development for Advanced Reactor Systems

    SciTech Connect

    Was, G.S.; Atzmon, M.; Wang, L.

    1999-12-22

    OAK B188 A Novel Approach to Material Development for Advanced Reactor Systems. Year one of this project had three major goals. First, to specify, order and install a new high current ion source for more rapid and stable proton irradiation. Second, to assess the use low temperature irradiation and chromium pre-enrichment in an effort to isolate a radiation damage microstructure in stainless steels without the effects of RIS. Third, to prepare for the irradiation of reactor pressure vessel steel and Zircaloy. In year 1 quarter 1, the project goal was to order the high current ion source and to procure and prepare samples of stainless steel for low temperature proton irradiation.

  20. A Novel Approach to Material Development for Advanced Reactor Systems

    SciTech Connect

    Was, G.S.; Atzmon, M.; Wang, L.

    2000-06-27

    OAK B188 A Novel Approach to Material Development for Advanced Reactor Systems. Year one of this project had three major goals. First, to specify, order and install a new high current ion source for more rapid and stable proton irradiation. Second, to assess the use of low temperature irradiation and chromium pre-enrichment in an effort to isolate a radiation damage microstructure in stainless steel without the effects of RIS. Third, to initiate irradiation of reactor pressure vessel steel and Zircaloy. In year 1 quarter 3, the project goal was to complete irradiation of model alloys of RPV steels for a range of doses and begin sample characterization. We also planned to prepare samples for microstructure isolation in stainless steels, and to identify sources of Zircaloy for irradiation and characterization.

  1. Evolutionary/advanced light water reactor data report

    SciTech Connect

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  2. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    SciTech Connect

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  3. Advanced Test Reactor Testing Experience: Past, Present and Future

    SciTech Connect

    Frances M. Marshall

    2005-04-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 48" long and 5.0" diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors -- US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, wherein the target material is placed in a capsule, or plate form, and the capsule is in direct contact with the primary coolant. The next level of complexity of an experiment is an instrumented lead experiment, which allows for active monitoring and control of experiment conditions during the irradiation. The highest level of complexity of experiment is the pressurized water loop experiment, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

  4. Metal fires and their implications for advanced reactors.

    SciTech Connect

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-10-01

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

  5. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  6. A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor

    SciTech Connect

    Brad J. Merrill; Richard L. Moore; J. Phillip Sharp

    2008-09-01

    A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). This beryllium dust oxidation model accounts for the diffusion of steam into a beryllium dust layer, the oxidation of the dust particles inside this layer based on the beryllium-steam oxidation equations developed at the INL, and the effective thermal conductivity of this beryllium dust layer. This paper details this oxidation model and presents the results of the application of this model to a wet bypass accident scenario in the ITER device.

  7. A magnetohydrodynamic stability study of reverse shear equilibria in the Tokamak Fusion Test Reactor

    SciTech Connect

    Phillips, M.W.; Zarnstorff, M.C.; Manickam, J.; Levinton, F.M.; Hughes, M.H.

    1996-05-01

    A study is presented of the low-{ital n} ({ital n}=1,2,3) magnetohydrodynamic stability of equilibria with reverse shear safety factor profiles. The low-{ital n} stability boundaries are found to be characterized by resonance structures due to internal so-called {open_quote}{open_quote}infernal{close_quote}{close_quote} mode types of instabilities. The parametric dependence of shear reversal width and depth, current, and pressure gradient on the beta limit are determined by using profile models that allow each parameter to be varied independently. Reverse magnetic shear is found to have a stabilizing influence for modes with toroidal mode numbers {ital n}{ge}2 leading to the possibility of improved {beta} limits in the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. {bold 26}, 11 (1984)]. {copyright} {ital 1996 American Institute of Physics.}

  8. Advanced Test Reactor -- Testing Capabilities and Plans AND Advanced Test Reactor National Scientific User Facility -- Partnerships and Networks

    SciTech Connect

    Frances M. Marshall

    2008-07-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plans for the NSUF.

  9. Advanced fuels for plutonium management in pressurized water reactors

    NASA Astrophysics Data System (ADS)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

  10. INEL advanced test reactor plutonium-238 production feasibility assessment

    SciTech Connect

    Schnitzler, B.G. )

    1993-01-10

    Results of a preliminary neutronics assessment indicate the feasibility of [sup 238]Pu production in the Idaho National Engineering Laboratory Advanced Test Reactor (ATR). Based on the results of this assessment, an annual production of 11.3 kg [sup 238]Pu can be achieved in the ATR. An annual loading of 102 kg [sup 237]Np is required for the particular target configuration and irradiation scenario examined. The [sup 236]Pu contaminant level is approximately 6 parts per million at zero cooling time. The product quality is about 90% [sup 238]Pu. Neptunium feedstock requirements, [sup 238]Pu production rates, or product purity can be optimized depending on their relative importances.

  11. Advanced High Temperature Reactor Systems and Economic Analysis

    SciTech Connect

    Holcomb, David Eugene; Peretz, Fred J; Qualls, A L

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience with

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

    NASA Astrophysics Data System (ADS)

    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).

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

    SciTech Connect

    Rimminen, H.; Kyynaeraeinen, J.

    2013-05-15

    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.

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

    SciTech Connect

    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)

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  17. Radiation streaming analysis for the beam ports of the laser fusion reactor, SENRI-I

    SciTech Connect

    Oomura, H.; Ido, S.; Nakai, S.; Nakashima, H.; Yamanaka, C.

    1984-01-01

    Three-dimensional Monte Carlo analyses are performed for the three-legged beam ports of the laser fusion reactor SENRI-I. The validity of two devices for attenuating neutron flux is examined, and it is shown that the fast neutron flux attenuation of 1.4 X 10/sup 5/ can be reached by adopting these two devices in combination. The effectiveness of variance reduction techniques is also examined. It is determined that path length stretching is useful for neutron groups above 1 MeV. Adopting adjoint flux-dependent Russian roulette weight cutoff also seems useful.

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

    SciTech Connect

    M.E. Lumia; C.A. Gentile

    2002-01-18

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

  19. Risk assessment of a fusion-reactor fuel-processing system

    SciTech Connect

    Bruske, S.Z.; Holland, D.F.

    1983-07-01

    The probabilistic risk assessment (PRA) methodology provides a means to systematically examine the potential for accidents that may result in a release of hazardous materials. This report presents the PRA for a typical fusion reactor fuel processing system. The system used in the analysis is based on the Tritium Systems Test Assembly, which is being operated at the Los Alamos National Laboratory. The results of the evaluation are presented in a probability-consequence plot that describes the probability of various accidental tritium release magnitudes.

  20. Recent results and challenges in development of metallic Hall sensors for fusion reactors

    SciTech Connect

    Ďuran, Ivan; Mušálek, Radek; Kovařík, Karel; Sentkerestiová, Jana; Kohout, Michal

    2014-08-21

    Reliable and precise diagnostic of local magnetic field is crucial for successful operation of future thermonuclear fusion reactors based on magnetic confinement. Magnetic sensors at these devices will experience an extremely demanding operational environment with large radiation and thermal loads in combination with required long term, reliable, and service-free performance. Neither present day commercial nor laboratory measurement systems comply with these requirements. Metallic Hall sensors based on e.g. copper or bismuth could potentially satisfy these needs. We present the technology for manufacturing of such sensors and some initial results on characterization of their properties.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  3. Local tests of parallel electrical resistivity in the Tokamak Fusion Test Reactor

    SciTech Connect

    Batha, S.H.; Levinton, F.M.; Ramsey, A.T.; Schmidt, G.L.; Zarnstorff, M.C.

    1997-01-01

    The motional Stark effect (MSE) polarimeter measures the local magnetic field pitch angle, proportional to the ratio of the poloidal to toroidal magnetic fields, in the Tokamak Fusion Test Reactor (TFTR). The authors have used the polarimeter to measure the temporal evolution of the local value of the magnetic field pitch angle during large changes in the current profile such as during a current ramp or discharge initiation. The measured evolution is compared to the evolution predicted by classical and neoclassical resistivity models. The neoclassical resistivity model is a better predictor of the local pitch angle temporal evolution than the classical model.

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

    PubMed

    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.

  5. Alpha particle losses from Tokamak Fusion Test Reactor deuterium{endash}tritium plasmas

    SciTech Connect

    Darrow, D.S.; Zweben, S.J.; Batha, S.; Budny, R.V.; Bush, C.E.; Chang, Z.; Cheng, C.Z.; Duong, H.H.; Fang, J.; Fisch, N.J.; Fischer, R.; Fredrickson, E.D.; Fu, G.Y.; Heeter, R.F.; Heidbrink, W.W.; Herrmann, H.W.; Herrmann, M.C.; Hill, K.; Jaeger, E.F.; James, R.; Majeski, R.; Medley, S.S.; Murakami, M.; Petrov, M.; Phillips, C.K.; Redi, M.H.; Ruskov, E.; Spong, D.A.; Strait, E.J.; Taylor, G.; White, R.B.; Wilson, J.R.; Wong, K.; Zarnstorff, M.C.

    1996-05-01

    Because alpha particle losses can have a significant influence on tokamak reactor viability, the loss of deuterium{endash}tritium alpha particles from the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire {ital et} {ital al}., Phys. Plasmas {bold 2}, 2176 (1995)] has been measured under a wide range of conditions. In TFTR, first orbit loss and stochastic toroidal field ripple diffusion are always present. Other losses can arise due to magnetohydrodynamic instabilities or due to waves in the ion cyclotron range of frequencies. No alpha particle losses have yet been seen due to collective instabilities driven by alphas. Ion Bernstein waves can drive large losses of fast ions from TFTR, and details of those losses support one element of the alpha energy channeling scenario. {copyright} {ital 1996 American Institute of Physics.}

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

    SciTech Connect

    Majumdar, S.; Smith, P.

    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.

  7. Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

    SciTech Connect

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity.

  8. Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

    SciTech Connect

    Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

    1981-04-01

    Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity.

  9. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  10. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  11. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    NASA Astrophysics Data System (ADS)

    Wali, W. A.; Hassan, K. H.; Cullen, J. D.; Al-Shamma'a, A. I.; Shaw, A.; Wylie, S. R.

    2011-08-01

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  12. The search for advanced remote technology in fast reactor reprocessing

    SciTech Connect

    Burch, W.D.; Herndon, J.N.; Stradley, J.G.

    1990-01-01

    Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. Germany and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in all of these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper will focus principally on the search for improved facility concepts and better maintenance systems in the CFRP and, in turn, on how developments at ORNL have influenced the technology elsewhere.

  13. Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.

    1999-07-23

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energymore » deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations.« less

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

    SciTech Connect

    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.

  15. Dense Z-pinch (DZP) as a fusion power reactor: preliminary scaling calculations and sysems energy balance

    SciTech Connect

    Hagenson, R.L.; Tai, A.S.; Krakowski, R.A.; Moses, R.W.

    1980-01-01

    A conceptual DT fusion reactor concept is described that is based upon the dense Z-pinch (DZP). This study emphasizes plasma modeling and the parametric assessment of the reactor energy balance. To this end simple analytic and numerical models have been developed and evaluated. The resulting optimal reactor operating point promises a high-Q, low-yield system of a scale that may allow the use of conventional high-voltage Marx/water-line technology to drive a potentially very small reactor system.

  16. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  17. Recurrent and pathological gene fusions in breast cancer: current advances in genomic discovery and clinical implications.

    PubMed

    Veeraraghavan, Jamunarani; Ma, Jiacheng; Hu, Yiheng; Wang, Xiao-Song

    2016-07-01

    Gene fusions have long been considered principally as the oncogenic events of hematologic malignancies, but have recently gained wide attention in solid tumors due to several milestone discoveries and the advancement of deep sequencing technologies. With the progress in deep sequencing studies of breast cancer transcriptomes and genomes, the discovery of recurrent and pathological gene fusions in breast cancer is on the focus. Recently, driven by new deep sequencing studies, several recurrent or pathological gene fusions have been identified in breast cancer, including ESR1-CCDC170, SEC16A-NOTCH1, SEC22B-NOTCH2, and ESR1-YAP1 etc. More important, most of these gene fusions are preferentially identified in the more aggressive breast cancers, such as luminal B, basal-like, or endocrine-resistant breast cancer, suggesting recurrent gene fusions as additional key driver events in these tumors other than the known drivers such as the estrogen receptor. In this paper, we have comprehensively summarized the newly identified recurrent or pathological gene fusion events in breast cancer, reviewed the contributions of new genomic and deep sequencing technologies to new fusion discovery and the integrative bioinformatics tools to analyze these data, highlighted the biological relevance and clinical implications of these fusion discoveries, and discussed future directions of gene fusion research in breast cancer. PMID:27372070

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

    NASA Astrophysics Data System (ADS)

    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.

  19. Fabrication and performance of AIN insulator coatings for application in fusion reactor blankets

    SciTech Connect

    Natesan, K.

    1995-09-01

    The liquid-metal blanket concept for fusion reactors requires an coating on the first-wall structural material to minimize the magnetohydrodynamic pressure drop that occurs during the flow of liquid metal in a magnetic field. Based on the thermodynamics of interactions betwen the coating and the liquid lithium on one side and the structural V-base alloy on the other side, an AIN coating was selected as a candidate. Detailed investigations were conducted on the fabrication, metallurgical microstructure, compatibility in liquid Li, and electrical characteristics of AIN material obtained from several sources. Lithium compatibility was studied in static systems by exposing AIN-coated specimens to liquid Li for several time periods. Electrical resistance was measured at room temperature on the specimens before and after exposure to liquid Li. The results obtained in this study indicate that AIN is a viable coating from the standpoint of chemical compatibility in Li, electrical insulation, and ease of fabrication; for these reasons, the coating should be examined further for fusion reactor applications.

  20. Performance of a palladium membrane reactor using a Ni catalyst for fusion fuel impurities processing

    SciTech Connect

    Willms, R.S.; Wilhelm, R.; Okuno, K.

    1994-07-01

    The palladium membrane reactor (PNM) provides a means to recover hydrogen isotopes from impurities expected to be present in fusion reactor exhaust. This recovery is based on reactions such as water-gas shift and steam reforming for which conversion is equilibrium limited. By including a selectively permeable membrane such as Pd/Ag in the catalyst bed, hydrogen isotopes can be removed from the reacting environment, thus promoting the reaction to complete conversion. Such a device has been built and operated at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). For the reactions listed above, earlier study with this unit has shown that hydrogen single-pass recoveries approaching 100% can be achieved. It was also determined that a nickel catalyst is a feasible choice for use with a PMR appropriate for fusion fuel impurities processing. The purpose of this study was to systematically assess the performance of the PMR using a nickel catalyst over a range of temperatures, feed compositions and flowrates. Reactions which were studied are the water-gas shift reaction and steam reforming.

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

    NASA Astrophysics Data System (ADS)

    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.

  2. Fusion Power Program biannual progress report, April-September 1979

    SciTech Connect

    Not Available

    1980-02-01

    This biannual report summarizes the Argonne National Laboratory work performed for the Office of Fusion Energy during the April-September 1979 quarter in the following research and development areas: materials; energy storage and transfer; tritium containment, recovery and control; advanced reactor design; atomic data; reactor safety; fusion-fission hybrid systems; alternate applications of fusion energy; and other work related to fusion power. Separate abstracts were prepared for three sections. (MOW)

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

    NASA Astrophysics Data System (ADS)

    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%.

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

    SciTech Connect

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

    1994-05-01

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

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

    SciTech Connect

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R.

    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.

  6. Multimodal options for materials research to advance the basis for fusion energy in the ITER era

    NASA Astrophysics Data System (ADS)

    Zinkle, S. J.; Möslang, A.; Muroga, T.; Tanigawa, H.

    2013-10-01

    Well-coordinated international fusion materials research on multiple fundamental feasibility issues can serve an important role during the next ten years. Due to differences in national timelines and fusion device concepts, a parallel-track (multimodal) approach is currently being used for developing fusion energy. An overview is given of the current state-of-the-art of major candidate materials systems for next-step fusion reactors, including a summary of existing knowledge regarding operating temperature and neutron irradiation fluence limits due to high-temperature strength and radiation damage considerations, coolant compatibility information, and current industrial manufacturing capabilities. There are two inter-related overarching objectives of fusion materials research to be performed in the next decade: (1) understanding materials science phenomena in the demanding DT fusion energy environment, and (2) application of this knowledge to develop and qualify materials to provide the basis for next-step facility construction authorization by funding agencies and public safety licensing authorities. The critical issues and prospects for development of high-performance fusion materials are discussed along with recent research results and planned activities of the international materials research community.

  7. The HiPER project for inertial confinement fusion and some experimental results on advanced ignition schemes

    NASA Astrophysics Data System (ADS)

    Batani, D.; Koenig, M.; Baton, S.; Perez, F.; Gizzi, L. A.; Koester, P.; Labate, L.; Honrubia, J.; Antonelli, L.; Morace, A.; Volpe, L.; Santos, J.; Schurtz, G.; Hulin, S.; Ribeyre, X.; Fourment, C.; Nicolai, P.; Vauzour, B.; Gremillet, L.; Nazarov, W.; Pasley, J.; Richetta, M.; Lancaster, K.; Spindloe, Ch; Tolley, M.; Neely, D.; Kozlová, M.; Nejdl, J.; Rus, B.; Wolowski, J.; Badziak, J.; Dorchies, F.

    2011-12-01

    This paper presents the goals and some of the results of experiments conducted within the Working Package 10 (Fusion Experimental Programme) of the HiPER Project. These experiments concern the study of the physics connected to 'advanced ignition schemes', i.e. the fast ignition and the shock ignition approaches to inertial fusion. Such schemes are aimed at achieving a higher gain, as compared with the classical approach which is used in NIF, as required for future reactors, and make fusion possible with smaller facilities. In particular, a series of experiments related to fast ignition were performed at the RAL (UK) and LULI (France) Laboratories and studied the propagation of fast electrons (created by a short-pulse ultra-high-intensity beam) in compressed matter, created either by cylindrical implosions or by compression of planar targets by (planar) laser-driven shock waves. A more recent experiment was performed at PALS and investigated the laser-plasma coupling in the 1016 W cm-2 intensity regime of interest for shock ignition.

  8. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Gregg L. Sharp; R. T. McCracken

    2003-06-01

    The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  9. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Sharp, G.L.; McCracken, R.T.

    2003-05-13

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  10. Overview of the US program of controls for advanced reactors

    SciTech Connect

    White, J.D.; Sackett, J.I.; Monson, R.; Lindsay, R.W.; Carroll, D.G.

    1989-01-01

    An automated control system can incorporate control goals and strategies, assessment of present and future plant status, diagnostic evaluation and maintenance planning, and signal and command validation. It has not been feasible to employ these capabilities in conventional hard-wired, analog, control systems. Recent advances in computer-based digital data acquisition systems, process controllers, fiber-optic signal transmission artificial intelligence tools and methods, and small inexpensive, fast, large-capacity computers---with both numeric and symbolic capabilities---have provided many of the necessary ingredients for developing large, practical automated control systems. Furthermore, recent reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. This paper presents an overall US national perspective for advanced controls research and development. The goals of high reliability, low operating cost and simple operation are described. The staged approach from conceptualization through implementation is discussed. Then the paper describes the work being done by ORNL, ANL and GE. The relationship of this work to the US commercial industry is also discussed.

  11. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    SciTech Connect

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  12. Supervisory Control System Architecture for Advanced Small Modular Reactors

    SciTech Connect

    Cetiner, Sacit M; Cole, Daniel L; Fugate, David L; Kisner, Roger A; Melin, Alexander M; Muhlheim, Michael David; Rao, Nageswara S; Wood, Richard Thomas

    2013-08-01

    This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

  13. The Advanced Test Reactor Irradiation Facilities and Capabilities

    SciTech Connect

    S. Blaine Grover; Raymond V. Furstenau

    2007-03-01

    The Advanced Test Reactor (ATR) is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR’s unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments.

  14. Advanced Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Romano, A.J.

    1980-01-01

    The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR safety evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  15. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  16. Advances in Hyperspectral and Multispectral Image Fusion and Spectral Unmixing

    NASA Astrophysics Data System (ADS)

    Lanaras, C.; Baltsavias, E.; Schindler, K.

    2015-08-01

    In this work, we jointly process high spectral and high geometric resolution images and exploit their synergies to (a) generate a fused image of high spectral and geometric resolution; and (b) improve (linear) spectral unmixing of hyperspectral endmembers at subpixel level w.r.t. the pixel size of the hyperspectral image. We assume that the two images are radiometrically corrected and geometrically co-registered. The scientific contributions of this work are (a) a simultaneous approach to image fusion and hyperspectral unmixing, (b) enforcing several physically plausible constraints during unmixing that are all well-known, but typically not used in combination, and (c) the use of efficient, state-of-the-art mathematical optimization tools to implement the processing. The results of our joint fusion and unmixing has the potential to enable more accurate and detailed semantic interpretation of objects and their properties in hyperspectral and multispectral images, with applications in environmental mapping, monitoring and change detection. In our experiments, the proposed method always improves the fusion compared to competing methods, reducing RMSE between 4% and 53%.

  17. Advanced neutron source reactor probabilistic flow blockage assessment

    SciTech Connect

    Ramsey, C.T.

    1995-08-01

    The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool.

  18. Fuel qualification plan for the Advanced Neutron Source Reactor

    SciTech Connect

    Copeland, G.L.

    1995-07-01

    This report describes the development and qualification plan for the fuel for the Advanced Neutron Source. The reference fuel is U{sub 3}Si{sub 2}, dispersed in aluminum and clad in 6061 aluminum. This report was prepared in May 1994, at which time the reference design was for a two-element core containing highly enriched uranium (93% {sup 235}U) . The reactor was in the process of being redesigned to accommodate lowered uranium enrichment and became a three-element core containing a higher volume fraction of uranium enriched to 50% {sup 235}U. Consequently, this report was not issued at that time and would have been revised to reflect the possibly different requirements of the lower-enrichment, higher-volume fraction fuel. Because the reactor is now being canceled, this unrevised report is being issued for archival purposes. The report describes the fabrication and inspection development plan, the irradiation tests and performance modeling to qualify performance, the transient testing that is part of the safety program, and the interactions and interfaces of the fuel development with other tasks.

  19. The neutron texture diffractometer at the China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  20. Advanced nuclear reactor public opinion project. Interim report

    SciTech Connect

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  1. Advanced Test Reactor National Scientific User Facility Progress

    SciTech Connect

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  2. Fusion-power demonstration

    NASA Astrophysics Data System (ADS)

    Henning, C. D.; Logan, B. G.; Carlson, G. A.; Neef, W. S.; Moir, R. W.; Campbell, R. B.; Botwin, R.; Clarkson, I. R.; Carpenter, T. J.

    1983-03-01

    As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment.

  3. 76 FR 5218 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ... Register on October 21, 2010 (75 FR 65038- 65039). Detailed meeting agendas and meeting transcripts are... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water...

  4. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES... developed using this Catalog along with the Operator Licensing Examination Standards for Power...

  5. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    SciTech Connect

    Not Available

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  6. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    SciTech Connect

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-06-19

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  7. Torus configuration and materials selection on a fusion DEMO reactor, SlimCS

    NASA Astrophysics Data System (ADS)

    Tobita, K.; Nishio, S.; Tanigawa, H.; Enoeda, M.; Isono, T.; Nakamura, H.; Tsuru, D.; Suzuki, S.; Hayashi, T.; Tsuchiya, K.; Hayashi, T.; Nishitani, T.; DEMO design Team

    2009-04-01

    SlimCS is the conceptual design of a compact fusion DEMO plant assuming technologies foreseeable in 2020-2030s. For continuity of blanket technology from the Japanese ITER-TBM, the prime option of blanket is water-cooled solid breeder with Li 2TiO 3 (or Li 4SiO 4) and Be. A reduced-activation ferritic-martensitic steel (RAFM) and subcritical water are chosen as the structural material and coolant, respectively. The reactor has somewhat complex torus configuration with a sector-wide conducting plate slipped in between the replaceable (front) and permanent (back) blanket. In order to allow flexibility of maintenance in such a complex configuration, sector transport hot cell maintenance scheme is adopted. This paper describes characteristics of SlimCS with a focus on materials selection.

  8. Calculation procedures for the analysis of integral experiments for fusion-reactor design

    NASA Astrophysics Data System (ADS)

    Santoro, R. T.; Barnes, J. M.; Alsmiller, R. G., Jr.; Oblow, E. M.

    1981-07-01

    The calculational models, nuclear data, and radiation transport codes that are used in the analysis of integral measurements of the transport of approxmately 14 MeV neutrons through laminated slabs of materials typical of those found in fusion reactor shields are described. The two dimensional discrete ordinates calculations to optimize the experimental configuration of reducing the neutron and gamma ray background levels and for obtaining and equivalent, reduced geometry of the calculational model to reduce computer core storage and running times are also presented. The equations and data to determine the energy-angle relations to neutrons produced in the reactions of 250 keV deuterons in a titanium tritide target are given. The procedures used to collapse the 17ln 36gamma VATAMIN C cross section library to a 53n 21 gamma broad group library are described.

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

    SciTech Connect

    Thomson, S.; Pilatzke, K.; McCrimmon, K.; Castillo, I.; Suppiah, S.

    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.

  10. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    SciTech Connect

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-08-01

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method.

  11. Removing tritium and other impurities during industrial recycling of beryllium from a fusion reactor

    SciTech Connect

    Dylst, K.; Seghers, J.; Druyts, F.; Braet, J.

    2008-07-15

    Recycling beryllium used in a fusion reactor might be a good way to overcome problems related to the disposal of neutron irradiated beryllium. The critical issues for the recycling of used first wall beryllium are the presence of tritium and (transuranic) impurities. High temperature annealing seems to be the most promising technique for detritiation. Purification of the de-tritiated beryllium can be achieved by chlorination of the irradiated beryllium and the subsequent reduction of beryllium chloride to highly pure metallic beryllium. After that, the beryllium can be re-fabricated into first wall tiles via powder metallurgy which is already a mature industrial practice. This paper outlines the path to define the experimental needs for beryllium recycling and tackles problems related to the detritiation and the purification via the chlorine route. (authors)

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

    SciTech Connect

    Ingesson, L. C.; 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.

  13. Investigation of Liquid Metal Embrittlement of Materials for use in Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Kennedy, Daniel; Jaworski, Michael

    2014-10-01

    Liquid metals can provide a continually replenished material for the first wall and extraction blankets of fusion reactors. However, research has shown that solid metal surfaces will experience embrittlement when exposed to liquid metals under stress. Therefore, it is important to understand the changes in structural strength of the solid metal materials and test different surface treatments that can limit embrittlement. Research was conducted to design and build an apparatus for exposing solid metal samples to liquid metal under high stress and temperature. The apparatus design, results of tensile testing, and surface imaging of fractured samples will be presented. This work was supported in part by the U.S. Department of Energy, Office of Science, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

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

    SciTech Connect

    Odette, G.R.

    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.

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

    SciTech Connect

    Purdy, I.M.

    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.

  16. Kβ spectra of heliumlike iron from tokomak-fusion-test-reactor plasmas

    NASA Astrophysics Data System (ADS)

    Smith, A. J.; Bitter, M.; Hsuan, H.; Hill, K. W.; von Goeler, S.; Timberlake, J.; Beiersdorfer, P.; Osterheld, A.

    1993-04-01

    Kβ spectra of heliumlike iron, Fe xxv, have been observed from tokamak-fusion-test-reactor (TFTR) plasmas with a high-resolution crystal spectrometer. The wavelength range of the Fe Kβ spectrum partially overlaps the spectrum of heliumlike nickel (Ni xxvii), which is used on TFTR and the joint European torus for ion-temperature measurements. The experimental arrangement made it possible to observe the Fe xxv Kβ lines and their satellite transitions of the type 1s2l'3l''-->1s22l', as well as the entire satellite spectrum of the Ni xxvii Kα line simultaneously. In order to identify the features of the Kβ spectra and to study their possible interference with the Ni xxvii spectrum, the intensity of the Kβ spectrum was enhanced by the injection of iron into the plasma. Accurate wavelengths and intensities have been obtained and compared with different theoretical calculations.

  17. Long- and short-term trends in vessel conditioning of TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    LaMarche, P.H.; Dylla, H.F.; Bell, M.G.; Boody, F.P.; Bush, C.E.; Groebuer, R.J.; Hawryluk, R.J.; Hill, K.W.; Mueller, D.; Owens, D.K.

    1986-10-01

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

  18. Divertor conditions relevant for fusion reactors achieved with linear plasma generator

    SciTech Connect

    Eck, H. J. N. van; Lof, A.; Meiden, H. J. van der; Rooij, G. J. van; Scholten, J.; Zeijlmans van Emmichoven, P. A.; Kleyn, A. W.

    2012-11-26

    Intense magnetized hydrogen and deuterium plasmas have been produced with electron densities up to 3.6 Multiplication-Sign 10{sup 20} m{sup -3} and electron temperatures up to 3.7 eV with a linear plasma generator. Exposure of a W target has led to average heat and particle flux densities well in excess of 4 MW m{sup -2} and 10{sup 24} m{sup -2} s{sup -1}, respectively. We have shown that the plasma surface interactions are dominated by the incoming ions. The achieved conditions correspond very well to the projected conditions at the divertor strike zones of fusion reactors such as ITER. In addition, the machine has an unprecedented high gas efficiency.

  19. Modelling of advanced structural materials for GEN IV reactors

    NASA Astrophysics Data System (ADS)

    Samaras, M.; Hoffelner, W.; Victoria, M.

    2007-09-01

    The choice of suitable materials and the assessment of long-term materials damage are key issues that need to be addressed for the safe and reliable performance of nuclear power plants. Operating conditions such as high temperatures, irradiation and a corrosive environment degrade materials properties, posing the risk of very expensive or even catastrophic plant damage. Materials scientists are faced with the scientific challenge to determine the long-term damage evolution of materials under service exposure in advanced plants. A higher confidence in life-time assessments of these materials requires an understanding of the related physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To overcome lengthy and expensive trial-and-error experiments, the multiscale modelling of materials behaviour is a promising tool, bringing new insights into the fundamental understanding of basic mechanisms. This paper presents the multiscale modelling methodology which is taking root internationally to address the issues of advanced structural materials for Gen IV reactors.

  20. Advanced Deuterium Fusion Rocket Propulsion for Manned Deep Space Missions

    NASA Astrophysics Data System (ADS)

    Winterberg, F.

    Excluding speculations about future breakthrough discoveries in physics, it is shown that with what is at present known, and also what is technically feasible, manned space flight to the limits of the solar system and beyond deep into the Oort cloud is quite possible. Using deuterium as the rocket fuel of choice, abundantly available on the comets of the Oort cloud, rockets driven by deuterium fusion can there be refuelled. To obtain a high thrust with high specific impulse favours the propulsion by deuterium micro-bombs, and it is shown that the ignition of deuterium micro-bombs is possible by intense GeV proton beams, generated in space by using the entire spacecraft as a magnetically insulated billion volt capacitor. The cost to develop this kind of a propulsion system in space would be very high, but it can also be developed on Earth by a magnetically insulated Super Marx Generator. Since the ignition of deuterium is theoretically possible with the Super Marx Generator, making obsolete the ignition of deuterium-tritium with a laser, where 80% of the energy goes into neutrons, this would also mean a breakthrough in fusion research, and therefore would justify the large development costs.

  1. (Fusion materials R D programs of the International Thermonuclear Experimental Reactor)

    SciTech Connect

    Reuther, T.C.

    1990-10-12

    The objective of this travel was to advance provisional planning of an activity to coordinate research and development (R D) activities on fusion materials among the existing fusion materials R D programs of the ITER parties. This objective was accomplished in discussions with the Executive Committee for the IEA Implementing Agreement on Fusion Materials in Karlsruhe, Germany, and with the ITER management and staff of Garching, Germany. The IEA Executive Committee deferred substantive consideration of the topic at the insistence of the Ex-Officio member from European Community (EC), Brussels. Discussions with ITER management and staff were positive. It was noted the the draft ITER Long-Term Technology Research and Development Plan contains recommendations for major program effort in materials R D and includes recommendations for coordination among the existing programs of the parties to meet those materials R D needs. ITER management discussions were in the context that decisions on specific activities for the ITER engineering design activity (EDA) must await formal action by the parties on the ITER EDA.

  2. High poloidal beta long-pulse experiments in the Tokamak Fusion Test Reactor*

    NASA Astrophysics Data System (ADS)

    Kesner, J.; Mauel, M. E.; Navratil, G. A.; Sabbagh, S. A.; Bell, M.; Budny, R.; Bush, C.; Fredrickson, E.; Grek, B.; Janos, A.; Johnson, D.; Mansfield, D.; McCune, D.; McGuire, K.; Park, H.; Ramsey, A.; Synakowski, E.; Taylor, G.; Zarnstorff, M.; Batha, S. H.; Levinton, F. M.

    1993-07-01

    Experiments have been performed in the Tokamak Fusion Test Reactor [D. M. Meade et al. in Plasma Physics Controlled Nuclear Fusion Research, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, p. 9] with neutral beam injection of up to 4 sec. duration, which is comparable to the time scale for resistive redistribution of the plasma current profile. These plasmas were created using a rapid decrease of the plasma current which initially created a plasma with enhanced stability and confinement. As the current profile evolved, a significantly reduced beta limit was observed. The high ɛβp plasmas had up to 90% of the current driven noninductively which significantly broadened the current profile during the long pulse lengths. These experiments demonstrated that high βN plasmas could not be sustained for times longer than the resistive relaxation of the outer current region which at early times after the current ramp-down carried negative current. At later times in lower βN discharges, beta collapses were sometimes observed as the current profile broadened at βN˜1.5. The appearance of disruptions was consistent with the predictions of ideal magnetohydrodynamics (MHD) stability analyses.

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

    SciTech Connect

    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.

  4. Multilayer mirror based monitors for impurity controls in large fusion reactor type devices

    SciTech Connect

    Regan, S.P.; May, M.J.; Soukhanovskii, V.; Finkenthal, M.; Moos, H.W.

    1995-12-31

    Multilayer Mirror (MLM) based monitors are compact, high throughput diagnostics capable of extracting XUV emissions (the wavelength range including the soft-x-ray and the extreme ultraviolet, 10 {angstrom} to 304 {angstrom}) of impurities from the harsh environment of large fusion reactor type devices. For several years the Plasma Spectroscopy Group at Johns Hopkins University has investigated the application of MLM based XUV spectroscopic diagnostics for magnetically confined fusion plasmas. MLM based monitors have been constructed for and extensively used on DIII-D, Alcator C-mod, TEXT, Phaedrus-T, and CDX-U tokamaks to study the impurity behavior of elements ranging from He to Mo. On ITER MLM based devices would be used to monitor the spectral line emissions from Li I-like to F I-like charge states of Fe, Cr, and Ni, as well as extractors for the bands of emissions from high Z elements such as Mo or W for impurity controls of the fusion plasma. In addition to monitoring the impurity emissions from the main plasma, MLM based devices can also be adapted for radiation measurements of low Z elements in the divertor. The concepts and designs of these MLM based monitors for impurity controls in ITER will be presented. The results of neutron irradiation experiments of the MLMs performed in the Los Alamos Spallation Radiation Effects Facility (LASREF) at the Los Alamos National Laboratory will also be discussed. These preliminary neutron exposure studies show that the dispersive and reflective qualities of the MLMs were not affected in a significant manner.

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

    SciTech Connect

    Bell, M.G.; Beer, M.; Batha, S.

    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.

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

    SciTech Connect

    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.

    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.}

  7. Magnetohydrodynamic modes analysis and control of Fusion Advanced Studies Torus high-current scenarios

    SciTech Connect

    Villone, F.; Mastrostefano, S.; Calabrò, G.; Vlad, G.; Crisanti, F.; Fusco, V.; Marchiori, G.; Bolzonella, T.; Marrelli, L.; Martin, P.; Liu, Y. Q.

    2014-08-15

    One of the main FAST (Fusion Advanced Studies Torus) goals is to have a flexible experiment capable to test tools and scenarios for safe and reliable tokamak operation, in order to support ITER and help the final DEMO design. In particular, in this paper, we focus on operation close to a possible border of stability related to low-q operation. To this purpose, a new FAST scenario has then been designed at I{sub p} = 10 MA, B{sub T} = 8.5 T, q{sub 95} ≈ 2.3. Transport simulations, carried out by using the code JETTO and the first principle transport model GLF23, indicate that, under these conditions, FAST could achieve an equivalent Q ≈ 3.5. FAST will be equipped with a set of internal active coils for feedback control, which will produce magnetic perturbation with toroidal number n = 1 or n = 2. Magnetohydrodynamic (MHD) mode analysis and feedback control simulations performed with the codes MARS, MARS-F, CarMa (both assuming the presence of a perfect conductive wall and using the exact 3D resistive wall structure) show the possibility of the FAST conductive structures to stabilize n = 1 ideal modes. This leaves therefore room for active mitigation of the resistive mode (down to a characteristic time of 1 ms) for safety purposes, i.e., to avoid dangerous MHD-driven plasma disruption, when working close to the machine limits and magnetic and kinetic energy density not far from reactor values.

  8. Conceptual design of a laser-fusion power plant. Part II. Two technical options: 1. JADE reactor; 2. Heat transfer by heat pipes

    SciTech Connect

    Not Available

    1981-07-01

    A laser fusion reactor concept is described that employs liquid metal walls. The concept envisions a porous medium, called the JADE, of specific geometry lining the reactor cavity. Some advantages and disadvantages of the concept are pointed out. The possibility of using heat pipes for passive cooling in ICF reactors is discussed. Some of the problems are outlined. (MOW)

  9. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels. PMID:23274823

  10. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

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

    SciTech Connect

    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).

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

    NASA Astrophysics Data System (ADS)

    Dekeyser, W.; Reiter, D.; Baelmans, M.

    2012-10-01

    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.

  13. A target station for plasma exposure of neutron irradiated fusion material samples to reactor relevant conditions

    NASA Astrophysics Data System (ADS)

    Rapp, Juergen; Giuliano, Dominic; Ellis, Ronald; Howard, Richard; Lore, Jeremy; Lumsdaine, Arnold; Lessard, Timothy; McGinnis, William; Meitner, Steven; Owen, Larry; Varma, Venugopal

    2015-11-01

    The Material Plasma Exposure eXperiment (MPEX) is a device planned to address scientific and technological gaps for the development of viable plasma facing components for fusion reactor conditions (FNSF, DEMO). It will have to address the relevant plasma conditions in a reactor divertor (electron density, electron temperature, ion fluxes) and it needs to be able to expose a-priori neutron irradiated samples. A pre design of a target station able to handle activated materials will be presented. This includes detailed MCNP as well as SCALE and MAVRIC calculations for all potential plasma-facing materials to estimate dose rates. Details on the remote handling schemes for the material samples will be presented. 2 point modeling of the linear plasma transport has been used to scope out the parameter range of the anticipated power fluxes to the target. This has been used to design the cooling capability of the target. The operational conditions of surface temperatures, plasma conditions, and oblique angle of incidence of magnetic field to target surface will be discussed. ORNL is managed by UT-Battelle, LLC, for the U.S. DOE under contract DE-AC-05-00OR22725.

  14. Fatigue behavior of irradiated helium-containing ferritic steels for fusion reactor applications*1

    NASA Astrophysics Data System (ADS)

    Grossbeck, M. L.; Vitek, J. M.; Liu, K. C.

    1986-11-01

    The martensitic alloys 12Cr-1MoVW and 9Cr-1MoVNb have been irradiated in the High Flux Isotope Reactor (HFIR) and subsequently tested in fatigue. In order to achieve helium levels characteristic of fusion reactors, the 12Cr-1MoVW was doped with 1 and 2% Ni, resulting in helium levels of 210 and 410 appm at damage levels of 25 dpa. The 9Cr-1MoVNb was irradiated to a damage level of 3 dpa and contained less than 5 appm He. Irradiations were carried out at 55°C and testing at 22°C. No significant changes were found in 9Cr-1MoVNb upon irradiation at this damage level, but effects that could possibly be attributed to helium were found in 12Cr-1MoVW. Levels of 210 and 410 appm He produced cyclic strengthening of 29 and 34% over unirradiated nickel-doped materials, respectively. This cyclic hardening, attributable largely to helium, resulted in degradation of the cyclic life. However, the fatigue life remained comparable to or better than unirradiated 20%-cold-worked 316 stainless steel.

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

    NASA Astrophysics Data System (ADS)

    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.

  16. Technical evaluation of major candidate blanket systems for fusion power reactor

    NASA Astrophysics Data System (ADS)

    Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

    1987-03-01

    The key functions required for tritium breeding blankets for a fusion power reactor are ; (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li/sub 2/O/H/sub 2/O/Be, Mo-alloy/Li/sub 2/O/He/Be, Mo-alloy/LiAlO/sub 2//He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies.

  17. Advanced SiC composites for fusion applications

    SciTech Connect

    Snead, L.L.; Schwarz, O.J.

    1995-04-01

    This is a short review of the motivation for and progress in the development of ceramic matrix composites for fusion. Chemically vapor infiltrated silicon carbide (SiC) composites have been fabricated from continuous fibers of either SiC or graphite and tested for strength and thermal conductivity. Of significance is the the Hi-Nicalon{trademark} SiC based fiber composite has superior unirradiated properties as compared to the standard Nicalon grade. Based on previous results on the stability of the Hi-Nicalon fiber, this system should prove more resistant to neutron irradiation. A graphite fiber composite has been fabricated with very good mechnical properties and thermal conductivity an order of magnitude higher than typical SiC/SiC composites.

  18. Anomalous Loss of DT Alpha Particles in the Tokamak Fusion Test Reactor

    SciTech Connect

    Herrmann, Hans W.

    1997-06-01

    Princeton's Tokamak Fusion Test Reactor (TFTR) is the first experimental fusion device to routinely use tritium to study the deuterium-tritium (DT) fusion reaction,allowing the first systematic study of DT alpha particles in tokamak plasmas. A crucial aspect of alpha-particle physics is the fraction of alphas that escape from the plasma, particularly since these energetic particles can do severe damage to the first wall of a reactor. An escaping alpha collector probe has been developed for TFTR's DT phase. Energy distributions of escaping alphas have been determined by measuring the range of alpha-particles implanted into nickel foils located within the alpha collector. Results at 1.0 MA of plasma current are in good agreement with predictions for first orbit alpha loss. Results at 1.8 MA, however, show a significant anomalous loss of partially thermalized alphas (in addition to the expected first orbit loss), which is not observed with the lost alpha scintillator detectors in DT plasmas, but does resemble the anomalous "delayed" loss seen in DD plasmas. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations. An experiment designed to study the effect of plasma major radius shifts on alpha-particle loss has led to a better understanding of alpha-particle dynamics in tokamaks. Intuitively, one might suppose that confined marginally passing alpha-particles forced to move toward higher magnetic field during an inward major radius shift (i.e. compression) would mirror and become trapped particles, leading to increased alpha loss. Such an effect was looked for during the shift experiment, however, no significant changes in alpha loss to the 90 degree lost alpha scintillator detector were observed during the shifts. It is calculated that the energy gained by an alpha-particle during the inward shift is sufficient to explain this result. However, an unexpected loss of partially thermalized alpha-particles near the

  19. A NATIONAL COLLABORATORY TO ADVANCE THE SCIENCE OF HIGH TEMPERATURE PLASMA PHYSICS FOR MAGNETIC FUSION

    SciTech Connect

    Allen R. Sanderson; Christopher R. Johnson

    2006-08-01

    This report summarizes the work of the University of Utah, which was a member of the National Fusion Collaboratory (NFC) Project funded by the United States Department of Energy (DOE) under the Scientific Discovery through Advanced Computing Program (SciDAC) to develop a persistent infrastructure to enable scientific collaboration for magnetic fusion research. A five year project that was initiated in 2001, it the NFC built on the past collaborative work performed within the U.S. fusion community and added the component of computer science research done with the USDOE Office of Science, Office of Advanced Scientific Computer Research. The project was itself a collaboration, itself uniting fusion scientists from General Atomics, MIT, and PPPL and computer scientists from ANL, LBNL, and Princeton University, and the University of Utah to form a coordinated team. The group leveraged existing computer science technology where possible and extended or created new capabilities where required. The complete finial report is attached as an addendum. The In the collaboration, the primary technical responsibility of the University of Utah in the collaboration was to develop and deploy an advanced scientific visualization service. To achieve this goal, the SCIRun Problem Solving Environment (PSE) is used on FusionGrid for an advanced scientific visualization service. SCIRun is open source software that gives the user the ability to create complex 3D visualizations and 2D graphics. This capability allows for the exploration of complex simulation results and the comparison of simulation and experimental data. SCIRun on FusionGrid gives the scientist a no-license-cost visualization capability that rivals present day commercial visualization packages. To accelerate the usage of SCIRun within the fusion community, a stand-alone application built on top of SCIRun was developed and deployed. This application, FusionViewer, allows users who are unfamiliar with SCIRun to quickly create

  20. Core design studies for advanced burner test reactor.

    SciTech Connect

    Yang, W. S.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

    2008-01-01

    The U.S. government announced in February 2006 the Global Nuclear Energy Partnership (GNEP) to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. The advanced burner reactor (ABR) based on a fast spectrum is one of the three major technologies to be demonstrated in GNEP. In FY06, a pre-conceptual design study was performed to develop an advanced burner test reactor (ABTR) that supports development of a prototype full-scale ABR, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR were (1) to demonstrate reactor-based transmutation of transuranics (TRU) as part of an advanced fuel cycle, (2) to qualify the TRU-containing fuels and advanced structural materials needed for a full-scale ABR, (3) to support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. Based on these objectives, core design and fuel cycle studies were performed to develop ABTR core designs, which can accommodate the expected changes of the TRU feed and the conversion ratio. Various option and trade-off studies were performed to determine the appropriate power level and conversion ratio. Both ternary metal alloy (U-TRU-10Zr) and mixed oxide (UO{sub 2}-TRUO{sub 2}) fuel forms have been considered with TRU feeds from weapons-grade plutonium (WG-Pu) and TRU recovered from light water reactor spent fuel (LWR-SF). Reactor performances were evaluated in detail including equilibrium cycle core parameters, mass flow, power distribution, kinetic parameters, reactivity feedback coefficient, reactivity control requirements and shutdown margins, and spent fuel characteristics. Trade-off studies on power level suggested that about 250 MWt is a reasonable compromise to allow a low project cost, at the same time providing a reasonable prototypic irradiation environment for demonstrating

  1. Integrated continuous dissolution, refolding and tag removal of fusion proteins from inclusion bodies in a tubular reactor.

    PubMed

    Pan, Siqi; Zelger, Monika; Jungbauer, Alois; Hahn, Rainer

    2014-09-20

    An integrated continuous tubular reactor system was developed for processing an autoprotease expressed as inclusion bodies. The inclusion bodies were suspended and fed into the tubular reactor system for continuous dissolving, refolding and precipitation. During refolding, the dissolved autoprotease cleaves itself, separating the fusion tag from the target peptide. Subsequently, the cleaved fusion tag and any uncleaved autoprotease were precipitated out in the precipitation step. The processed exiting solution results in the purified soluble target peptide. Refolding and precipitation yields performed in the tubular reactor were similar to batch reactor and process was stable for at least 20 h. The authenticity of purified peptide was also verified by mass spectroscopy. Productivity (in mg/l/h and mg/h) calculated in the tubular process was twice and 1.5 times of the batch process, respectively. Although it is more complex to setup a tubular than a batch reactor, it offers faster mixing, higher productivity and better integration to other bioprocessing steps. With increasing interest of integrated continuous biomanufacturing, the use of tubular reactors in industrial settings offers clear advantages.

  2. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-12

    ... COMMISSION Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES:...

  3. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    SciTech Connect

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  4. On-Line NDE for Advanced Reactor Designs

    NASA Astrophysics Data System (ADS)

    Nakagawa, N.; Inanc, F.; Thompson, R. B.; Junker, W. R.; Ruddy, F. H.; Beatty, J. M.; Arlia, N. G.

    2003-03-01

    This expository paper introduces the concept of on-line sensor methodologies for monitoring the integrity of components in next generation power systems, and explains general benefits of the approach, while describing early conceptual developments of suitable NDE methodologies. The paper first explains the philosophy behind this approach (i.e. the design-for-inspectability concept). Specifically, we describe where and how decades of accumulated knowledge and experience in nuclear power system maintenance are utilized in Generation IV power system designs, as the designs are being actively developed, in order to advance their safety and economy. Second, we explain that Generation IV reactor design features call for the replacement of the current outage-based maintenance by on-line inspection and monitoring. Third, the model-based approach toward design and performance optimization of on-line sensor systems, using electromagnetic, ultrasonic, and radiation detectors, will be explained. Fourth, general types of NDE inspections that are considered amenable to on-line health monitoring will be listed. Fifth, we will describe specific modeling developments to be used for radiography, EMAT UT, and EC detector design studies.

  5. Advances in crack-arrest technology for reactor pressure vessels

    SciTech Connect

    Bass, B.R.; Pugh, C.E.

    1988-01-01

    The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs.

  6. Advanced Test Reactor probabilistic risk assessment methodology and results summary

    SciTech Connect

    Eide, S.A.; Atkinson, S.A.; Thatcher, T.A.

    1992-01-01

    The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 report documents a comprehensive and state-of-the-art study to establish and reduce the risk associated with operation of the ATR, expressed as a mean frequency of fuel damage. The ATR Level 1 PRA effort is unique and outstanding because of its consistent and state-of-the-art treatment of all facets of the risk study, its comprehensive and cost-effective risk reduction effort while the risk baseline was being established, and its thorough and comprehensive documentation. The PRA includes many improvements to the state-of-the-art, including the following: establishment of a comprehensive generic data base for component failures, treatment of initiating event frequencies given significant plant improvements in recent years, performance of efficient identification and screening of fire and flood events using code-assisted vital area analysis, identification and treatment of significant seismic-fire-flood-wind interactions, and modeling of large loss-of-coolant accidents (LOCAs) and experiment loop ruptures leading to direct damage of the ATR core. 18 refs.

  7. Damage analysis and fundamental studies for fusion reactor materials development for the period March 1, 1991--February 28, 1994. Final report

    SciTech Connect

    Odette, G.R.; Lucas, G.E.

    1995-01-01

    The philosophy of the program at the University of California Santa Barbara has been to develop a fundamental understanding of both the basic damage processes and microstructural evolution that take place in a material during neutron irradiation and the consequent dimensional and mechanical property changes. This fundamental understanding can be used in conjunction with empirical data obtained from a variety of irradiation facilities to develop physically-based models of neutron irradiation effects in structural materials. The models in turn can be used to guide alloy development and to help extrapolate the irradiation data base to the fusion reactor regime. This philosophy is consistent with that of the national and international programs for developing structural materials for fusion reactors. During this period work has encompassed: (1) analysis of the degradation of the mechanical properties of austenitic stainless steels for the purpose of assessing the feasibility of using these steels in ITER; (2) examining helium effects on radiation damage in austenitic and ferritic stainless steels; (3) development and application of electropotential drop techniques to monitor the growth of cracks in steel specimens for a variety of specimen geometries (4) development of advanced methods of measuring fracture properties; (5) combining micromechanical modeling of fracture with finite element calculations of crack and notch-tip stress and strain fields to predict failure; (6) developing a data base on flow and fracture properties of ferritic steels. Each of these activities is described in more detail below and in greater detail in the attached publications.

  8. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    NASA Astrophysics Data System (ADS)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  9. Superconducting magnets for fusion applications

    SciTech Connect

    Henning, C.D.

    1987-07-02

    Fusion magnet technology has made spectacular advances in the past decade; to wit, the Mirror Fusion Test Facility and the Large Coil Project. However, further advances are still required for advanced economical fusion reactors. Higher fields to 14 T and radiation-hardened superconductors and insulators will be necessary. Coupled with high rates of nuclear heating and pulsed losses, the next-generation magnets will need still higher current density, better stability and quench protection. Cable-in-conduit conductors coupled with polyimide insulations and better steels seem to be the appropriate path. Neutron fluences up to 10/sup 19/ neutrons/cm/sup 2/ in niobium tin are achievable. In the future, other amorphous superconductors could raise these limits further to extend reactor life or decrease the neutron shielding and corresponding reactor size.

  10. Preliminary Advanced Test Reactor LEU Fuel Conversion Feasibility Study

    SciTech Connect

    G. S. Chang; R. G. Ambrosek

    2005-11-01

    The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. As a result, the ATR is a representative candidate for assessing the necessary modifications and evaluating the subsequent operating effects associated with low-enriched uranium (LEU) fuel conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed for the fuel cycle burnup comparison analysis. Using the current HEU 235U enrichment of 93.0 % as a baseline, an analysis can be performed to determine the LEU uranium density and 235U enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the 235U loading in the LEU core, such that the differences in Keff between the HEU and LEU core can be minimized for operation at 150 EFPD with a total core power of 115 MW. The Monte-Carlo with ORIGEN-2 (MCWO) method was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the LEU core conversion designer should be able to optimize the 235U content of each fuel plate, so that the Keff and relative radial fission heat flux profile are similar to the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Upgraded Final Safety Analysis Report (UFSAR) safety requirements, a further study will be required in order to investigate the detailed radial, axial, and azimuthal heat flux profile variations versus EFPDs.

  11. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    SciTech Connect

    T. R. Allen; J. B. Benson; J. A. Foster; F. M. Marshall; M. K. Meyer; M. C. Thelen

    2009-05-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  12. Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

    NASA Astrophysics Data System (ADS)

    Hançerlioğullari, Aybaba; Cini, Mesut; Güdal, Murat

    2013-08-01

    Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B-V-VI nuclear data. Around the fusion chamber, molten salts Flibe (LI2BeF4), lead-lithium (PbLi), Li-Sn, thin-lityum (Li20Sn80) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI2BeF4) and Li-Sn, for blanket layers. In case of usage of the Flibe (LI2BeF4), PbLi, and thin-lityum (Li20Sn80) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.

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

    SciTech Connect

    Smith, Roger J.

    2008-10-15

    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{sub 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{sub e}, n{sub e}, and B{sub ||} 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{sub e}B{sub ||} product and higher n{sub 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.

  14. Low Temperature Fatigue Properties of Advanced Cyanate-Ester Blends after Reactor Irradiation

    NASA Astrophysics Data System (ADS)

    Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.; Fabian, P. E.; Munshi, N. A.; Feucht, S. W.

    2004-06-01

    Fiber reinforced composites offer a broad spectrum of applications due to their excellent material performance under demanding conditions. Therefore, these materials will also be employed as insulation systems for the superconducting magnets in fusion devices. However, high doses of gamma and neutron irradiation lead to a drastic damage mostly of the organic matrices, such as pure epoxy resins. An improvement of these composites with regard to higher radiation resistance is of special importance to ensure stable coil operation over the plant lifetime. Recently, a series of advanced S2-glass fiber composites was developed, which consist of novel cyanate ester (CE) blends. All systems were irradiated in the TRIGA reactor (Vienna, Austria) to a neutron fluence of 1×1021 and 1×1022 m-2 (E>0.1 MeV), in order to assess the radiation hardness of their ultimate tensile strength. Furthermore, the material performance under cyclic load was investigated by tension-tension fatigue measurements at 77 K in view of the pulsed ITER operating conditions.

  15. Low Temperature Fatigue Properties of Advanced Cyanate-Ester Blends after Reactor Irradiation

    SciTech Connect

    Bittner-Rohrhofer, K.; Humer, K.; Weber, H.W.; Fabian, P.E.; Munshi, N.A.; Feucht, S.W.

    2004-06-28

    Fiber reinforced composites offer a broad spectrum of applications due to their excellent material performance under demanding conditions. Therefore, these materials will also be employed as insulation systems for the superconducting magnets in fusion devices. However, high doses of gamma and neutron irradiation lead to a drastic damage mostly of the organic matrices, such as pure epoxy resins. An improvement of these composites with regard to higher radiation resistance is of special importance to ensure stable coil operation over the plant lifetime. Recently, a series of advanced S2-glass fiber composites was developed, which consist of novel cyanate ester (CE) blends. All systems were irradiated in the TRIGA reactor (Vienna, Austria) to a neutron fluence of 1x1021 and 1x1022 m-2 (E>0.1 MeV), in order to assess the radiation hardness of their ultimate tensile strength. Furthermore, the material performance under cyclic load was investigated by tension-tension fatigue measurements at 77 K in view of the pulsed ITER operating conditions.

  16. Big fusion, little fusion

    NASA Astrophysics Data System (ADS)

    Chen, Frank; ddtuttle

    2016-08-01

    In reply to correspondence from George Scott and Adam Costley about the Physics World focus issue on nuclear energy, and to news of construction delays at ITER, the fusion reactor being built in France.

  17. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  18. Fusion Studies in Japan

    NASA Astrophysics Data System (ADS)

    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.

  19. Advanced Fusion Power Plant Studies. Annual Report for 1999

    SciTech Connect

    Chan, V.S.; Chu, M.S.; Greenfield, C.M.; Kinsey, J.E.; et al.

    2000-01-01

    Significant progress in physics understanding of the reversed shear advanced tokamak regime has been made since the last ARIES-RS study was completed in 1996. The 1999 study aimed at updating the physics design of ARIES-RS, which has been renamed ARIES-AT, using the improved understanding achieved in the last few years. The new study focused on: Improvement of beta-limit stability calculations to include important non-ideal effects such as resistive wall modes and neo-classical tearing modes; Use of physics based transport model for internal transport barrier (ITB) formation and sustainment; Comparison of current drive and rotational flow drive using fast wave, electron cyclotron wave and neutral particle beam; Improvement in heat and particle control; Integrated modeling of the optimized scenario with self-consistent current and transport profiles to study the robustness of the bootstrap alignment, ITB sustainment, and stable path to high beta and high bootstrap fraction operation.

  20. Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

    SciTech Connect

    White, J.D.; Monson, L.R.; Carrol, D.G.; Dayal, Y.; Argonne National Lab., IL; General Electric Co., San Jose, CA )

    1989-01-01

    Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs.