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Sample records for advanced blanket performance

  1. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    International Nuclear Information System (INIS)

    WONG, CPC; MALANG, S; NISHIO, S; RAFFRAY, R; SAGARA, S

    2002-01-01

    OAK A271 ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS. First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

  2. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Malang, S.; Nishio, S.; Raffray, R.; Sagara, A.

    2002-01-01

    First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

  3. Limitations on blanket performance

    International Nuclear Information System (INIS)

    Malang, S.

    1999-01-01

    The limitations on the performance of breeding blankets in a fusion power plant are evaluated. The breeding blankets will be key components of a plant and their limitations with regard to power density, thermal efficiency and lifetime could determine to a large degree the attractiveness of a power plant. The performance of two rather well known blanket concepts under development in the frame of the European Blanket Programme is assessed and their limitations are compared with more advanced (and more speculative) concepts. An important issue is the question of which material (structure, breeder, multiplier, coatings) will limit the performance and what improvement would be possible with a 'better' structural material. This evaluation is based on the premise that the performance of the power plant will be limited by the blankets (including first wall) and not by other components, e.g. divertors, or the plasma itself. However, the justness of this premise remains to be seen. It is shown that the different blanket concepts cover a large range of allowable power densities and achievable thermal efficiencies, and it is concluded that there is a high incentive to go for better performance in spite of possibly higher blanket cost. However, such high performance blankets are usually based on materials and technologies not yet developed and there is a rather high risk that the development could fail. Therefore, it is explained that a part of the development effort should be devoted to concepts where the materials and technologies are more or less in hand in order to ensure that blankets for a DEMO reactor can be developed and tested in a given time frame. (orig.)

  4. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li 2 ZrO 3 was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li 2 O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D- 3 He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view

  5. Development of advanced blanket materials for solid breeder blanket of fusion reactor

    International Nuclear Information System (INIS)

    Ishitsuka, E.

    2002-01-01

    Advanced solid breeding blanket design in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high dose of neutron irradiation. Therefore, the development of such advanced blanket materials is indispensable. In this paper, the cooperation activities among JAERI, universities and industries in Japan on the development of these advanced materials are reported. Advanced tritium breeding material to prevent the grain growth in high temperature had to be developed because the tritium release behavior degraded by the grain growth. As one of such materials, TiO 2 -doped Li 2 TiO 3 has been studied, and TiO 2 -doped Li 2 TiO 3 pebbles was successfully fabricated. For the advanced neutron multiplier, the beryllium intermetallic compounds that have high melting point and good chemical stability have been studied. Some characterization of Be 12 Ti was studied. The pebble fabrication study for Be 12 Ti was also performed and Be 12 Ti pebbles were successfully fabricated. From these activities, the bright prospect to realize the DEMO blanket by the application of TiO 2 -doped Li 2 TiO 3 and beryllium intermetallic compounds was obtained. (author)

  6. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-1 through 4 and PULSAR 1 and 2. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. Also, the requirements of engineering and physics systems for a pulsed reactor were evaluated by the PULSAR design studies. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies

  7. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  8. Effect of blanket assembly shuffling on LMR neutronic performance

    International Nuclear Information System (INIS)

    Khalil, H.; Fujita, E.K.

    1987-01-01

    Neutronic analyses of advanced liquid-metal reactors (LMRs) have generally been performed with assemblies in different batches scatter-loaded but not shuffled among the core lattice positions between cycles. While this refueling approach minimizes refueling time, significant improvements in thermal performance are believed to be achievable by blanket assembly shuffling. These improvements, attributable to mitigation of the early-life overcooling of the blankets, include reductions in peak clad temperatures and in the temperature gradients responsible for thermal striping. Here the authors summarize results of a study performed to: (1) assess whether the anticipated gains in thermal performance can be realized without sacrificing core neutronic performance, particularly the burnup reactivity swing rho/sub bu/, which determines the rod ejection worth; (2) determine the effect of various blanket shuffling operations on reactor performance; and (3) determine whether shuffling strategies developed for an equilibrium (plutonium-fueled) core can be applied during the transition from an initial uranium-fueled core as is being considered in the US advanced LMR program

  9. Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

    International Nuclear Information System (INIS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

    2006-10-01

    Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

  10. Effects of buffer thickness on ATW blanket performances

    International Nuclear Information System (INIS)

    Yang, Won Sik

    2001-01-01

    This paper presents the preliminary results of target and buffer design studies for a lead-bismuth eutectic (LBE) cooled accelerator transmutation of waste (ATW) system, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using an 840 MWt LBE cooled ATW design, the effects of buffer thickness on the blanket performances have been studied. Varying the buffer thickness for a given blanket configuration, system performances have been estimated by a series of calculations using MCNPX and REBUS-3 codes. The effects of source importance change are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. As the irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. The results show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable

  11. Effects of buffer thickness on ATW blanket performance

    International Nuclear Information System (INIS)

    Yang, W. S.; Mercatali, L.; Taiwo, T. A.; Hill, R. N.

    2001-01-01

    This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy ( and lt; 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level

  12. Effects of Buffer Thickness on ATW Blanket Performance

    International Nuclear Information System (INIS)

    Yang, W.S.; Mercatali, L.; Taiwo, T.A.; Hill, R.N.

    2002-01-01

    This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level. (authors)

  13. Advancement in tritium transport simulations for solid breeding blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Ying, Alice, E-mail: ying@fusion.ucla.edu [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Zhang, Hongjie [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Merrill, Brad J. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    In this paper, advancement on tritium transport simulations was demonstrated for a solid breeder blanket HCCR TBS, where multi-physics and detailed engineering descriptions are considered using a commercial simulation code. The physics involved includes compressible purge gas fluid flow, heat transfer, chemical reaction, isotope swamping effect, and tritium isotopes mass transport. The strategy adopted here is to develop numerical procedures and techniques that allow critical details of material, geometric and operational heterogeneity in a most complete engineering description of the TBS being incorporated into the simulation. Our application focuses on the transient assessment in view of ITER being pulsed operations. An immediate advantage is a more realistic predictive and design analysis tool accounting pulsed operations induced temperature variations which impact helium purge gas flow as well as Q{sub 2} composition concentration time and space evolutions in the breeding regions. This affords a more accurate prediction of tritium permeation into the He coolant by accounting correct temperature and partial pressure effects and realistic diffusion paths. The analysis also shows that by introducing by-pass line to accommodate ITER pulsed operations in the TES loop allows tritium extraction design being more cost effective.

  14. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

    2015-11-04

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and

  15. Advanced methods comparisons of reaction rates in the Purdue Fast Breeder Blanket Facility

    International Nuclear Information System (INIS)

    Hill, R.N.; Ott, K.O.

    1988-01-01

    A review of worldwide results revealed that reaction rates in the blanket region are generally underpredicted with the discrepancy increasing with penetration; however, these results vary widely. Experiments in the large uniform Purdue Fast Breeder Blanket Facility (FBBF) blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50 group cross sections), a consistent Calculated/Experimental (C/E) drop-off was observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the blanket is necessary for agreement with experiments. The usefulness of refined group constant generation utilizing specialized weighting spectra and transport theory methods in correcting this discrepancy was analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The surprising result was that transport methods had no effect on the blanket deviations; thus, transport theory considerations do not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by some approximations which are applied in all current methods. 27 refs., 3 figs., 1 tab

  16. Performance of silvered Teflon (trademark) thermal control blankets on spacecraft

    Science.gov (United States)

    Pippin, Gary; Stuckey, Wayne; Hemminger, Carol

    1993-01-01

    Silverized Teflon (Ag/FEP) is a widely used passive thermal control material for space applications. The material has a very low alpha/e ratio (less than 0.1) for low operating temperatures and is fabricated with various FEP thicknesses (as the Teflon thickness increases, the emittance increases). It is low outgassing and, because of its flexibility, can be applied around complex, curved shapes. Ag/FEP has achieved multiyear lifetimes under a variety of exposure conditions. This has been demonstrated by the Long Duration Exposure Facility (LDEF), Solar Max, Spacecraft Charging at High Altitudes (SCATHA), and other flight experiments. Ag/FEP material has been held in place on spacecraft by a variety of methods: mechanical clamping, direct adhesive bonding of tapes and sheets, and by Velcro(TM) tape adhesively bonded to back surfaces. On LDEF, for example, 5-mil blankets held by Velcro(TM) and clamping were used for thermal control over 3- by 4-ft areas on each of 17 trays. Adhesively bonded 2- and 5-mil sheets were used on other LDEF experiments, both for thermal control and as tape to hold other thermal control blankets in place. Performance data over extended time periods are available from a number of flights. The observed effects on optical properties, mechanical properties, and surface chemistry will be summarized in this paper. This leads to a discussion of performance life estimates and other design lessons for Ag/FEP thermal control material.

  17. Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

    2016-11-01

    Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

  18. Heating performances of a IC in-blanket ring array

    Energy Technology Data Exchange (ETDEWEB)

    Bosia, G., E-mail: gbosia@to.infn.it [Department of Physics, University of Turin (Italy); Ragona, R. [Laboratory for Plasma Physics-LPP-ERM/KMS, Brussels (Belgium)

    2015-12-10

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  19. Conceptual study on high performance blanket in a spherical tokamak fusion-driven transmuter

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Yican

    2000-01-01

    A preliminary conceptual design on high performance dual-cooled blanket of fusion-driven transmuter is presented based on neutronic calculation. The dual-cooled system has some attractive advantages when utilized in transmutation of HLW (High Level Wastes). The calculation results show that this kind of blanket could safely transmute about 6 ton minor actinides (produced by 170 GW(e) Year PWRs approximately) and 0.4 ton fission products per year, and output 12 GW thermal power. In addition, the variation of power and critical factor of this blanket is relatively little during its 1-year operation period. This blanket is also tritium self-sustainable

  20. A solid-breeder blanket and power conversion system for the Mirror Advanced Reactor Study (MARS)

    International Nuclear Information System (INIS)

    Bullis, R.; Clarkson, I.

    1983-01-01

    A solid-breeder blanket has been designed for a commercial fusion power reactor based on the tandem mirror concept (MARS). The design utilizes lithium oxide, cooled by helium which powers a conventional steam electric generating cycle. Maintenance and fabricability considerations led to a modular configuration 6 meters long which incorporates two magnets, shield, blanket and first wall. The modules are arranged to form the 150 meter long reactor central cell. Ferritic steel is used for the module primary structure. The lithium oxide is contained in thin-walled vanadium alloy tubes. A tritium breeding ratio of 1.25 and energy multiplication of 1.1 is predicted. The blanket design appears feasible with only a modest advance in current technology

  1. Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

    International Nuclear Information System (INIS)

    Swami, H.L.; Danani, C.

    2015-01-01

    In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

  2. Lithium-vanadium advanced blanket development. ITER final report on U.S. contribution: Task T219/T220

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Mattas, R.F. [comps.

    1997-07-01

    The objective of this task is to develop the required data base and demonstrate the performance of a liquid lithium-vanadium advanced blanket design. The task has two main activities related to vanadium structural material and liquid lithium system developments. The vanadium alloy development activity included four subtasks: (1.1) baseline mechanical properties of non irradiated base metal and weld metal joints; (1.2) compatibility with liquid lithium; (1.3) material irradiation tests; and (1.4) development of material manufacturing and joining methods. The lithium blanket technology activity included four subtasks: (2.1) electrical insulation development and testing for liquid metal systems; (2.2) MHD pressure drop and heat transfer study for self-cooled liquid metal systems; (2.3) chemistry of liquid lithium; and (2.4) design, fabrication and testing of ITER relevant size blanket mockups. A summary of the progress and results obtained during the period 1995 and 1996 in each of the subtask areas is presented in this report.

  3. Lithium-vanadium advanced blanket development. ITER final report on U.S. contribution: Task T219/T220

    International Nuclear Information System (INIS)

    Smith, D.L.; Mattas, R.F.

    1997-07-01

    The objective of this task is to develop the required data base and demonstrate the performance of a liquid lithium-vanadium advanced blanket design. The task has two main activities related to vanadium structural material and liquid lithium system developments. The vanadium alloy development activity included four subtasks: (1.1) baseline mechanical properties of non irradiated base metal and weld metal joints; (1.2) compatibility with liquid lithium; (1.3) material irradiation tests; and (1.4) development of material manufacturing and joining methods. The lithium blanket technology activity included four subtasks: (2.1) electrical insulation development and testing for liquid metal systems; (2.2) MHD pressure drop and heat transfer study for self-cooled liquid metal systems; (2.3) chemistry of liquid lithium; and (2.4) design, fabrication and testing of ITER relevant size blanket mockups. A summary of the progress and results obtained during the period 1995 and 1996 in each of the subtask areas is presented in this report

  4. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    Energy Technology Data Exchange (ETDEWEB)

    Pereslavtsev, Pavel, E-mail: pavel.pereslavtsev@kit.edu [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Boltzmannstrasse 2, 85748 Garching (Germany); Fischer, Ulrich [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, {sup 6}Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  5. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    International Nuclear Information System (INIS)

    Pereslavtsev, Pavel; Bachmann, Christian; Fischer, Ulrich

    2016-01-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, "6Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  6. Performance evaluation on force control for ITER blanket installation

    Energy Technology Data Exchange (ETDEWEB)

    Aburadani, A., E-mail: aburadani.atsushi@jaea.go.jp [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Nakahira, M.; Hamilton, D.; Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation.

  7. Performance evaluation on force control for ITER blanket installation

    International Nuclear Information System (INIS)

    Aburadani, A.; Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S.; Nakahira, M.; Hamilton, D.; Tesini, A.

    2013-01-01

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation

  8. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

  9. Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

    International Nuclear Information System (INIS)

    Patterson-Hine, F.A.

    1984-05-01

    Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs. The effects of processing on blanket performance have been assessed for three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The level of salt processing was found to have little effect on the behavior of the blanket during reactor operation; however, significant effects were observed during the decay period after reactor shutdown

  10. ITER structural design criteria and their extension to advanced reactor blankets

    International Nuclear Information System (INIS)

    Majumdar, S.; Kalinin, G.

    2000-01-01

    Applications of the recent ITER structural design criteria (ISDC) are illustrated by two components. First, the low-temperature-design rules are applied to copper alloys that are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures. Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. Next, the high-temperature-design rules of ISDC are applied to evaporation of lithium and vapor extraction (EVOLVE), a blanket design concept currently being investigated under the US Advanced Power Extraction (APEX) program. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method

  11. Advanced Electrochemical Machining (ECM) for tungsten surface micro-structuring in blanket applications

    Energy Technology Data Exchange (ETDEWEB)

    Holstein, Nils, E-mail: nils.holstein@kit.edu [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Baden-Württemberg (Germany); Krauss, Wolfgang; Konys, Jürgen [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Baden-Württemberg (Germany); Heuer, Simon; Weber, Thomas [Research Center Jülich, Institute of Energy- and Climate Research – Plasma Physics (IEK-4), D-52425 Jülich (Germany)

    2016-11-01

    Highlights: • Electrochemical Machining is an appropriate tool for tungsten shaping. • Progress in shaping achieved by combination of ECM with advanced micro-lithography. • Application in First Wall for connection of plasma facing material to breeder blanket. • Successful development of adhesion promotors by ECM for plasma spraying interlayers. • Microstructure electrochemical manufacturing of tungsten in sizes of 100 μm achieved. - Abstract: Plasma facing components for fusion applications must have to exhibit long-term stability under extreme physical conditions, and therefore any material imperfections caused by mechanical and/or thermal stresses in the shaping processes cannot be tolerated due to a high risk of possible technical failures under fusion conditions. To avoid such defects, the method of Electrochemical Machining (ECM) enables a complete defect-free processing of removal of tungsten material during the desired shaping, also for high penetration depths. Furthermore, supported by lithographic mask pretreatment, three-dimensional distinct geometric structures can be positive-imaged via the directional galvanic dissolution applying M-ECM process into the tungsten bulk material. New required applications for tungsten components, e.g. as adhesion promotors in W-surfaces to enable sure grip and bonding of thick plasma-spraying layers for blanket components, will define the way of further miniaturization of well-established millimeter dimensioned M-ECM shaping processes to dimensions of 100 μm and furthermore down to 50 μm. Besides current M-ECM limits the article describes inevitable needs of further developments for mask resists, mask materials and the resulting ECM parameters, to reach the needed accuracy in tungsten microstructure. The achieved progress and observed correlations of processing parameters will be manifested by produced demonstrators made by the new “μM”-ECM process.

  12. Advanced Electrochemical Machining (ECM) for tungsten surface micro-structuring in blanket applications

    International Nuclear Information System (INIS)

    Holstein, Nils; Krauss, Wolfgang; Konys, Jürgen; Heuer, Simon; Weber, Thomas

    2016-01-01

    Highlights: • Electrochemical Machining is an appropriate tool for tungsten shaping. • Progress in shaping achieved by combination of ECM with advanced micro-lithography. • Application in First Wall for connection of plasma facing material to breeder blanket. • Successful development of adhesion promotors by ECM for plasma spraying interlayers. • Microstructure electrochemical manufacturing of tungsten in sizes of 100 μm achieved. - Abstract: Plasma facing components for fusion applications must have to exhibit long-term stability under extreme physical conditions, and therefore any material imperfections caused by mechanical and/or thermal stresses in the shaping processes cannot be tolerated due to a high risk of possible technical failures under fusion conditions. To avoid such defects, the method of Electrochemical Machining (ECM) enables a complete defect-free processing of removal of tungsten material during the desired shaping, also for high penetration depths. Furthermore, supported by lithographic mask pretreatment, three-dimensional distinct geometric structures can be positive-imaged via the directional galvanic dissolution applying M-ECM process into the tungsten bulk material. New required applications for tungsten components, e.g. as adhesion promotors in W-surfaces to enable sure grip and bonding of thick plasma-spraying layers for blanket components, will define the way of further miniaturization of well-established millimeter dimensioned M-ECM shaping processes to dimensions of 100 μm and furthermore down to 50 μm. Besides current M-ECM limits the article describes inevitable needs of further developments for mask resists, mask materials and the resulting ECM parameters, to reach the needed accuracy in tungsten microstructure. The achieved progress and observed correlations of processing parameters will be manifested by produced demonstrators made by the new “μM”-ECM process.

  13. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

  14. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1995-09-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified

  15. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1996-01-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as primary blanket materials, which have the greatest influence in determining the overall design and performance, and secondary blanket materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified. (orig.)

  16. Nuclear performance optimization of the Be/Li/Th blanket for the fusion breeder

    International Nuclear Information System (INIS)

    Lee, J.D.; Bandini, B.R.

    1985-01-01

    More rigorous nuclear analysis, including treatment of resonance self-shielding effects coupled with an optimization procedure, has resulted in improved performance of the Be/Li/Th blanket. Net U-233 breeding ratio has increased 36% (to 0.84) while at an average U-233/Th ratio of 0.5 a/o average energy multiplication has increased only 12% (to 2.1) compared with earlier results

  17. Blanket materials for fusion reactors: comparisons of thermochemical performance

    International Nuclear Information System (INIS)

    Johnson, C.E.; Fischer, A.K.; Tetenbaum, M.

    1984-01-01

    Thermodynamic calculations have been made to predict the thermochemical performance of the fusion reactor breeder materials, Li 2 O, LiAlO 2 , and Li 4 SiO 4 in the temperature range 900 to 1300 0 K and in the oxygen activity range 10 -25 to 10 -5 . Except for a portion of these ranges, the performance of LiAlO 2 is predicted to be better than that of Li 2 O and Li 4 SiO 4 . The protium purge technique for enhancing tritium release is explored for the Li 2 O system; it appears advantageous at higher temperatures but should be used cautiously at lower temperatures. Oxygen activity is an important variable in these systems and must be considered in executing and interpreting measurements on rates of tritium release, the form of released tritium, diffusion of tritiated species and their identities, retention of tritium in the condensed phase, and solubility of hydrogen isotope gases

  18. Control of Sound Radiation and Reflection With Advanced Smart Foam Blankets

    National Research Council Canada - National Science Library

    Fuller, Chris

    2003-01-01

    .... The past few years of the project have demonstrated the high potential of using smart foam blankets for efficiently reducing the interior sound levels in the payloads of launch vehicles in the low...

  19. US-DOE Fusion-Breeder Program: blanket design and system performance

    International Nuclear Information System (INIS)

    Lee, J.D.

    1983-01-01

    Conceptual design studies are being used to assess the technical and economic feasibility of fusion's potential to produce fissile fuel. A reference design of a fission-suppressed blanket using conventional materials is under development. Theoretically, a fusion breeder that incorporates this fusion-suppressed blanket surrounding a 3000-MW tandem mirror fusion core produces its own tritium plus 5600 kg of 233 U per year. The 233 U could then provide fissile makeup for 21 GWe of light-water reactor (LWR) power using a denatured thorium fuel cycle with full recycle. This is 16 times the net electric power produced by the fusion breeder (1.3 GWe). The cost of electricity from this fusion-fission system is estimated to be only 23% higher than the cost from LWRs that have makeup from U 3 O 8 at present costs (55 $/kg). Nuclear performance, magnetohydrodynamics (MHD), radiation effects, and other issues concerning the fission-suppressed blanket are summarized, as are some of the present and future objectives of the fusion breeder program

  20. The second advanced lead lithium blanket concept using ODS steel as structural material and SiCf/SiC flow channel inserts as electrical and thermal insulators (Task PPA 2.5). Final report

    International Nuclear Information System (INIS)

    Norajitra, P.; Buehler, L.; Fischer, U.

    1999-12-01

    Preparatory work on the advanced dual coolant (A-DCL) blanket concept using SiC f /SiC flow channel inserts as electrical and thermal insulators has been carried out at the Forschungszentrum Karlsruhe in co-operation with CEA as a conceptual design proposal to the EU fusion power plant study planned to be launched in 2000 within the framework of the EU fusion programme with the main objective of specifying the characteristics of an attractive and viable commercial D-T fusion power plant. The basic principles and design characteristics of this A-DCL blanket concept are presented and its potential with regard to performance (neutron wall load, lifetime, availability) is discussed in this report. The results of this study show that the A-DCL blanket concept has a high potential for further development due to its high thermal efficiency and its simple concept solution. (orig.) [de

  1. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Waganer, L.M.

    1985-01-01

    The economic impact of employing the highly ranked blankets in the Blanket Comparison and Selection Study (BCSS) was evaluated in the context of both a tokamak and a tandem mirror power reactor (TMR). The economic evaluation criterion was determined to be the cost of electricity. The influencing factors that were considered are the direct cost of the blankets and related systems; the annual cost of blanket replacement; and the performance of the blanket, heat transfer, and energy conversion systems. The technical and cost bases for comparison were those of the STARFIRE and Mirror Advanced Reactor Study conceptual design power plants. The economic evaluation results indicated that the nitrate-salt-cooled blanket concept is an economically attractive concept for either reactor type. The water-cooled, solid breeder blanket is attractive for the tokamak and somewhat less attractive for the TMR. The helium-cooled, liquidlithium breeder blanket is the least economically desirable of higher ranked concepts. The remaining self-cooled liquid-metal and the helium-cooled blanket concepts represent moderately attractive concepts from an economic standpoint. These results are not in concert with those found in the other BCSS evaluation areas (engineering feasibility, safety, and research and development (R and D) requirements). The blankets faring well economically had generally lower cost components, lower pumping power requirements, and good power production capability. On the other hand, helium- and lithium-cooled systems were preferred from the standpoints of safety, engineering feasibility, and R and D requirements

  2. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.; Jankhah, M.H.

    1979-01-01

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  3. Octalithium plumbate as breeding blanket ceramic: Neutronic performances, synthesis and partial characterization

    International Nuclear Information System (INIS)

    Colominas, S.; Palermo, I.; Abellà, J.; Gómez-Ros, J.M.; Sanz, J.; Sedano, L.

    2012-01-01

    Highlights: ► Definition of a suitable configuration for the Li 8 PbO 6 breeding blanket design. ► Demonstration of the feasibility of Li 8 PbO 6 as a breeding material. ► Synthesis optimization in the Li 8 PbO 6 production. ► Characterization of Li 8 PbO 6 by X-ray phase analysis is discussed. - Abstract: A neutronic assessment of the performances of a helium-cooled Li 8 PbO 6 breeding blanket (BB) for the conceptual design of a DEMO fusion reactor is given. Different BB configurations have been considered in order to minimize the amount of beryllium required for neutron multiplication, including the use of graphite as reflector material. The calculated neutronic responses: tritium breeding ratio (TBR), power deposition in TF coils and power amplification factor, indicate the feasibility of Li 8 PbO 6 as breeding material. Furthermore, the synthesis and characterization of Li 8 PbO 6 by X-ray phase analysis are also discussed.

  4. Nuclear performance of target-blanket assemblies for electronuclear fuel production

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Gabriel, T.A.; Barish, J.

    1978-01-01

    The results of calculations of high energy transport carried out to evaluate the nuclear performance of several of the designs that have been proposed for electronuclear fuel production are presented. Topics covered include: results for 1-GeV protons and 1-GeV deuterons incident on a lithium target surrounded by a 238 U blanket; results for 1 GeV protons and 1-GeV deuterons incident on a thorium salt; results for 1-GeV protons, incident on a gas-cooled system fueled by either 238 UO 2 or 232 ThO 2 ; and results for protons in the energy range 0.5 to 10 GeV incident on a very large natural uranium target

  5. Mirror reactor blankets

    International Nuclear Information System (INIS)

    Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

    1976-01-01

    The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

  6. Blanket testing in NET

    International Nuclear Information System (INIS)

    Chazalon, M.; Daenner, W.; Libin, B.

    1989-01-01

    The testing stages in NET for the performance assessment of the various breeding blanket concepts developed at the present time in Europe for DEMO (LiPb and ceramic blankets) and the requirements upon NET to perform these tests are reviewed. Typical locations available in NET for blanket testing are the central outboard segments and the horizontal ports of in-vessel sectors. These test positions will be connectable with external test loops. The number of test loops (helium, water, liquid metal) will be such that each major class of blankets can be tested in NET. The test positions, the boundary conditions and the external test loops are identified and the requirements for test blankets are summarized (author). 6

  7. Concept and nuclear performance of direct-enrichment fusion breeder blanket using UO2 powder

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Kasahara, Takayasu; An, Shigehiro

    1985-01-01

    A new concept is presented for direct enrichment of fissile fuel in the blanket of a fusion-fission hybrid reactor. The enriched fuel produced by this means can be used in fission reactors without reprocessing. The outstanding feature of the concept is the powdered form in which UO 2 fuel is placed in the reactor blanket, where it is irradiated to the requisite enrichment for use as fuel in burner reactor, e.g. 3%. After removal from blanket, the powder is mixed to homogenize the enrichment. Fuel pellets and assemblies are then fabricated from the powder without reprocessing. The concept of irradiating UO 2 in powder eliminates the problems of spatial nonuniformity in fissile enrichment, and of radiation damage to fuel clad, encountered in attempting to enrich prefabricated fuel. Powder mixing for homogenization brings the additional benefit of removing volatile fission products. Also burnable poison can be added, as necessary, after irradiation. An extensive neutronic parameter survey showed that the optimum blanket arrangement for this enrichment concept is one presenting a fission suppressing configuration and with beryllium adopted as moderator. By this arrangement, the average 239 Pu enrichment obtained on the natural UO 2 fuel in the blanket reaches 3% after only 0.56 MW.yr/m"2 exposure. A conceptual design is presented of the blanket, together with associated fusion breeder, from which, practical application of the concept is shown to be promising. (author)

  8. Membrane installation for enhanced up-flow anaerobic sludge blanket (UASB) performance.

    Science.gov (United States)

    Liu, Yin; Zhang, Kaisong; Bakke, Rune; Li, Chunming; Liu, Haining

    2013-09-01

    It is postulated that up-flow anaerobic sludge blanket (UASB) reactor efficiency can be enhanced by a membrane immersed in the reactor to operate it as an anaerobic membrane bioreactor (AnMBR) for low-strength wastewater treatment. This postulate was tested by comparing the performance with and without a hollow fiber microfiltration membrane module immersed in UASB reactors operated at two specific organic loading rates (SOLR). Results showed that membrane filtration enhanced process performance and stability, with over 90% total organic carbon (TOC) removal consistently achieved. More than 91% of the TOC removal was achieved by suspended biomass, while less than 6% was removed by membrane filtration and digestion in the membrane attached biofilm during stable AnMBRs operation. Although the membrane and its biofilm played an important role in initial stage of the high SOLR test, linear increased TOC removal by bulk sludge mainly accounted for the enhanced process performance, implying that membrane led to enhanced biological activity of the suspended sludge. The high retention of active fine sludge particles in suspension was the main reason for this significant improvement of performance and biological activity, which led to decreased SOLR with time to a theoretical optimal level around 2  g COD/g MLVSS·d and the establishment of a microbial community dominated by Methanothrix-like microbes. It was concluded that UASB process performance can be enhanced by transforming such to AnMBR operation when the loading rate is too high for sufficient sludge retention, and/or when the effluent water quality demands are especially stringent. Copyright © 2013 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  9. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

    The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

  10. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  11. Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

    International Nuclear Information System (INIS)

    Patterson-Hine, F.A.; Davidson, J.W.; Klein, D.E.; Lee, J.D.

    1985-01-01

    Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs: fluorination only, fluorination plus reductive extraction, and fluorination, plus reductive extraction, plus metal transfer. The effects of processing on blanket performance have been assessed for these three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis, which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The method of salt processing was found to have little affect on the level of radioactivity, toxicity, or the thermal behavior of the salt during operation of the reactor. The processing rates necessary to maintain the desired uranium concentrations in the suppressed-fission environment were quite low, which permitted only long-lived species to be removed from the salt. The effects of the processing therefore became apparent only after the radioactivity due to the short-lived species diminished. The effect of the additional processing (reductive extraction and metal transfer) could be seen after approximately 1 year of decay, but were not significant at times closer to shutdown. The reduced radioactivity and corresponding heat deposition were thus of no consequence in accident or maintenance situations. Net fissile production in the Be/MS blanket concept at a fusion power level of 3000 MW at 70% capacity ranged from 5100 kg/year to 5170 kg/year for uranium concentrations of 0.11% and 1.0% 233 U in thorium, respectively, with fluorination-only processing. The addition of processing by reductive extraction resulted in 5125 kg/year for the 0.11% 233 U case and 5225 kg/year for the 1.0% 233 U case

  12. Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Mas de les Valls, E., E-mail: elisabet.masdelesvalls@gits.ws [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Heat Engines, Barcelona (Spain); Batet, L. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Physics and Nuclear Engineering, Barcelona (Spain); Medina, V. de [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Sediment Transport Research Group, Department of Engineering Hydraulic, Marine and Environmental Engineering, Barcelona (Spain); Fradera, J. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Physics and Nuclear Engineering, Barcelona (Spain); Sanmarti, M. [bFUS-IREC, Jardins de les Dones de Negre 1, 08930 Sant Adria del Besos (Spain); Sedano, L.A. [EURATOM-CIEMAT Association, 28040 Madrid (Spain)

    2012-08-15

    Spanish Breeding Blanket Technology Programme TECNO{sub F}US is exploring the technological capabilities of a Dual-Coolant He/Pb15.7Li breeding blanket for DEMO and studying new breeding blanket design specifications. The progress of the channel conceptual design is being conducted in parallel with the extension of MHD computational capabilities of CFD tools and the underlying physics of MHD models. A qualification of MHD effects under present blanket design specifications and some approaches to their modelling were proposed by the authors in . The analysis was accomplished with the 2D transient algorithm from Sommeria and Moreau and implemented in the OpenFOAM CFD toolbox . The thermal coupling was implemented by means of the Boussinesq hypothesis. Previous analyses showed the need of improvement of FCI thickness and thermal properties in order to obtain a desirable liquid metal temperature gain of 300 Degree-Sign C. In the present study, an assessment through sensitivity and parametric analyses of the required FCI thickness is performed. Numerical simulations have been carried out considering a Robin-type thermal boundary condition which assumes 1D steady state thermal balance across the solid FCI and Eurofer layers. Such boundary condition has been validated with a fluid-solid coupled domain analysis. Results for the studied flow conditions and channel dimensions show that, in order to obtain a liquid metal temperature gain of about 300 Degree-Sign C, the required FCI material should have a very small effective heat transfer coefficient ((k/{delta}) {<=} 1 W/m{sup 2}K) and fluid velocities should be about 0.2 m/s or less. Moreover, special attention has to be placed on the temperature difference across the FCI layer. However, for a maximised liquid metal thermal gain, higher velocities would be preferable, what would also imply a reduced temperature difference across the FCI layer.

  13. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  14. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

    In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

  15. Advances in the optimisation of apparel heating products: A numerical approach to study heat transport through a blanket with an embedded smart heating system

    International Nuclear Information System (INIS)

    Neves, S.F.; Couto, S.; Campos, J.B.L.M.; Mayor, T.S.

    2015-01-01

    The optimisation of the performance of products with smart/active functionalities (e. g. in protective clothing, home textiles products, automotive seats, etc.) is still a challenge for manufacturers and developers. The aim of this study was to optimise the thermal performance of a heating product by a numerical approach, by analysing several opposing requirements and defining solutions for the identified limitations, before the construction of the first prototype. A transfer model was developed to investigate the transport of heat from the skin to the environment, across a heating blanket with an embedded smart heating system. Several parameters of the textile material and of the heating system were studied, in order to optimise the thermal performance of the heating blanket. Focus was put on the effects of thickness and thermal conductivity of each layer, and on parameters associated with the heating elements, e.g. position of the heating wires relative to the skin, distance between heating wires, applied heating power, and temperature range for operation of the heating system. Furthermore, several configurations of the blanket (and corresponding heating powers) were analysed in order to minimise the heat loss from the body to the environment, and the temperature distribution along the skin. The results show that, to ensure an optimal compromise between the thermal performance of the product and the temperature oscillation along its surface, the distance between the wires should be small (and not bigger than 50 mm), and each layer of the heating blanket should have a specific thermal resistance, based on the expected external conditions during use and the requirements of the heating system (i.e. requirements regarding energy consumption/efficiency and capacity to effectively regulate body exchanges with surrounding environment). The heating system should operate in an ON/OFF mode based on the body heating needs and within a temperature range specified based on

  16. Advanced fuel technology and performance

    International Nuclear Information System (INIS)

    1985-10-01

    The purpose of the Advisory Group Meeting on Advanced Fuel Technology and Performance was to review the experience of advanced fuel fabrication technology, its performance, peculiarities of the back-end of the nuclear fuel cycle with regard to all types of reactors and to outline the future trends. As a result of the meeting recommendations were made for the future conduct of work on advanced fuel technology and performance. A separate abstract was prepared for each of the 20 papers in this issue

  17. SIEX3: A correlated computer code for prediction of fast reactor mixed oxide fuel and blanket pin performance

    International Nuclear Information System (INIS)

    Baker, R.B.; Wilson, D.R.

    1986-04-01

    The SIEX3 computer program was developed to calculate the fuel and cladding performance of oxide fuel and oxide blanket pins irradiated in the fast neutron environment of a liquid metal cooled reactor. The code is uniquely designed to be accurate yet quick running and use a minimum of computer core storage. This was accomplished through the correlation of physically based models to very large data bases of irradiation test results. Data from over 200 fuel pins and over 800 transverse fuel microscopy samples were used in the calibrations

  18. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2009-01-01

    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

  19. Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

    Directory of Open Access Journals (Sweden)

    Toshio Wakabayashi

    2013-01-01

    Full Text Available An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

  20. Advanced Performance Hydraulic Wind Energy

    Science.gov (United States)

    Jones, Jack A.; Bruce, Allan; Lam, Adrienne S.

    2013-01-01

    The Jet Propulsion Laboratory, California Institute of Technology, has developed a novel advanced hydraulic wind energy design, which has up to 23% performance improvement over conventional wind turbine and conventional hydraulic wind energy systems with 5 m/sec winds. It also has significant cost advantages with levelized costs equal to coal (after carbon tax rebate). The design is equally applicable to tidal energy systems and has passed preliminary laboratory proof-of-performance tests, as funded by the Department of Energy.

  1. Recovering hydrogen production performance of upflow anaerobic sludge blanket reactor (UASBR) fed with galactose via repeated heat treatment strategy.

    Science.gov (United States)

    Sivagurunathan, Periyasamy; Anburajan, Parthiban; Kumar, Gopalakrishnan; Park, Jong-Hun; Kim, Sang-Hyoun

    2017-09-01

    This study evaluated the effect of repeated heat treatment towards the enhancement of hydrogen fermentation from galactose in an upflow anaerobic sludge blanket reactor with the hydraulic retention time of 6h and the operation temperature of 37°C. The hydrogen production rate (HPR) and hydrogen yield (HY) gradually increased up to 9.1L/L/d and 1.1mol/mol galactose, respectively, until the 33rd day of operation. When heat treatment at 80°C for 30min was applied, hydrogen production performance was enhanced by 37% with the enrichment of hydrogen producing bacteria population. The HPR and HY were achieved at 12.5L/L/d and 1.5mol/mol hexose, respectively, during further 30 cycles of reactor operation. The repeated heat treatment would be a viable strategy to warrant reliable continuous hydrogen production using mixed culture. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    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

  3. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

    Strebkov, Yu.; Sidorov, A.; Danilov, I.

    1994-01-01

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

  4. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1991-01-01

    In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

  5. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1990-11-01

    The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  7. Magnetoconvection in HCLL blankets

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Buehler, L.

    2014-01-01

    In the present work we consider magneto-convective flows in one of the proposed European liquid metal blankets that will be tested in the experimental fusion reactor ITER. Here the PbLi alloy is used as breeder material and helium as coolant. In order to finalize the design of the helium cooled lead lithium (HCLL) blanket, studies are still required to fully understand the behavior of the electrically conducting breeder under the influence of the intense magnetic field that confines the fusion plasma and in case of non-uniform thermal conditions. Liquid metal HCLL blanket flows are expected to be mainly driven by buoyancy forces caused by non-isothermal operating conditions due to neutron volumetric heating and cooling of walls, since only a weak forced ow is foreseen for tritium extraction in external ancillary systems. Buoyancy can therefore become very important and modify the velocity distribution and related heat transfer performance of the blanket. The present numerical study aims at clarifying the influence of electromagnetic and thermal coupling of neighboring fluid domains on magneto-convective flows in geometries relevant for the HCLL blanket concept. According to the last design review two internal cooling plates subdivide the fluid domain into three slender flow regions, which are thermally and electrically coupled through common walls. First a uniform volumetric heat source is considered to identify the basic convective patterns that establish in the liquid metal. Results are then compared with those obtained by applying a realistic radial distribution of the power density as obtained from a neutronic analysis. Velocity and temperature distributions are discussed for various volumetric heat sources and magnetic field strengths.

  8. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

    The design features of the hybrid reactor blanket and main parameters are presented. The stress analysis is performed by using computer codes SAP5p and SAP6 with the three kinds of blanket module loadings, i.e, the pressure of coolant, the blanket weight and the thermal loading. Numerical calculation results indicate that the stresses of the blanket are smaller than the allowable ones of the material, the blanket design is therefore reasonable

  9. Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

    International Nuclear Information System (INIS)

    Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

    1979-01-01

    In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

  10. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

  11. Performance evaluation of a mesophilic (37 deg. C) upflow anaerobic sludge blanket reactor in treating distiller's grains wastewater

    International Nuclear Information System (INIS)

    Gao Mengchun; She Zonglian; Jin Chunji

    2007-01-01

    The performance of a laboratory-scale upflow anaerobic sludge blanket (UASB) reactor treating distiller's grains wastewater was investigated for 420 days at 37 deg. C. After a successful start-up, 80-97.3% chemical oxygen demand (COD) removal efficiencies were achieved at hydraulic retention times (HRT) of 82-11 h with organic loading rates (OLR) of 5-48.3 kg COD m -3 d -1 . The biogas mainly consisted of methane and carbon dioxide, and the methane and carbon dioxide content in the biogas was 57-60 and 38-41%, respectively. The yield coefficient of methane production was 0.3182 l CH 4 g -1 COD removed until OLR at 33.3 kg COD m -3 d -1 , but afterwards began to decrease. The volatile fatty acid (VFA) in the effluent mainly consisted of acetate and propionate, accounting for more than 95% of total VFA as COD, and other VFA was detected at insignificant concentrations. The mesophilic granules developed in this study showed an excellent specific methanogenic activity (SMA) at 0.91 and 1.12 g methane COD g -1 VSS -1 d -1 using sucrose and acetate as individual substrates on day 200, respectively

  12. Performance experiments with alternative advanced teleoperator control modes for a simulated solar maximum satellite repair

    Science.gov (United States)

    Das, H.; Zak, H.; Kim, W. S.; Bejczy, A. K.; Schenker, P. S.

    1992-01-01

    Experiments are described which were conducted at the JPL Advanced Teleoperator Lab to demonstrate and evaluate the effectiveness of various teleoperator control modes in the performance of a simulated Solar Max Satellite Repair (SMSR) task. THe SMSR was selected as a test because it is very rich in performance capability requirements and it actually has been performed by two EVA astronauts in the Space Shuttle Bay in 1984. The main subtasks are: thermal blanket removal; installation of a hinge attachment for electrical panel opening; opening of electrical panel; removal of electrical connectors; relining of cable bundles; replacement of electrical panel; securing parts and cables; re-mate electrical connectors; closing of electrical panel; and reinstating thermal blanket. The current performance experiments are limited to thermal blanket cutting, electrical panel unbolting and handling electrical bundles and connectors. In one formal experiment even different control modes were applied to the unbolting and reinsertion of electrical panel screws subtasks. The seven control modes are alternative combinations of manual position and rate control with force feedback and remote compliance referenced to force-torque sensor information. Force-torque sensor and end effector position data and task completion times were recorded for analysis and quantification of operator performance.

  13. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

    2000-07-01

    This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

  14. Impact of material system thermomechanics and thermofluid performance on He-cooled ceramic breeder blanket designs with SiCf/SiC

    International Nuclear Information System (INIS)

    Ying, Alice Y.; Yokomine, Takehiko; Shimizu, Akihiko; Abdou, Mohamed; Kohyama, Akira

    2004-01-01

    This paper presents results from a recent effort initiated under the JUPITER-II collaborative program for high temperature gas-cooled blanket systems using SiC f /SiC as a structural material. Current emphasis is to address issues associated with the function of the helium gas considered in the DREAM and ARIES-I concepts by performing thermomechanical and thermofluid analysis. The objective of the analysis is to guide future research focus for a task in the project. It is found that the DREAM concept has the advantage of achieving uniform temperature without threatening blanket pebble bed integrity by differential thermal stress. However, its superiority needs to be further justified by investigating the feasibility and economic issues involved in the tritium extraction technology

  15. Impact of material system thermomechanics and thermofluid performance on He-cooled ceramic breeder blanket designs with SiCf/SiC

    International Nuclear Information System (INIS)

    Ying, A.Y.; Abdou, M.; Yokomine, T.; Shimizu, A.; Kohyama, A.

    2008-01-01

    This paper presents results from a recent effort initiated under the JUPITER-II collaborative program for high temperature gas-cooled blanket systems using SiC/SiC as a structural material. Current emphasis is to address issues associated with the function of the helium gas considered in the DREAM and ARIES-I concepts by performing thermomechanical and thermofluid analysis. The objective of the analysis is to guide future research focus for a task in the project. It is found that the DREAM concept has the advantage of achieving uniform temperature without threatening blanket pebble bed integrity by differential thermal stress. However, its superiority needs to be further justified by investigating the feasibility and economic issues involved in the tritium extraction technology. (author)

  16. Performance Tests of a Permeation Sensor for Test Blanket Modules Using Liquid Metal

    International Nuclear Information System (INIS)

    Choi, B. G.; Lee, D. W.; Lee, E. H.; Yoon, J. S.; Kim, S. K.; Shin, K. I.; Jin, H. G.

    2013-01-01

    The tritium extraction from a breeder is one of the key technologies and its methods have been investigated. For developing the tritium extraction methods and evaluating the amount of tritium in the system, a reliable and correct sensor is required to measure the hydrogen concentration in liquid metal breeder. There are several researches for developing the sensors in the ITER participants and especially, EU has developed the permeation sensors trying to selecting materials with low Serviette's constant (solubility) and high hydrogen diffusivity coefficient. However, EU's response time is still too long time about tens of minutes to measure the tritium concentration in the online system. We have been performing the preliminary tests with designed and fabricated sensors to solve the late response of sensor. However, we could not continue the tests because of the membrane's oxidation (pure Fe) and the difficulty of welding nonferrous metals. In present study, a permeation sensor made of vacuum flanges with a porous plate inside is proposed not only to eliminate the difficulty of the fabrication but to optimize the performance of sensor. The permeation sensor to measure the hydrogen isotopes in liquid metal breeder has been proposed and evaluated to overcome the limitation of a long response time for various shapes and materials. We found that the previous sensors have limitation; the oxidation problems (pure Fe) and the difficulty in welding (nonferrous metals). Therefore we proposed a permeation sensor with the vacuum flanges filled with porous disks to eliminate the problems. By using the CF flanges, the problem caused by welding is removed. But the permeable response time of sensors took a long time to reach the pressure equivalent

  17. LMFBR blanket physics project progress report No. 4

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Lanning, D.D.; Kaplan, I.; Supple, A.T.

    1973-01-01

    During the period covered by the report, July 1, 1972, through June 30, 1973, work was devoted to completion of experimental measurements and data analysis on Blanket Mockup No. 3, a graphite-reflected blanket, and to initiation of experimental work on Blanket Mockup No. 4, a steel-reflected assembly designed to mock up a demonstration plant blanket. Work was also carried out on the analysis of a number of advanced blanket concepts, including the use of high-albedo reflectors, the use of thorium in place of uranium in the blanket region, and the ''parfait'' or completely internal blanket concept. Finally, methods development work was initiated to develop the capability for making gamma heating measurements in the blanket mockups. (U.S.)

  18. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    Proust, E.; Giancarli, L.

    1992-01-01

    This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently investigated within the framework of the European Test-Blanket Development Programme

  19. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

    Selecting relevant features is in demand when a large data set is of interest in a classification task. It produces a tractable number of features that are sufficient and possibly improve the classification performance. This paper studies a statistical method of Markov blanket induction algorithm...... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....

  20. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    Gohar, Y.; Baker, C.C.; Smith, D.L.

    1987-10-01

    The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of ∼2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs

  1. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-12-01

    This report includes electromagnetic analyses for ITER shielding blanket modules, fabrication methods for the blanket modules and the back plate, the design and the fabrication methods for port limiter have been investigated. Studies on the runaway electron impact for Be armor have been also performed. (J.P.N.)

  2. Performance test of diamond-like carbon films for lubricating ITER blanket maintenance equipment under GPa-level high contact stress

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi

    2007-01-01

    Diamond-like carbon (DLC) coating was tested as a candidate solid lubricant for transmission gears of the maintenance equipment of the blanket of the ITER instead of an oil lubricant. The wear tests using the pin-on-disk method were performed on disks with SCM440 and SNCM420 as the base materials and coated with soft, layered, and hard DLCs. All cases satisfied the required allowable contact stress (2 GPa) and lifetime (10 4 cycles), and therefore the feasibility of the DLC coating was validated. Among the three types of DLCs, the soft DLC showed the best performance. (author)

  3. (D,T) Driven thorium hybrid blankets

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.; Khan, S.; Sahin, S.

    1983-01-01

    Recently, a project has started, with the aim to establish the neutronic performance and the basic design of an experimental fusionfission (hybrid) reactor facility, called AYMAN, in cylinderical geometry. The fusion reactor will have to be simulated by a (D,T) neutron generator. Fissile and fertile fuel will have to surround the neutron generator as a cylinderical blanket to simulate the boundary conditions of the hybrid blanket in a proper way. This geometry is consistent with Tandem Mirror Hybrid Blanket design and with most of the ICF blanket designs. A similar experimental installation will become operational around 1984 at the Swiss Federal Institute of Technology in Lausanne, Switzerland known under the project LOTUS. Due to the limited dimensions of the experimental cavity of the LOTUS-hybrid reactor, the LOTUS blankets have to be designed in plane geometry. Also, the bulky form of the Haefely neutron generator of the LOTUS facility obliges one to design a blanket in the plane geometry. This results in a vacuum left boundary conditions for the LOTUS blanket. The importance of a reflecting left boundary condition on the overall neutronic performance of a hybrid blanket has been analyzed in previous work in detail

  4. Development of advanced blanket performance under irradiation and system integration through JUPITER-II project

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Katsunori; Kohyama, Akira; Tanaka, Satoru; Namba, C.; Terai, T.; Kunugi, T.; Muroga, Takeo; Hasegawa, Akira; Sagara, A.; Berk, S.; Zinkle, Steven J.; Sze, Dai Kai; Petti, D. A.; Abdou, Mohamed A.; Morley, Neil B.; Kurtz, Richard J.; Snead, Lance L.; Ghoniem, Nasr M.

    2008-12-01

    This report describes an outline of the activities of the JUPITER-II collaboration (japan-USA program of Irradiation/Integration test for Fusion Research-II), Which has bee carried out through six years (2001-2006) under Phase 4 of the collabroation implemented by Amendment 4 of Annex 1 to the DOE (United States Department of Energy)-MEXT (Ministry of Education ,Culture,Sports,Science and Technology) Cooperation. This program followed the RTNS-II Program (Phase1:1982-4986), the FFTF/MOTA Program (Phase2:1987-1994) and the JUPITER Program (Phase 3: 1995-2000) [1].

  5. Advanced Modular Software Performance Monitoring

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    The LHCb software is based on the Gaudi framework, on top of which are built several large and complex software applications. The LHCb experiment is now in the active phase of collecting and analyzing data and significant performance problems arise in the Gaudi based software beginning from High Level Trigger (HLT) programs and ending with data analysis frameworks (DaVinci). It’s not easy to find hot spots in the code - only special tools can help to understand where CPU or memory usage is not reasonable. There exist many performance analyzing tools, but the main problem is that they show reports in terms of class and function names and such information usually is not very useful - the majority of algorithm developers use the Gaudi framework abstractions and usually do not know about functions which lie at the lower level. We will show a new approach which adds to performance reports a higher abstraction level based on knowledge of framework architecture and run-time object properties. A set of profiling to...

  6. Advanced modular software performance monitoring

    CERN Document Server

    Mazurov, A

    2012-01-01

    The LHCb software is based on the Gaudi framework, on top of which are built several large and complex software applications. As the LHCb experiment is now in the active phase of collecting and analyzing data, performance problems arise in various parts of the software, from the High Level Trigger (HLT) programs to data analysis frameworks. It is not easy to find hotspots in the code - only specialized tools can help to understand where CPU or memory usage are not reasonable. There exist many performance analyzing tools, but the main problem is that they show reports in terms of class and function names and such information usually is not very useful - the majority of algorithm developers use the Gaudi framework abstractions and usually do not know about functions which lie at the lower level. We will show a new approach which adds to performance reports a higher abstraction level based on knowledge of framework architecture and run-time object properties. A set of profiling tools (based on Intel VTune Amplif...

  7. Thermonuclear blankets

    International Nuclear Information System (INIS)

    Kubota, Tadashi.

    1986-01-01

    Purpose: To increase the effective thermal conductivity between lithium ceramic spheres and a metal container. Constitution: The surface of lithium ceramic spheres is coated with a nickel metal, which is further coated with a thin copper layer. Then, copper spheres with a diameter smaller than that of the lithium ceramic spheres are packed in admixture together with the lithium ceramic spheres in an appropriate volume ratio. Since, copper as a relatively soft metal is coated on the surface of the lithium ceramic spheres and the copper spheres are charged to the gaps between each of the lithium ceramic spheres, the area of contact between the lithium ceramic spheres to each other and that between the lithium ceramic spheres and the metal container are easily increased to improve the effective thermal conductivity, by which the heat removing performance of the plant can be improved. (Yoshino, Y.)

  8. Advanced Airfoils Boost Helicopter Performance

    Science.gov (United States)

    2007-01-01

    Carson Helicopters Inc. licensed the Langley RC4 series of airfoils in 1993 to develop a replacement main rotor blade for their Sikorsky S-61 helicopters. The company's fleet of S-61 helicopters has been rebuilt to include Langley's patented airfoil design, and the helicopters are now able to carry heavier loads and fly faster and farther, and the main rotor blades have twice the previous service life. In aerial firefighting, the performance-boosting airfoils have helped the U.S. Department of Agriculture's Forest Service control the spread of wildfires. In 2003, Carson Helicopters signed a contract with Ducommun AeroStructures Inc., to manufacture the composite blades for Carson Helicopters to sell

  9. Monte Carlo analysis of the effects of a blanket-shield penetration on the performance of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1977-05-01

    Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively

  10. Low cost, high yield IFE reactors: Revisiting Velikhov's vaporizing blankets

    International Nuclear Information System (INIS)

    Logan, B.G.

    1992-01-01

    The performance (efficiency and cost) of IFE reactors using MHD conversion is explored for target blanket shells of various materials vaporized and ionized by high fusion yields (5 to 500 GJ). A magnetized, prestressed reactor chamber concept is modeled together with previously developed models for the Compact Fusion Advanced Rankine II (CFARII) MHD Balance-of-Plant (BoP). Using conservative 1-D neutronics models, high fusion yields (20 to 80 GJ) are found necessary to heat Flibe, lithium, and lead-lithium blankets to MHD plasma temperatures, at initial solid thicknesses sufficient to capture most of the fusion yield. Advanced drivers/targets would need to be developed to achieve a ''Bang per Buck'' figure-of-merit approx-gt 20 to 40 joules yield per driver $ for this scheme to be competitive with these blanket materials. Alternatively, more realistic neutronics models and better materials such as lithium hydride may lower the minimum required yields substantially. The very low CFARII BoP costs (contributing only 3 mills/kWehr to CoE) allows this type of reactor, given sufficient advances that non-driver costs dominate, to ultimately produce electricity at a much lower cost than any current nuclear plant

  11. A blanket design, apparatus, and fabrication techniques for the mass production of multilayer insulation blankets for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.; Otavka, J.G.; Ruschman, M.K.; Schoo, C.J.

    1989-09-01

    The multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) consists of full cryostat length assemblies of aluminized polyester film fabricated in the form of blankets and installed as blankets to the 4.5K cold mass and the 20K and 80K thermal radiation shields. Approximately 40,000 MLI blankets will be required in the 10,000 cryogenic devices comprising the SSC accelerator. Each blanket is nearly 17 meters long and 1.8 meters wide. This paper reports the blanket design, an apparatus, and the fabrication method used to mass produce pre-fabricated MLI blankets. Incorporated in the blanket design are techniques which automate quality control during installation of the MLI blankets in the SSC cryostat. The apparatus and blanket fabrication method insure consistency in the mass produced blankets by providing positive control of the dimensional parameters which contribute to the thermal performance of the MLI blanket. By virtue of the fabrication process, the MLI blankets have inherent features of dimensional stability three-dimensional uniformity, controlled layer density, layer-to-layer registration, interlayer cleanliness, and interlayer material to accommodate thermal contraction differences. 11 refs., 6 figs., 1 tab

  12. R and D status on Water Cooled Ceramic Breeder Blanket Technology

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji

    2014-10-15

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li{sub 2}TiO{sub 3} pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

  13. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

    1993-11-01

    The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design

  14. Neutronic performance of two European breeder-inside-tube (BIT) blankets for DEMO: the helium-cooled ceramic LiAlO2 with Be multiplier and the water-cooled liquid Li17Pb

    International Nuclear Information System (INIS)

    Petrizzi, L.; Rado, V.

    1995-01-01

    In support of ENEA activity in the European Community Test Programme, neutron analysis has been performed on the two latest blanket designs: helium-cooled ceramic breeder-inside-tube (BIT) (with LiAlO 2 and Be multiplier) and water-cooled liquid Li 17 Pb in cylindrical modules (CM). The powerful MCNP Monte Carlo code was used (version 4.2). A detailed and accurate description of the geometrical model has been performed by inserting the main reactor details and avoiding breeder material dilution inside the modules. The tritium breeding ratio (TBR) performance is low for the solid breeder BIT blanket (with 10 ports 1.011) due mainly to low blanket coverage near the exhaust duct, and this solution should be revised. The CM Li 17 Pb blanket reaches a sufficient TBR (1.059, with ports) to rely on tritium self-sufficiency. Shielding properties, with respect to the toroidal field coils, have been estimated in a simplified model by means of the ANISN code, supplied with a nuclear data library consistent with that used by MCNP. The analysis suggests that a careful shield thickness/composition design should be used to ensure the shielding capability of the whole blanket plus shield system. (orig.)

  15. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    International Nuclear Information System (INIS)

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production

  16. Cognitive correlates of performance in advanced mathematics.

    Science.gov (United States)

    Wei, Wei; Yuan, Hongbo; Chen, Chuansheng; Zhou, Xinlin

    2012-03-01

    Much research has been devoted to understanding cognitive correlates of elementary mathematics performance, but little such research has been done for advanced mathematics (e.g., modern algebra, statistics, and mathematical logic). To promote mathematical knowledge among college students, it is necessary to understand what factors (including cognitive factors) are important for acquiring advanced mathematics. We recruited 80 undergraduates from four universities in Beijing. The current study investigated the associations between students' performance on a test of advanced mathematics and a battery of 17 cognitive tasks on basic numerical processing, complex numerical processing, spatial abilities, language abilities, and general cognitive processing. The results showed that spatial abilities were significantly correlated with performance in advanced mathematics after controlling for other factors. In addition, certain language abilities (i.e., comprehension of words and sentences) also made unique contributions. In contrast, basic numerical processing and computation were generally not correlated with performance in advanced mathematics. Results suggest that spatial abilities and language comprehension, but not basic numerical processing, may play an important role in advanced mathematics. These results are discussed in terms of their theoretical significance and practical implications. ©2011 The British Psychological Society.

  17. Fusion fuel blanket technology

    International Nuclear Information System (INIS)

    Hastings, I.J.; Gierszewski, P.

    1987-05-01

    The fusion blanket surrounds the burning hydrogen core of a fusion reactor. It is in this blanket that most of the energy released by the nuclear fusion of deuterium-tritium is converted into useful product, and where tritium fuel is produced to enable further operation of the reactor. As fusion research turns from present short-pulse physics experiments to long-burn engineering tests in the 1990's, energy removal and tritium production capabilities become important. This technology will involve new materials, conditions and processes with applications both to fusion and beyond. In this paper, we introduce features of proposed blanket designs and update and status of international research. In focusing on the Canadian blanket technology program, we discuss the aqueous lithium salt blanket concept, and the in-reactor tritium recovery test program

  18. High Performance Grinding and Advanced Cutting Tools

    CERN Document Server

    Jackson, Mark J

    2013-01-01

    High Performance Grinding and Advanced Cutting Tools discusses the fundamentals and advances in high performance grinding processes, and provides a complete overview of newly-developing areas in the field. Topics covered are grinding tool formulation and structure, grinding wheel design and conditioning and applications using high performance grinding wheels. Also included are heat treatment strategies for grinding tools, using grinding tools for high speed applications, laser-based and diamond dressing techniques, high-efficiency deep grinding, VIPER grinding, and new grinding wheels.

  19. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Sawan, M.E.; Cheng, E.T.

    1985-01-01

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

  20. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Sawan, M.L.; Cheng, E.T.

    1986-01-01

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li 17 Pb 83 blankets. (author)

  1. Performance Enhancements for Advanced Database Management Systems

    OpenAIRE

    Helmer, Sven

    2000-01-01

    New applications have emerged, demanding database management systems with enhanced functionality. However, high performance is a necessary precondition for the acceptance of such systems by end users. In this context we developed, implemented, and tested algorithms and index structures for improving the performance of advanced database management systems. We focused on index structures and join algorithms for set-valued attributes.

  2. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kuroda, Toshimasa; Enoeda, Mikio; Kikuchi, Shigeto [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-11-01

    The ITER breeding blanket employs a ceramic breeder and Be neutron multiplier both in small spherical pebble form. Radial-poloidal cooling panels are arranged in the blanket box to remove the nuclear heating in these materials and to reinforce the blanket structure. At the first wall, Be armor is bonded onto the stainless steel (SS) structure to provide a low Z plasma-compatible surface and to protect the first wall/blanket structure from the direct contact with the plasma during off-normal events. Thermo-mechanical analyses and investigation of fabrication procedure have been performed for this breeding blanket. To evaluate thermo-mechanical behavior of the pebble beds including the dependency of the effective thermal conductivity on stress, analysis methods have been preliminary established by the use of special calculation option of ABAQUS code, which are briefly summarized in this report. The structural response of the breeding blanket module under internal pressure of 4 MPa (in case of in-blanket LOCA) resulted in rather high stress in the blanket side (toroidal end) wall, thus addition of a stiffening rib or increase of the wall thickness will be needed. Two-dimensional elasto-plastic analyses have been performed for the Be/SS bonded interface at the first wall taking a fabrication process based on HIP bonding and thermal cycle due to pulsed plasma operation into account. The stress-strain hysteresis during these process and operation was clarified, and a procedure to assess and/or confirm the bonding integrity was also proposed. Fabrication sequence of the breeding blanket module was preliminarily developed based on the procedure to fabricate part by part and to assemble them one by one. (author)

  3. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-05-01

    Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

  4. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

    1999-07-01

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  5. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

    2017-01-15

    Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

  6. Thermomechanical analysis of the DFLL test blanket module for ITER

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

    The finite element code is used to simulate two kinds of blanket design structure, which are SLL (Quasi-Static Lithium Lead) and DLL (Dual-cooled Lithium Lead) blanket concepts for the Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM) submitted to the ITER test blanket working group. The temperature and stress distributions have been presented for the two kinds of blanket structure on the basis of the structural design, thermal-hydraulic design and neutronics analysis. Also the mechanical performance is presented for the high temperature component of blanket structure according to the ITER Structural Design Criteria (ISDC). The rationality and feasibility of the two kinds of blanket structure design of DFLL-TBM have been analyzed based on the above results which also acted as the theoretical base for further optimized analysis. (authors)

  7. Rapid thermal cycling of new technology solar array blanket coupons

    Science.gov (United States)

    Scheiman, David A.; Smith, Bryan K.; Kurland, Richard M.; Mesch, Hans G.

    1990-01-01

    NASA Lewis Research Center is conducting thermal cycle testing of a new solar array blanket technologies. These technologies include test coupons for Space Station Freedom (SSF) and the advanced photovoltaic solar array (APSA). The objective of this testing is to demonstrate the durability or operational lifetime of the solar array interconnect design and blanket technology within a low earth orbit (LEO) or geosynchronous earth orbit (GEO) thermal cycling environment. Both the SSF and the APSA array survived all rapid thermal cycling with little or no degradation in peak performance. This testing includes an equivalent of 15 years in LEO for SSF test coupons and 30 years of GEO plus ten years of LEO for the APSA test coupon. It is concluded that both the parallel gap welding of the SSF interconnects and the soldering of the APSA interconnects are adequately designed to handle the thermal stresses of space environment temperature extremes.

  8. NREL Evaluates Advanced Solar Inverter Performance for Hawaiian Electric

    Science.gov (United States)

    Companies | Energy Systems Integration Facility | NREL NREL Evaluates Advanced Solar Inverter Performance for Hawaiian Electric Companies NREL Evaluates Advanced Solar Inverter Performance for Hawaiian performance and impacts of today's advanced solar inverters, as well as proprietary feedback to the inverter

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

    International Nuclear Information System (INIS)

    Yoshikawa, K.; Ishikawa, M.; Umoto, J.; Fukuyama, A.; Mitarai, O.; Okamoto, M.; Sekimoto, H.; Nagatsu, M.

    1995-01-01

    Preliminary key issues for a synchrotron radiation-enhanced compact fusion advanced Brayton (CFAB) cycle fusion reactor similar to the CFAR (compact fusion advanced Rankine) cycle reactor are presented. These include plasma operation windows as a function of the first wall reflectivity and related issues, to estimate an allowance for deterioration of the first wall reflectivity due to dpa effects. It was found theoretically that first wall reflectivities down to 0.8 are still adequate for operation at an energy confinement scaling of 3 times Kaye-Goldston. Measurements of the graphite first wall reflectivities at Nagoya University indicate excellent reflectivities in excess of 90% for CC-312, PCC-2S, and PD-330S in the submillimeter regime, even at high temperatures in excess of 1000K. Some engineering issues inherent to the CFAB cycle are also discussed briefly in comparison with the CFAR cycle which uses hazardous limited-resource materials but is capable of using mercury as coolant for high heat removal. The CFAB cycle using helium coolant is found to achieve higher net plant conversion efficiencies in excess 60% using a non-equilibrium magnetohydrodynamic disk generator in the moderate pressure range, even at the cost of a relatively large pumping power, and at the penalty of high temperature materials, although excellent heat removal characteristics in the moderate pressure range need to be guaranteed in the future. (orig.)

  10. TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

    2011-11-30

    This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

  11. APT target-blanket fabrication development

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, D.L.

    1997-06-13

    Concepts for producing tritium in an accelerator were translated into hardware for engineering studies of tritium generation, heat transfer, and effects of proton-neutron flux on materials. Small-scale target- blanket assemblies were fabricated and material samples prepared for these performance tests. Blanket assemblies utilize composite aluminum-lead modules, the two primary materials of the blanket. Several approaches are being investigated to produce large-scale assemblies, developing fabrication and assembly methods for their commercial manufacture. Small-scale target-blanket assemblies, designed and fabricated at the Savannah River Site, were place in Los Alamos Neutron Science Center (LANSCE) for irradiation. They were subjected to neutron flux for nine months during 1996-97. Coincident with this test was the development of production methods for large- scale modules. Increasing module size presented challenges that required new methods to be developed for fabrication and assembly. After development, these methods were demonstrated by fabricating and assembling two production-scale modules.

  12. Tritium breeding blanket

    International Nuclear Information System (INIS)

    Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

    1991-01-01

    The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

  13. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  14. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

    Purpose: To produce tritium more than consumed, through thermonuclear reaction. Constitution: The energy spectrum of neutron generated by neutron multiplying reaction in a neutron multiplying blanket and moderated neutrons has a large ratio in a low energy section. In the low-energy absorption region of stainless steel which is a material of cooling pipes constituting a neutron multiplying blanket cooling channel, the neutrons are absorbed, lessening the neutron multiplying effect. To prevent this, the neutron multiplying blanket cooling channel is covered with tritium breeding blankets, thereby enabling the production of a substantially great amount of tritium more than the amount of tritium to be consumed by the thermonuclear reaction by preventing neutron absorption by the component materials of the cooling channel, improving the tritium breeding ratio by 20 to 25 %, and increasing the efficiency of use of neutrons for tritium generation. (Horiuchi, T.)

  15. Improved Dechlorinating Performance of Upflow Anaerobic Sludge Blanket Reactors by Incorporation of Dehalospirillum multivorans into Granular Sludge

    Science.gov (United States)

    Hörber, Christine; Christiansen, Nina; Arvin, Erik; Ahring, Birgitte K.

    1998-01-01

    Dechlorination of tetrachloroethene, also known as perchloroethylene (PCE), was investigated in an upflow anaerobic sludge blanket (UASB) reactor after incorporation of the strictly anaerobic, reductively dechlorinating bacterium Dehalospirillum multivorans into granular sludge. This reactor was compared to the reference 1 (R1) reactor, where the granules were autoclaved to remove all dechlorinating abilities before inoculation, and to the reference 2 (R2) reactor, containing only living granular sludge. All three reactors were fed mineral medium containing 3 to 57 μM PCE, 2 mM formate, and 0.5 mM acetate and were operated under sterile conditions. In the test reactor, an average of 93% (mole/mole) of the effluent chloroethenes was dichloroethene (DCE), compared to 99% (mole/mole) in the R1 reactor. The R2 reactor, with no inoculation, produced only trichloroethene (TCE), averaging 43% (mole/mole) of the effluent chloroethenes. No dechlorination of PCE was observed in an abiotic control consisting of sterile granules without inoculum. During continuous operation with stepwise-reduced hydraulic retention times (HRTs), both the test reactor and the R1 reactor showed conversion of PCE to DCE, even at HRTs much lower than the reciprocal maximum specific growth rate of D. multivorans, indicating that this bacterium was immobilized in the living and autoclaved granular sludge. In contrast, the R2 reactor, with no inoculation of D. multivorans, only converted PCE to TCE under the same conditions. Immobilization could be confirmed by using fluorescein-labeled antibody probes raised against D. multivorans. In granules obtained from the R1 reactor, D. multivorans grew mainly in microcolonies located in the centers of the granules, while in the test reactor, the bacterium mainly covered the surfaces of granules. PMID:9572963

  16. Blanket comparison and selection study. Final report. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  17. Blanket comparison and selection study. Final report. Volume 3

    International Nuclear Information System (INIS)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li 2 O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N 2 ) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li 2 O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  18. Blanket comparison and selection study. Final report. Volume 1

    International Nuclear Information System (INIS)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li 2 O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N 2 ) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li 2 O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  19. Blanket comparison and selection study. Final report. Volume 2

    International Nuclear Information System (INIS)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li 2 O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N 2 ) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li 2 O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  20. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1980-01-01

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  1. Modeling the performance of 'up-flow anaerobic sludge blanket' reactor based wastewater treatment plant using linear and nonlinear approaches-A case study

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Kunwar P., E-mail: kpsingh_52@yahoo.com [Environmental Chemistry Division, Indian Institute of Toxicology Research (Council of Scientific and Industrial Research), Post Box No. 80, MG Marg, Lucknow-226 002, UP (India); Basant, Nikita [School of Graduate Studies-Multiscale Modeling, Computational Simulations and Characterization in Material and Life Sciences, University of Modena and Reggio E., Modena (Italy); Malik, Amrita; Jain, Gunja [Environmental Chemistry Division, Indian Institute of Toxicology Research (Council of Scientific and Industrial Research), Post Box No. 80, MG Marg, Lucknow-226 002, UP (India)

    2010-01-18

    The paper describes linear and nonlinear modeling of the wastewater data for the performance evaluation of an up-flow anaerobic sludge blanket (UASB) reactor based wastewater treatment plant (WWTP). Partial least squares regression (PLSR), multivariate polynomial regression (MPR) and artificial neural networks (ANNs) modeling methods were applied to predict the levels of biochemical oxygen demand (BOD) and chemical oxygen demand (COD) in the UASB reactor effluents using four input variables measured weekly in the influent wastewater during the peak (morning and evening) and non-peak (noon) hours over a period of 48 weeks. The performance of the models was assessed through the root mean squared error (RMSE), relative error of prediction in percentage (REP), the bias, the standard error of prediction (SEP), the coefficient of determination (R{sup 2}), the Nash-Sutcliffe coefficient of efficiency (E{sub f}), and the accuracy factor (A{sub f}), computed from the measured and model predicted values of the dependent variables (BOD, COD) in the WWTP effluents. Goodness of the model fit to the data was also evaluated through the relationship between the residuals and the model predicted values of BOD and COD. Although, the model predicted values of BOD and COD by all the three modeling approaches (PLSR, MPR, ANN) were in good agreement with their respective measured values in the WWTP effluents, the nonlinear models (MPR, ANNs) performed relatively better than the linear ones. These models can be used as a tool for the performance evaluation of the WWTPs.

  2. Modeling the performance of 'up-flow anaerobic sludge blanket' reactor based wastewater treatment plant using linear and nonlinear approaches-A case study

    International Nuclear Information System (INIS)

    Singh, Kunwar P.; Basant, Nikita; Malik, Amrita; Jain, Gunja

    2010-01-01

    The paper describes linear and nonlinear modeling of the wastewater data for the performance evaluation of an up-flow anaerobic sludge blanket (UASB) reactor based wastewater treatment plant (WWTP). Partial least squares regression (PLSR), multivariate polynomial regression (MPR) and artificial neural networks (ANNs) modeling methods were applied to predict the levels of biochemical oxygen demand (BOD) and chemical oxygen demand (COD) in the UASB reactor effluents using four input variables measured weekly in the influent wastewater during the peak (morning and evening) and non-peak (noon) hours over a period of 48 weeks. The performance of the models was assessed through the root mean squared error (RMSE), relative error of prediction in percentage (REP), the bias, the standard error of prediction (SEP), the coefficient of determination (R 2 ), the Nash-Sutcliffe coefficient of efficiency (E f ), and the accuracy factor (A f ), computed from the measured and model predicted values of the dependent variables (BOD, COD) in the WWTP effluents. Goodness of the model fit to the data was also evaluated through the relationship between the residuals and the model predicted values of BOD and COD. Although, the model predicted values of BOD and COD by all the three modeling approaches (PLSR, MPR, ANN) were in good agreement with their respective measured values in the WWTP effluents, the nonlinear models (MPR, ANNs) performed relatively better than the linear ones. These models can be used as a tool for the performance evaluation of the WWTPs.

  3. Optimization of up-flow anaerobic sludge blanket reactor for ...

    African Journals Online (AJOL)

    Optimization of up-flow anaerobic sludge blanket reactor for treatment of composite ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... Granules grown in the bottom part of UASB reactor were more compact and tense ...

  4. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

    Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

    1985-01-01

    Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

  5. Trade-off study of liquid metal self-cooled blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of this study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. The primary results of the study are as follows: a) the lithium-lead blanket achieves a higher TBR with a smaller blanket thickness relative to the lithium blanket; b) the lithium blanket generates more energy per fusion neutron relative to the lithium-lead blanket; c) among the possible reflector materials, the carbon reflector produces the highest TBR; d) the high-Z reflector materials (Mo, Cu, W, or steel) generate more energy per fusion neutron and produce smaller TBRs relative to the carbon reflector; e) lithium-6 enrichment is required for the lithium-lead blanket to reduce the total blanket thickness; and f) the energy deposition per fusion neutron reaches a saturation as the blanket thickness, the fraction of the high-Z material in the reflector, or the reflector zone thickness increases (this allows one to design the blanket for a specific TBR without reducing the energy production)

  6. The performance enhancements of upflow anaerobic sludge blanket (UASB) reactors for domestic sludge treatment--a state-of-the-art review.

    Science.gov (United States)

    Chong, Siewhui; Sen, Tushar Kanti; Kayaalp, Ahmet; Ang, Ha Ming

    2012-07-01

    Nowadays, carbon emission and therefore carbon footprint of water utilities is an important issue. In this respect, we should consider the opportunities to reduce carbon footprint for small and large wastewater treatment plants. The use of anaerobic rather than aerobic treatment processes would achieve this aim because no aeration is required and the generation of methane can be used within the plant. High-rate anaerobic digesters receive great interests due to their high loading capacity and low sludge production. Among them, the upflow anaerobic sludge blanket (UASB) reactors have been most widely used. However, there are still unresolved issues inhibiting the widespread of this technology in developing countries or countries with climate temperature fluctuations (such as subtropical regions). A large number of studies have been carried out in order to enhance the performance of UASB reactors but there is a lack of updated documentation. In face of the existing limitations and the increasing importance of this technology, the authors present an up-to-date review on the performance enhancements of UASB reactors over the last decade. The important aspects of this article are: (i) enhancing the start-up and granulation in UASB reactors, (ii) coupling with post-treatment unit to overcome the temperature constraint, and (iii) improving the removal efficiencies of the organic matter, nutrients and pathogens in the final effluent. Finally the authors have highlighted future research direction based on their critical analysis. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

  7. Performance and model of a full-scale up-flow anaerobic sludge blanket (UASB) to treat the pharmaceutical wastewater containing 6-APA and amoxicillin.

    Science.gov (United States)

    Chen, Zhiqiang; Wang, Hongcheng; Chen, Zhaobo; Ren, Nanqi; Wang, Aijie; Shi, Yue; Li, Xiaoming

    2011-01-30

    A full-scale test was conducted with an up-flow anaerobic sludge blanket (UASB) pre-treating pharmaceutical wastewater containing 6-aminopenicillanic acid (6-APA) and amoxicillin. The aim of the study is to investigate the performance of UASB in the condition of a high chemical oxygen demand (COD) loading rate from 12.57 to 21.02 kgm(-3)d(-1) and a wide pH from 5.57 to 8.26, in order to provide a reference for treating the similar chemical synthetic pharmaceutical wastewater containing 6-APA and amoxicillin. The results demonstrated that the UASB average percentage reduction in COD, 6-APA and amoxicillin were 52.2%, 26.3% and 21.6%, respectively. In addition, three models, built on the back propagation neural network (BPNN) theory and linear regression techniques were developed for the simulation of the UASB system performance in the biodegradation of pharmaceutical wastewater containing 6-APA and amoxicillin. The average error of COD, 6-APA and amoxicillin were -0.63%, 2.19% and 5.40%, respectively. The results indicated that these models built on the BPNN theory were well-fitted to the detected data, and were able to simulate and predict the removal of COD, 6-APA and amoxicillin by UASB. Crown Copyright © 2010. Published by Elsevier B.V. All rights reserved.

  8. Performance appraisal and advancement exercise 2005

    CERN Multimedia

    Human Resources Department

    2004-01-01

    The 2005 performance appraisal and advancement exercise will start in the usual way with annual interviews between staff and their supervisors. These interviews should be held in the period from 15 November 2004 to 15 February 2005 [see Administrative Circular No 26 (Rev. 5) Annex II, § 3. For this exercise, the paper MAPS form has been replaced by an electronic form in the EDH system, which is available via EDH (on the EDH desktop under Other Tasks / HR & Training) No changes have been made to the contents of the form. HR Department will shortly provide further information on this subject. Human Resources Department Tel. 73566

  9. Materials performance in advanced fossil technologies

    International Nuclear Information System (INIS)

    Natesan, K.

    1991-01-01

    A number of advanced technologies are being developed to convert coal into clean fuels for use as a feedstock in chemical plants and for power generation. From the standpoint of component materials, the environments created by coal conversion and combustion in these technologies and their interactions with materials are of interest. This article identifies several modes of materials degradation and possible mechanisms for metal wastage. Available data on the performance of materials in several of the environments are highlighted, and examples of promising research activities to improve the corrosion resistance of materials are presented

  10. Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Yoshinori, E-mail: kawamura.yoshinori@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Sato, Satoshi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Konno, Chikara; Edao, Yuki; Hayashi, Takumi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Yamanishi, Toshihiko [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Thermo-hydraulic calculation in the TBM at the water ingress event has been done. • Shielding calculations for the ITER equatorial port #18 were conducted by using C-lite model. • Prototypic pebbles of Be{sub 17}Ti{sub 2} and Be{sub 12}V had a good oxidation property similar to Be{sub 12}Ti pebble. • Li rich Li{sub 2}TiO{sub 3} pebbles were successfully fabricated using the emulsion method by controlling sintering atmosphere. • New tritium production/recovery experiments at FNS have been started by using ionization chamber as on-line gas monitor. - Abstract: The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

  11. The scaling of economic and performance parameters of DT and advanced fuel fusion reactors

    International Nuclear Information System (INIS)

    Roth, J.R.

    1983-01-01

    In this study, the plasma stability index beta and the fusion power density in the plasma were treated as independent variables to determine how they influenced three economic performance parameters of fusion reactors burning the DT and four advanced fusion fuel cycles. The economic/performance parameters included the total power produced per unit length of reactor; the mass per unit length, and the specific mass in kilograms/kilowatt. The scaling of these parameters with beta and fusion power density was examined for a common set of engineering assumptions on the allowable wall loading limits, the maximum magnetic field existing in the plasma, average blanket mass density, etc. It was found that the power per unit length decreased as the plasma power density and beta increased. This is a consequence of the fact that the first wall is a bottleneck in the energy flow from the plasma to the generating equipment, and the wall power flux will exceed wall loading limits if the plasma radius exceeds a critical value. If one wished to build an engineering test reactor which produced a burning plasma at the lowest possible initial cost, and without regard to whether such a reactor would ultimately produce the cheapest power, then one would minimize the mass per unit length. The mass per unit length decreases with increasing plasma power density and beta, with the DT reaction being the most expensive at a fixed plasma power density (because of its thicker blanket), and the least expensive at a fixed value of beta, at least up to values of beta of 50%. The specific mass, in kg/kw, which is a rough measure of the cost of the power generated by the reactor, shows an opposite trend. It increases with increasing plasma power density and beta. At a given plasma power density and low beta, the DT reaction gives the lowest specific mass, but at a fixed beta above 10%, the advanced fuel cycles have the lowest specific mass

  12. Novel blanket design for ICTR's

    International Nuclear Information System (INIS)

    Abdel-Khalik, S.I.; Conn, R.W.; Wolfer, W.G.; Larsen, E.N.; Sviatoslavsky, I.N.

    1978-01-01

    A novel blanket design for ICTRs is described. This blanket is used in SOLASE, the conceptual laser fusion reactor of the University of Wisconsin. The blanket to be described offers numerous advantages, including low cost, low weight, low induced radioactivity levels, the potential for hands-on maintenance, modular construction, low pressure, ability to decouple first wall and blanket coolant temperatures, adequate breeding, low tritium inventory and leakage, and sufficiently long life

  13. Minimum thickness blanket-shield for fusion reactors

    International Nuclear Information System (INIS)

    Karni, Y.; Greenspan, E.

    1989-01-01

    A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

  14. Performance appraisal and advancement exercise 2006

    CERN Multimedia

    Human Resources Department

    2005-01-01

    The 2006 performance appraisal and advancement exercise will start in the usual way with annual interviews between staff and their supervisors. This year, these interviews should be held in the period from 1 December 2005 to 31 March 2006. In this connection, a number of improvements to the procedures relating to the interviews and performance appraisal are currently under study. Administrative Circular No 26 (Procedures governing the Career Development of Staff Members) and the electronic MAPS form in EDH are being reviewed and will be available from December onwards. In the meantime supervisors can start the interview procedure. HR Department will shortly provide further information on this subject. The 2005 MAPS report can be retrieved for consultation at any time via EDH. Human Resources Department Tel. 73566

  15. Performance appraisal and advancement exercise 2006 (Rev)

    CERN Multimedia

    Human Resources Department

    2005-01-01

    The 2006 performance appraisal and advancement exercise will start in the usual way with annual interviews between staff and their supervisors. This year, these interviews should be held in the period from 1 December 2005 to 15 March 2006*. In this connection, a number of improvements to the procedures relating to the interviews and performance appraisals are currently under study. Administrative Circular No. 26 (Procedures Governing the Career Development of Staff Members) and the electronic MAPS form in EDH are being reviewed and will be available from December onwards. In the meantime supervisors can start the interview procedure. HR Department will shortly provide further information on this subject. The 2005 MAPS report can be retrieved for consultation at any time via EDH. * Instead of 31 March, as indicated in the electronic Bulletin No. 47 of 21 November 2005 Human Resources Department Tel. 73566

  16. Advisory group meeting on design and performance of reactor and subcritical blanket systems with lead and lead-bismuth as coolant and/or target material. Working material

    International Nuclear Information System (INIS)

    2001-01-01

    The purpose of the IAEA Advisory Group Meeting (AGM) on Design and Performance of Reactor and Sub-critical Blanket Systems with Lead and Lead-Bismuth as Coolant and/or Target Material was to provide a forum for international information exchange on all the topics relevant to Pb and Pb/Bi cooled critical and sub-critical reactors. In addition, the AGM aimed at: (1) finding ways and means to improve international co-ordination efforts in this area; (2) obtaining advice from the Member States with regard to the activities to be implemented in this area by the IAEA, in order to best meet their needs; and (3) laying out the plans for an effective co-ordination and support of the R and D activities in this area. The AGM stressed that nuclear energy is a realistic solution to satisfy the energy demand, considering the limited resources of fossil fuel, its uneven distribution in the world and the impact of its use on the planet, and taking into account the expected doubling of the world population in the 21st century and tripling of the electricity demand (especially in the developing countries). However, the AGM concluded that the development of an innovative nuclear technology meeting the following requirements must be pursued: (a) deterministic exclusion of any severe accident; (b) proliferation resistance; (c) cost competitiveness with alternative energy sources; (d) sustainable fuel supply; and (e) solution of the radioactive waste management problem

  17. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

  18. Tritium transport in the water cooled Pb-17Li blanket concept of DEMO

    International Nuclear Information System (INIS)

    Reiter, F.; Tominetti, S.; Perujo, A.

    1992-01-01

    The code TIRP has been used to calculate the time dependence of tritium inventory and tritium permeation into the coolant and into the first wall boxes in the water cooled Pb-17Li blanket concept of DEMO. The calculations have been performed for the martensitic steel MANET and the austenitic steel AISI 316L as blanket structure materials, for water or helium cooling and for convective or no motion of the liquid breeder in the blanket. Tritium inventories are rather low in blankets with MANET structure and higher in those with AISI 316L structure. Tritium permeation rates are too high in both blankets. Further calculations on tritium inventory and permeation are therefore presented for blankets with TiC permeation barriers of 1 μm thickness on various surfaces of the blanket structure and for blankets with any permeation barriers in function of their thickness, tritium diffusivities, tritium surface recombination rates and atomic densities. These last calculations have been performed for a blanket with coatings on the outer surfaces of the blanket and with a tritium residence time of 10 4 s and for a blanket with coatings on both sides of the cooling tubes and stagnant Pb-17Li in the blanket. The second case for a blanket with MANET structure presents a very interesting solution for tritium recovery by permeation into and pumping from the first wall boxes. (orig.)

  19. Test Blanket Working Group's recent activities

    International Nuclear Information System (INIS)

    Vetter, J.E.

    2001-01-01

    The ITER Test Blanket Working Group (TBWG) has continued its activities during the period of extension of the EDA with a revised charter on the co-ordination of the development work performed by the Parties and by the JCT leading to a co-ordinated test programme on ITER for a DEMO-relevant tritium breeding blanket. This follows earlier work carried out until July 1998, which formed part of the ITER Final Design Report (FDR), completed in 1998. Whilst the machine parameters for ITER-FEAT have been significantly revised compared to the FDR, testing of breeding blanket modules remains a main objective of the test programme and the development of a reactor-relevant breeding blanket to ensure tritium fuel self-sufficiency is recognized a key issue for fusion. Design work and R and D on breeding blanket concepts, including co-operation with the other Contacting Parties of the ITER-EDA for testing these concepts in ITER, are included in the work plans of the Parties

  20. Integrating advanced facades into high performance buildings

    International Nuclear Information System (INIS)

    Selkowitz, Stephen E.

    2001-01-01

    Glass is a remarkable material but its functionality is significantly enhanced when it is processed or altered to provide added intrinsic capabilities. The overall performance of glass elements in a building can be further enhanced when they are designed to be part of a complete facade system. Finally the facade system delivers the greatest performance to the building owner and occupants when it becomes an essential element of a fully integrated building design. This presentation examines the growing interest in incorporating advanced glazing elements into more comprehensive facade and building systems in a manner that increases comfort, productivity and amenity for occupants, reduces operating costs for building owners, and contributes to improving the health of the planet by reducing overall energy use and negative environmental impacts. We explore the role of glazing systems in dynamic and responsive facades that provide the following functionality: Enhanced sun protection and cooling load control while improving thermal comfort and providing most of the light needed with daylighting; Enhanced air quality and reduced cooling loads using natural ventilation schemes employing the facade as an active air control element; Reduced operating costs by minimizing lighting, cooling and heating energy use by optimizing the daylighting-thermal tradeoffs; Net positive contributions to the energy balance of the building using integrated photovoltaic systems; Improved indoor environments leading to enhanced occupant health, comfort and performance. In addressing these issues facade system solutions must, of course, respect the constraints of latitude, location, solar orientation, acoustics, earthquake and fire safety, etc. Since climate and occupant needs are dynamic variables, in a high performance building the facade solution have the capacity to respond and adapt to these variable exterior conditions and to changing occupant needs. This responsive performance capability

  1. Performance characteristics for advanced control systems

    International Nuclear Information System (INIS)

    Kisner, R.A.

    1989-01-01

    A growing collection of control techniques is becoming available to the design engineer. This make selection of the most appropriate technique for a given application a difficult task. A systematic approach to evaluating alternative control schemes is needed. The approach discussed in this paper expands the traditional concepts of quantitative performance analysis to include other relevant factors such as robustness of the technique, resource requirements, and effects on operators and other personnel. This collection of factors, termed measures of utility, may be used as qualitative and quantitative means of evaluating and comparing properties of alternative control system designs. This paper, although not an in-depth study, serves to outline several measures of utility and suggests a general structure for control system development. This method of comparing the usefulness of alternative control system will prove valuable to the ORNL Advanced Controls Program (ACTO) for optimizing compatibility with actual systems and equipment

  2. Neutron dosimetry for the TFTR Lithium-Blanket-Module program

    International Nuclear Information System (INIS)

    Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

    1981-01-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory

  3. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

    Kakudate, S.; Nakahira, M.; Oka, K.; Taguchi, K.; Obara, K.; Tada, E.; Shibanuma, K.; Tesini, A.; Haange, R.; Maisonnier, D.

    2001-01-01

    In ITER, the in-vessel components such as blanket are to be maintained or replaced remotely since they will be activated by 14 MeV neutrons, and a complete exchange of shielding blanket with breeding blanket is foreseen after the Basic Performance Phase. The blanket is segmented into about seven hundred modules to facilitate remote maintainability and allow individual module replacement. For this, the remote handing equipment for blanket maintenance is required to handle a module with a dead weight of about 4 tonne within a positioning accuracy of a few mm under intense gamma radiation. According to the ITER R and D program, a rail-mounted vehicle manipulator system was developed and the basic feasibility of this system was verified through prototype testing. Following this, development of full-scale remote handling equipment has been conducted as one of the ITER Seven R and D Projects aiming at a remote handling demonstration of the ITER blanket. As a result, the Blanket Test Platform (BTP) composed of the full-scale remote handling equipment has been completed and the first integrated performance test in March 1998 has shown that the fabricate remote handling equipment satisfies the main requirements of ITER blanket maintenance. (author)

  4. High performance anode for advanced Li batteries

    Energy Technology Data Exchange (ETDEWEB)

    Lake, Carla [Applied Sciences, Inc., Cedarville, OH (United States)

    2015-11-02

    The overall objective of this Phase I SBIR effort was to advance the manufacturing technology for ASI’s Si-CNF high-performance anode by creating a framework for large volume production and utilization of low-cost Si-coated carbon nanofibers (Si-CNF) for the battery industry. This project explores the use of nano-structured silicon which is deposited on a nano-scale carbon filament to achieve the benefits of high cycle life and high charge capacity without the consequent fading of, or failure in the capacity resulting from stress-induced fracturing of the Si particles and de-coupling from the electrode. ASI’s patented coating process distinguishes itself from others, in that it is highly reproducible, readily scalable and results in a Si-CNF composite structure containing 25-30% silicon, with a compositionally graded interface at the Si-CNF interface that significantly improve cycling stability and enhances adhesion of silicon to the carbon fiber support. In Phase I, the team demonstrated the production of the Si-CNF anode material can successfully be transitioned from a static bench-scale reactor into a fluidized bed reactor. In addition, ASI made significant progress in the development of low cost, quick testing methods which can be performed on silicon coated CNFs as a means of quality control. To date, weight change, density, and cycling performance were the key metrics used to validate the high performance anode material. Under this effort, ASI made strides to establish a quality control protocol for the large volume production of Si-CNFs and has identified several key technical thrusts for future work. Using the results of this Phase I effort as a foundation, ASI has defined a path forward to commercialize and deliver high volume and low-cost production of SI-CNF material for anodes in Li-ion batteries.

  5. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

    The primary function of beryllium in a fusion reactor blanket is neutron multiplication to enhance tritium breeding. However, because heat, tritium and helium will be generated in and/or transported through beryllium and because the beryllium is in contact with other blanket materials, the thermal, mechanical, tritium/helium and compatibility properties of beryllium are important in blanket design. In particular, tritium retention during normal operation and release during overheating events are safety concerns. Accommodating beryllium thermal expansion and helium-induced swelling are important issues in ensuring adequate lifetime of the structural components adjacent to the beryllium. Likewise, chemical/metallurgical interactions between beryllium and structural components need to be considered in lifetime analysis. Under accident conditions the chemical interaction between beryllium and coolant and breeding materials may also become important. The performance of beryllium in fusion blanket applications depends on fabrication variables and operational parameters. First the properties database is reviewed to determine the state of knowledge of beryllium performance as a function of these variables. Several design calculations are then performed to indicate ranges of fabrication and operation variables that lead to optimum beryllium performance. Finally, areas for database expansion and improvement are highlighted based on the properties survey and the design sensitivity studies

  6. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

    Cheng, E.T.; Engholm, B.A.; Su, S.D.

    1981-01-01

    The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design

  7. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2005-03-01

    Japan Atomic Energy Research Institute (JAERI) has been performing the development of breeding blanket for fusion power plant, as a leading institute of the development of solid breeder blankets, according to the long-term R and D program of the blanket development established by the Fusion Council of Japan in 1999. This report is an overview of development plan, achievements of element technology development and future prospect and plan of the development of the solid breeding blanket in JAERI. In this report, the mission of the blanket development activity in JAERI, key issues and roadmap of the blanket development have been clarified. Then, achievements of the element technology development were summarized and showed that the development has progressed to enter the engineering testing phase. The specific development target and plan were clarified with bright prospect. Realization of the engineering test phase R and D and completion of ITER test blanket module testing program, with universities/NIFS cooperation, are most important steps in the development of breeding blanket of fusion power demonstration plant. (author)

  8. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    International Nuclear Information System (INIS)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m 2 and a surface heat flux of 1 MW/m 2 . The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO 2 rods. The helium coolant pressure is 5 MPa, entering the module at 297 0 C and exiting at 550 0 C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter

  9. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-12-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations

  10. Neutronics analysis for aqueous self-cooled fusion reactor blankets

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Jaffa, R.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1986-06-01

    The tritium breeding performance of several Aqueous Self-Cooled Blanket (ASCB) configurations for fusion reactors has been evaluated. The ASCB concept employs small amounts of lithium compound dissolved in light or heavy water to serve as both coolant and breeding medium. The inherent simplicity of this concept allows the development of blankets with minimal technological risk. The tritium breeding performance of the ASCB concept is a critical issue for this family of blankets. Contrary to conventional blanket designs there will be a significant contribution to the tritium breeding ratio (TBR) in the water coolant/breeder of duct shields, and the 3-D TBR will therefore be similar to the 1-D TBR. The tritium breeding performance of an ASCB for a MARS-like-tandem reactor and an ASCB based breeding-shield for the Next European Torus (NET) are assessed. Two design options for the MARS-like blanket are discussed. One design employs a vanadium first wall, and zircaloy for the structural material. The trade-offs between light water and heavy water cooling options for this zircaloy blanket are discussed. The second design option for MARS relies on the use of a vanadium alloy as the stuctural material, and heavy water as the coolant. It is demonstrated that both design options lead to low-activation blankets that allow class C burial. The breeder-shield for NET consists of a water-cooled stainless steel shield

  11. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Naruse, Yuji; Yamaoka, Mitsuaki; Ohara, Atsushi; Ono, Kiyoshi; Kobayashi, Shigetada.

    1992-06-01

    Aqueous solution blanket using lithium salts such as LiNO 3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

  12. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  13. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    International Nuclear Information System (INIS)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li 2 O) and lithium zirconate (Li 2 ZrO 3 ) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers

  14. Flow balancing in liquid metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Morley, N.B.

    1995-01-01

    Non-uniform flow distribution between parallel channels is one of the most serious concerns for self-cooled liquid metal blankets with electrically insulated walls. We show that uncertainties in flow distribution can be dramatically reduced by relatively simple design modifications. Several design features which impose flow uniformity by electrically coupling parallel channels are surveyed. Basic mechanisms for ''flow balancing'' are described, and a particular self-regulating concept using discrete passive electrodes is proposed for the US ITER advanced blanket concept. Scoping calculations suggest that this simple technique can be very powerful in equalizing the flow, even with massive insulator failures in individual channels. More detailed analyses and experimental verification will be required to demonstrate this concept for ITER. (orig.)

  15. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

    1993-01-01

    Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

  16. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    Ozawa, Yoshihiro; Uda, Tatsuhiko; Maki, Koichi.

    1993-01-01

    The present invention provides a blanket of a thermonuclear device which produces tritium fuels consumed in plasmas while converting neutrons generated in the plasmas into heat energy. That is, zirconium is coated to at least one of neutron breeder pebbles and breeder pebbles, to suppress reaction between them by being in direct contact with each other at a high temperature. Further, fins are attached to a cooling pipe at a pitch smaller than the diameter of both of the pebbles, to prevent direct contact at whole surface of the pebbles and the cooling pipe, which would lower a temperature excessively. The length of the fin is controlled to control the thickness of a helium gas gap. With such constitution, direct contact of neutron breeder pebbles and the breeder pebble which are to be filled and mixed, and tend to react at a high temperature, can be prevented. The temperature of the breeding blanket is reliably prevented from lowering below a tritium emitting temperature. The structure is simplified and the production is facilitated. (I.S.)

  17. Preliminary analyses of neutronics schemes for three kinds waste transmutation blankets of fusion-fission hybrid

    International Nuclear Information System (INIS)

    Zhang Mingchun; Feng Kaiming; Li Zaixin; Zhao Fengchao

    2012-01-01

    The neutronics schemes of the helium-cooled waste transmutation blanket, sodium-cooled waste transmutation blanket and FLiBe-cooled waste transmutation blanket were preliminarily calculated and analysed by using the spheroidal tokamak (ST) plasma configuration. The neutronics properties of these blankets' were compared and analyzed. The results show that for the transmutation of "2"3"7Np, FLiBe-cooled waste transmutation blanket has the most superior transmutation performance. The calculation results of the helium-cooled waste transmutation blanket show that this transmutation blanket can run on a steady effective multiplication factor (k_e_f_f), steady power (P), and steady tritium production rate (TBR) state for a long operating time (9.62 years) by change "2"3"7Np's initial loading rate of the minor actinides (MA). (authors)

  18. Feasibility study of fusion breeding blanket concept employing graphite reflector

    International Nuclear Information System (INIS)

    Cho, Seungyon; Ahn, Mu-Young; Lee, Cheol Woo; Kim, Eung Seon; Park, Yi-Hyun; Lee, Youngmin; Lee, Dong Won

    2015-01-01

    Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

  19. Feasibility study of fusion breeding blanket concept employing graphite reflector

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seungyon, E-mail: sycho@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Woo; Kim, Eung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

  20. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

    Enoeda, M.; Sato, S.; Hatano, T.

    2001-01-01

    Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

  1. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

    Le Marois, G.; Federzoni, L.; Bucci, P.; Revirand, P.

    2000-01-01

    The benefit of HIP techniques applied to the fabrication of fusion internal components for higher performances, reliability and cost savings are emphasized. To demonstrate the potential of the techniques, design of new blankets concepts and mock-ups fabrication are currently performed by CEA. A coiled tube concept that allows cooling arrangement flexibility, strong reduction of the machining and number of welds is proposed for ITER IAM. Medium size mock-ups according to the WCLL breeding blanket concept have been manufactured. The fabrication of a large size mock-up is under progress. These activities are supported by numerical calculations to predict the deformations of the parts during HIP'ing. Finally, several HIP techniques issues have been identified and are discussed

  2. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

  3. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    1983-10-01

    This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies

  4. Performance of the Advanced Photon Source

    International Nuclear Information System (INIS)

    Decker, G.

    1997-01-01

    The Advanced Photon Source (APS) positron storage ring is a 100-mA, 7-GeV, third-generation x-ray synchrotron radiation source which began operation in March 1995. Since that time, significant progress on beamline construction and commissioning has taken place, with many of the x-ray user beamlines in operation. Operational design goals which have been met or exceeded include 10-hour lifetime, > 90% availability, > 100-mA average current, > 5-mA single-bunch current, < 10% uncorrected coupling, 8-mm full vertical apertures for insertion devices, and ultra-stable orbit (<4.5 microm rms vertically, 17 microm horizontally). Progress beyond these design goals and a report on development plans, including top-up operation (injection with x-ray beamline shutters open), are presented

  5. Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Malang, S.; Sawan, M.

    2005-01-01

    As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li 2 BeF 4 and the low melting point molten salts such as LiBeF 3 and LiNaBeF 4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC f /SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R and D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan

  6. Imploding-liner reactor nucleonic studies: the LINUS blanket

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1977-09-01

    Scoping nucleonic studies have been performed for a small imploding-liner fusion reactor concept. Tritium breeding ratio and time-dependent energy deposition rates were the primary parameters of interest in the study. Alloys of Pb and LiPb were considered for the liquid liner (blanket), and tritium breeding was found to be more than adequate with blankets less than 1 m thick. However, neutron leakages into the solid cylinder block surrounding the liquid liner are generally quite high, so considerable effort was concentrated on minimizing these values. Time-dependent calculations reveal that 89% of the energy is deposited in the blanket within 2 μs. Thus, LINUS's blanket should remain intact for the requisite neutron and gamma-ray lifetimes

  7. Application of vanadium alloys to a fusion reactor blanket

    Energy Technology Data Exchange (ETDEWEB)

    Bethin, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center)

    1984-05-01

    Vanadium and vanadium alloys are of interest in fusion reactor blanket applications due to their low induced radioactivity and outstanding elevated temperature mechanical properties during neutron irradiation. The major limitation to the use of vanadium is its sensitivity to oxygen impurities in the blanket environment, leading to oxygen embrittlement. A quantitative analysis was performed of the interaction of gaseous impurities in a helium coolant with vanadium and the V-15Cr-5Ti alloy under conditions expected in a fusion reactor blanket. It was shown that the use of unalloyed V would impose severe restrictions on the helium gas cleanup system due to excessive oxygen buildup and embrittlement of the metal. However, internal oxidation effects and the possibly lower terminal oxygen solubility in the alloy would impose much less severe cleanup constraints. It is suggested that V-15Cr-5Ti is a promising candidate for certain blanket applications and deserves further consideration.

  8. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  9. Li2O-pebble type tritium breeding blanket for fusion experimental reactor, 1

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

    1984-01-01

    The fusion experimental reactor is the next stage device in Japan, which is planned to be constructed following the critical plasma experimental device JT-60 being constructed at present. The breeding blanket installed in nuclear fusion reactors is one of most important structures, and it is required to satisfy the fundamental performance of producing and continuously recovering tritium as the nuclear fusion fuel, and other requirement in good coordination. The Li 2 O pebble type breeding blanket that Kawasaki Heavy Industries Ltd. has examined is the concept for resolving the problems of the mass transfer and thermal stress cracking of Li 2 O, which are important in blanket design. In this paper, the concept and characteristics of this breeding blanket are discussed from the viewpoint of the breeding and continuous recovery of tritium, the ease of manufacture and the maintenance of soundness. The breeding blanket is composed of breeding region, tritium purge region, cooling region, plasma stabilizing conductors and blanket container. Li 2 O is excellent in its tritium breeding performance and heat conductivity. The functions required for the breeding blanket, the fundamental structure, the examples of breeding blanket concept, the selection of breeding blanket concept, the characteristics of Li 2 O pebble type blanket and its future prospect are described. (Kako, I.)

  10. Advanced transport systems analysis, modeling, and evaluation of performances

    CERN Document Server

    Janić, Milan

    2014-01-01

    This book provides a systematic analysis, modeling and evaluation of the performance of advanced transport systems. It offers an innovative approach by presenting a multidimensional examination of the performance of advanced transport systems and transport modes, useful for both theoretical and practical purposes. Advanced transport systems for the twenty-first century are characterized by the superiority of one or several of their infrastructural, technical/technological, operational, economic, environmental, social, and policy performances as compared to their conventional counterparts. The advanced transport systems considered include: Bus Rapid Transit (BRT) and Personal Rapid Transit (PRT) systems in urban area(s), electric and fuel cell passenger cars, high speed tilting trains, High Speed Rail (HSR), Trans Rapid Maglev (TRM), Evacuated Tube Transport system (ETT), advanced commercial subsonic and Supersonic Transport Aircraft (STA), conventionally- and Liquid Hydrogen (LH2)-fuelled commercial air trans...

  11. Measurement of leakage neutron spectra from advanced blanket materials and structural materials induced by D-T neutrons. Correction for energy loss of charged particle in sample materials

    International Nuclear Information System (INIS)

    Nishio, Takashi; Kondo, Tetsuo; Takagi, Hiroyuki; Murata, Isao; Takahashi, Akito; Kokooo; Maekawa, Fujio; Ikeda, Yujiro; Takeuchi, Hiroshi

    2000-01-01

    D-T neutron benchmark experiments for LiAlO 2 , Li 2 TiO 3 , Li 2 ZrO 3 , Cu and W have been conducted at FNS of JAERI to validate five nuclear data files. The former three are promising advanced breeder materials and the latter two are important structural materials in a fusion reactor. From the results, all the nuclear data files were confirmed to be fairly reliable with respect to the prediction of neutron spectrum in the use of Li 2 TiO 3 and Cu. For LiAlO 2 and W, some large discrepancies between the experimental and calculated data were observed. For Li 2 ZrO 3 , the C/E values became very large for all the nuclear data files. (author)

  12. Advances in Nb3Sn Performance

    International Nuclear Information System (INIS)

    Godeke, Arno

    2008-01-01

    Nb 3 Sn wires with non-Cu critical current densities (J c ) that surpass 3 kAmm -2 at 12 T and 4.2 K are commercially available in piece lengths longer than 10 km. Accelerator-type magnets that utilize these conductors have achieved record magnetic fields. This article summarizes key developments in the last decade that have led to these significant improvements in the performance of Nb 3 Sn wires.

  13. On the safety performance of the advanced nuclear energy systems

    International Nuclear Information System (INIS)

    Li Shounan

    1999-01-01

    Some features on the safety performances of the Advanced Nuclear Energy Systems are discussed. The advantages and some peculiar problems on the safety of Advanced Nuclear Energy Systems with subcritical nuclear reactor driven by external neutron sources are also pointed out in comparison with conventional nuclear reactors

  14. Analysis of ER string test thermally instrumented interconnect 80-K MLI blanket

    International Nuclear Information System (INIS)

    Daly, E.; Pletzer, R.

    1992-04-01

    An 80-K Multi Layer Insulation (MLI) blanket in the interconnect region between magnets DD0019 and DD0027 in the Fermi National Accelerator Laboratory (FNAL) ER string was instrumented with temperature sensors to obtain the steady-state temperature gradient through the blanket after string cooldown. A thermal model of the 80-K blanket assembly was constructed to analyze the steady-state temperature gradient data. Estimates of the heat flux through the 80-K MLI blanket assembly and predicted temperature gradients were calculated. The thermal behavior of the heavy polyethylene terapthalate (PET) cover layers separating the shield and inner blanket and inner and outer blankets was derived empirically from the data. The results of the analysis predict a heat flux of 0.363--0.453 W/m 2 based on the 11 sets of data. These flux values are 33--46% below the 80-K MLI blanket heat leak budget of 0.676 W/m 2 . The effective thermal resistance of the two heavy PET cover layers between the shield and inner blanket was found to be 2.1 times that of a single PET spacer layer, and the effective resistance of the two heavy PET cover layers between the inner blanket and outer blanket was found to be 7 times that of a single PET spacer layer. Based on these results, the 80-K MLI blanket assembly appears to be performing more than adequately to meet the 80-K static IR heat leak budget. However, these results should not be construed as a verification of the 80-K static IR heat leak, since no actual heat leak was measured. The results have been used to improve the empirically based model data in the 80-K MLI blanket thermal model, which has previously not included the effects of heavy PET cover layers on 80-K MLI blanket thermal performance

  15. Thermal Model Predictions of Advanced Stirling Radioisotope Generator Performance

    Science.gov (United States)

    Wang, Xiao-Yen J.; Fabanich, William Anthony; Schmitz, Paul C.

    2014-01-01

    This presentation describes the capabilities of three-dimensional thermal power model of advanced stirling radioisotope generator (ASRG). The performance of the ASRG is presented for different scenario, such as Venus flyby with or without the auxiliary cooling system.

  16. Design, Manufacture, and Experimental Serviceability Validation of ITER Blanket Components

    Science.gov (United States)

    Leshukov, A. Yu.; Strebkov, Yu. S.; Sviridenko, M. N.; Safronov, V. M.; Putrik, A. B.

    2017-12-01

    In 2014, the Russian Federation and the ITER International Organization signed two Procurement Arrangements (PAs) for ITER blanket components: 1.6.P1ARF.01 "Blanket First Wall" of February 14, 2014, and 1.6.P3.RF.01 "Blanket Module Connections" of December 19, 2014. The first PA stipulates development, manufacture, testing, and delivery to the ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels intended for withstanding the heat flux from the plasma up to 4.7MW/m2. Two Russian institutions, NIIEFA (Efremov Institute) and NIKIET, are responsible for the implementation of this PA. NIIEFA manufactures plasma-facing components (PFCs) of the EHF FW panels and performs the final assembly and testing of the panels, and NIKIET manufactures FW beam structures, load-bearing structures of PFCs, and all elements of the panel attachment system. As for the second PA, NIKIET is the sole official supplier of flexible blanket supports, electrical insulation key pads (EIKPs), and blanket module/vacuum vessel electrical connectors. Joint activities of NIKIET and NIIEFA for implementing PA 1.6.P1ARF.01 are briefly described, and information on implementation of PA 1.6.P3.RF.01 is given. Results of the engineering design and research efforts in the scope of the above PAs in 2015-2016 are reported, and results of developing the technology for manufacturing ITER blanket components are presented.

  17. Stress analysis of blanket vessel for JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

    A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

  18. Nuclear Analysis of an ITER Blanket Module

    Science.gov (United States)

    Chiovaro, P.; Di Maio, P. A.; Parrinello, V.

    2013-08-01

    ITER blanket system is the reactor's plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.

  19. Structural materials for fusion reactor blanket systems

    International Nuclear Information System (INIS)

    Bloom, E.E.; Smith, D.L.

    1984-01-01

    Consideration of the required functions of the blanket and the general chemical, mechanical, and physical properties of candidate tritium breeding materials, coolants, structural materials, etc., leads to acceptable or compatible combinations of materials. The presently favored candidate structural materials are the austenitic stainless steels, martensitic steels, and vanadium alloys. The characteristics of these alloy systems which limit their application and potential performance as well as approaches to alloy development aimed at improving performance (temperature capability and lifetime) will be described. Progress towards understanding and improving the performance of structural materials has been substantial. It is possible to develop materials with acceptable properties for fusion applications

  20. Composite beryllium-ceramics breeder pin elements for a gas cooled solid blanket

    International Nuclear Information System (INIS)

    Carre, F.; Chevreau, G.; Gervaise, F.; Proust, E.

    1986-06-01

    Helium coolant have main advantages compared to water for solid blankets. But limitations exist too and the development of attractive helium cooled blankets based on breeder pin assemblies has been essentially made possible by the derivation from recent CEA neutronic studies of an optimized composite beryllium/ceramics breeder arrangement. Description of the proposed toroidal blanket layout for Net is made together with the analysis of its main performance. Merits of the considered composite Be/ceramics breeder elements are discussed

  1. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  2. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    Smith, D.L.; Sze, D.K.

    1986-02-01

    The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

  3. Reducing beryllium content in mixed bed solid-type breeder blankets

    Energy Technology Data Exchange (ETDEWEB)

    Shimwell, J., E-mail: mail@jshimwell.com [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Lilley, S.; Morgan, L.; Packer, L.; Kovari, M.; Zheng, S. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); McMillan, J. [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom)

    2016-11-01

    Highlights: • The ratio of breeder ceramic to neutron multiplier of breeder blankets was varied linearly with depth. • Blankets with varying composition were found to perform better than uniform composition breeder blankets. • It was also possible to reduce the amount of beryllium required by the blanket. - Abstract: Beryllium (Be) is a precious resource with many high value uses, the low energy threshold (n,2n) reaction makes Be an excellent neutron multiplier for use in fusion breeder blankets. Estimates of Be requirements and available resources suggest that this could represent a major supply difficulty for solid-type blanket concepts. Reducing the quantity of Be required by breeder blankets would help to alleviate the problem to some extent. In addition, it is important that the reduction in the Be quantity does not diminish the blanket's performance in key aspects such as the tritium breeding ratio (TBR), energy multiplication and peak nuclear heating. Mixed pebble bed designs allow for the multiplier fraction to be varied throughout the blanket. This neutronics study used MCNP 6 to investigate linear variations of the multiplier fraction in relation to blanket depth, in order to better utilise the important multiplying Be(n,2n) and breeding reactions. Blankets with a uniform multiplier fraction showed little scope for reduction in Be mass. Blankets with varying multiplier fractions were able to simultaneously use 10% less Be, increase the energy amplification by 1%, reduce the peak heating by 7% and maintaining a sufficient TBR when compared to the performance achievable using a uniform composition.

  4. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

    Sze, D.K.; Jung, J.; Cheng, E.T.

    1986-01-01

    Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

  5. The blanket interface to TSTA

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Grimm, T.L.; Sze, D.K.; Anderson, J.L.; Bartlit, J.R.; Naruse, Y.; Yoshida, H.

    1988-01-01

    The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

  6. Blanket maintenance by remote means using the cassette blanket approach

    International Nuclear Information System (INIS)

    Werner, R.W.

    1978-01-01

    Induced radioactivity in the blanket and other parts of a fusion reactor close to the plasma zone will dictate remote assembly, disassembly, and maintenance procedures. Time will be of the essence in these procedures. They must be practicable and certain. This paper discusses the reduction of a complicated Tokamak reactor to a simpler assembly via the use of a vacuum building in which to house the reactor and the introduction in this new model of cassette blanket modules. The cassettes significantly simplify remote handling

  7. A Study on Performance Requirements for Advanced Alarm System

    International Nuclear Information System (INIS)

    Seong, Duk Hyun; Jeong, Jae Hoon; Sim, Young Rok; Ko, Jong Hyun; Kim, Jung Seon; Jang, Gwi Sook; Park, Geun Ok

    2005-01-01

    A design goals of advanced alarm system is providing advanced alarm information to operator in main control room. To achive this, we applied computer based system to Alarm System. Because, It should apply data management and advanced alarm processing(ie. Data Base Mangegment System and S/W module for alarm processing). These are not impossible in analog based alarm system. And, preexitance research examples are made on digital computer. We have digital systems for test of advanced alarm system table and have tested and studied using by test equipment in the view point of the system performance, stability and security. In this paper, we discribed about general software architecture of preexitance research examples. Also, CPU performance and requirements of system software that served to accommodate it, stability and security

  8. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H

    2001-11-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

  9. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

    2001-01-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

  10. Breeding blanket design for ITER and prototype (DEMO) fusion reactors and breeding materials issues

    Energy Technology Data Exchange (ETDEWEB)

    Takatsu, H; Enoeda, M [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    Current status of the designs of the ITER breeding blanket and DEMO blankets is introduced placing emphasis on the breeding materials selection and related issues. The former design is based on the up-to-date design activities, as of October 1997, being performed jointly by Joint Central Team (JCT) and Home Teams (HT`s), while the latter is based on the DEMO blanket test module designs being proposed by each Party at the TBWG (Test Blanket Working Group) meetings. (J.P.N.)

  11. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

    General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

  12. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  13. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou

    1998-01-01

    ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

  14. Development of blanket remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

  15. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-06-01

    The fusion blanket surrounds the burning hydrogen core of the fusion reactor. It is in this blanket that most of the energy released by the DT fusion reaction is converted into useable product, and where tritium fuel is produced to enable further operation of the reactor. Blankets will involve new materials, conditions and processes. Several recent fusion blanket concepts are presented to illustrate the range of ideas

  16. Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

    International Nuclear Information System (INIS)

    Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

    2006-01-01

    In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

  17. Assessment of alkali metal coolants for the ITER blanket

    International Nuclear Information System (INIS)

    Natesan, K.; Reed, C.B.; Mattas, R.F.

    1994-01-01

    The blanket system is one of the most important components of 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 different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

  18. Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

    2015-10-15

    Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

  19. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

    Terai, T.; Tanaka, S.; Sze, D.-K.

    2000-01-01

    Force-free helical reactor, FFHR, is a demo-relevant heliotron-type D-T fusion reactor based on the great amount of R and D results obtained in the LHD project. Since 1993, collaboration works have made great progress in design studies of FFHR with standing on the major advantage of current-less steady operation with no dangerous plasma disruptions. There are two types of reference designs, FFHR-1 and FFHR-2, where molten Flibe (LiF-BeF2) is utilized as tritium breeder and coolant. In this paper, we present the outline of FFHR blanket design and some related R and D topics focusing on Be utilization. Beryllium is used as a neutron multiplier in the design and Be pebbles are placed in the front part of the tritium breeding zone. In a Flibe blanket, HF (TF) generated due to nuclear transmutation will be a problem because of its corrosive property. Though nickel-based alloys are thought to be intact in such a corrosive environment, FFHR blanket design does not adopt the alloys because of their induced radioactivity. The present candidate materials for the structure are low-activated ferritic steel (JLF-1), V-4Cr-4Ti, etc. They are capable to be corroded by HF in the operation condition, and Be is expected to work as a reducing agent in the system as well. Whether Be pebbles placed in a Flibe flow can work well or not is a very important matter. From this point, Be solubility in Flibe, reaction rate of the Redox reaction with TF in the liquid and on the surface of Be pebbles under irradiation, flowing behavior of Flibe through a Be pebble bed, etc. should be investigated. In 1997, in order to establish more practical and new data bases for advanced design works, we started a collaboration work of R and D on blanket engineering, where the Be research above mentioned is included. Preliminary dipping-test of Be sheets and in-situ tritium release experiment from Flibe with Be sheets have got started. (orig.)

  20. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

    Billone, M.C.; Dienst, W.; Noda, K.; Roux, N.

    1993-11-01

    The databases for solid breeder ceramics (Li 2 ,O, Li 4 SiO 4 , Li 2 ZrO 3 and LiAlO 2 ) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized

  1. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

    Materials with structures suitable for this purpose are studied. A bibliographic review of the main solid and liquid lithiated compounds is then presented. Erosion, dimensioning and maintenance problems associated with the limiter and the first wall of the reactor are studied from the point of view of the constraints they impose on the design of the blankets. Detailed studies of the main solid and liquid blanket concepts enable the best technological compromises to be determined for the indispensable functions of the blanket to be assured under acceptable conditions. Our analysis leads to four classes of solution, which cannot at this stage be considered as final recommendations, but which indicate what sort of solutions it is worthwhile exploring and comparing in order to be in a position to suggest a realistic blanket at the time when plasma control is sufficiently good for power reactors to be envisaged. Some considerations on the general architecture of the reactor are indicated. Energy storage with pulsed reactors is discussed in the appendix, and a first approach made to minimizing the total tritium recovery [fr

  2. Status of blanket design for RTO/RC ITER

    International Nuclear Information System (INIS)

    Yamada, M.; Ioki, K.; Cardella, A.; Elio, F.; Miki, N.

    2000-01-01

    Design has progressed on the FW/blanket for the RTO/RC (reduced technical objective/ reduced cost) ITER. The basic functions and structures are the same as for the 1998 ITER design. However, design and fabrication methods of the FW/blanket have been improved to achieve ∝ 50% reduction of the construction cost compared to that for the 1998 ITER design. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the EDA (engineering design activity) is still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed. (orig.)

  3. Disruption problematics in segmented blanket concepts

    International Nuclear Information System (INIS)

    Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

    1994-01-01

    In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

  4. The transpiration cooled first wall and blanket concept

    International Nuclear Information System (INIS)

    Barleon, Leopold; Wong, Clement

    2002-01-01

    To achieve high thermal performance at high power density the EVOLVE concept was investigated under the APEX program. The EVOLVE W-alloy first wall and blanket concept proposes to use transpiration cooling of the first wall and boiling or vaporizing lithium (Li) in the blanket zone. Critical issues of this concept are: the Magnetohydrodynamic (MHD) pressure losses of the Li circuit, the evaporation through a capillary structure and the needed superheating of the Li at the first wall and blanket zones. Application of the transpiration concept to the blanket region results in the integrated transpiration cooling concept (ITCC) with either toroidal or poloidal first wall channels. For both orientations the routing of the liquid Li and the Li vapor has been modeled and the corresponding pressure losses have been calculated by varying the width of the supplying slot and the capillary diameter. The concept works when the sum of the active and passive pumping head is higher than the total system pressure losses and when the temperature at the inner side of the first wall does not override the superheating limit of the coolant. This cooling concept has been extended to the divertor design, and the removal of a surface heat flux of up to 10 MW/m 2 appears to be possible, but this paper will focus on the transpiration cooled first wall and blanket concept assessment

  5. Performance evaluation of the sulfur-redox-reaction-activated up-flow anaerobic sludge blanket and down-flow hanging sponge anaerobic/anoxic sequencing batch reactor system for municipal sewage treatment.

    Science.gov (United States)

    Hatamoto, Masashi; Ohtsuki, Kota; Maharjan, Namita; Ono, Shinya; Dehama, Kazuya; Sakamoto, Kenichi; Takahashi, Masanobu; Yamaguchi, Takashi

    2016-03-01

    A sulfur-redox-reaction-activated up-flow anaerobic sludge blanket (UASB) and down-flow hanging sponge (DHS) system, combined with an anaerobic/anoxic sequencing batch reactor (A2SBR), has been used for municipal sewage treatment for over 2 years. The present system achieved a removal rate of 95±14% for BOD, 74±22% for total nitrogen, and 78±25% for total phosphorus, including low water temperature conditions. Sludge conversion rates during the operational period were 0.016 and 0.218 g-VSS g-COD-removed(-1) for the UASB, and DHS, respectively, which are similar to a conventional UASB-DHS system, which is not used of sulfur-redox-reaction, for sewage treatment. Using the sulfur-redox reaction made advanced treatment of municipal wastewater with minimal sludge generation possible, even in winter. Furthermore, the occurrence of a unique phenomenon, known as the anaerobic sulfur oxidation reaction, was confirmed in the UASB reactor under the winter season. Copyright © 2016. Published by Elsevier Ltd.

  6. The Advanced Photon Source: Performance and results from early operation

    International Nuclear Information System (INIS)

    Moncton, D.E.

    1997-01-01

    The Advanced Photon Source at Argonne National Laboratory is now providing researchers with extreme-brilliance undulator radiation in the hard x-ray region of the spectrum. All technical facilities and components are operational and have met design specifications. Fourteen research teams, occupying 20 sectors on the APS experiment hall floor, are currently installing beamline instrumentation or actively taking data. An overview is presented for the first operational years of the Advanced Photon Source. Emphasis is on the performance of accelerators and insertion devices, as well as early scientific results and future plans

  7. Performance and safety design of the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel

  8. Advanced Performance Modeling with Combined Passive and Active Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Dovrolis, Constantine [Georgia Inst. of Technology, Atlanta, GA (United States); Sim, Alex [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2015-04-15

    To improve the efficiency of resource utilization and scheduling of scientific data transfers on high-speed networks, the "Advanced Performance Modeling with combined passive and active monitoring" (APM) project investigates and models a general-purpose, reusable and expandable network performance estimation framework. The predictive estimation model and the framework will be helpful in optimizing the performance and utilization of networks as well as sharing resources with predictable performance for scientific collaborations, especially in data intensive applications. Our prediction model utilizes historical network performance information from various network activity logs as well as live streaming measurements from network peering devices. Historical network performance information is used without putting extra load on the resources by active measurement collection. Performance measurements collected by active probing is used judiciously for improving the accuracy of predictions.

  9. EU DEMO blanket concepts safety assessment. Final report of Working Group 6a of the Blanket Concept Selection Exercise

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Porfiri, T.

    1996-06-01

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four blanket concepts under development. Two of them use lithium ceramics, the other two concepts employ an eutectic lead-lithium alloy (Pb-17Li) as breeder material. The two most promising concepts were to select in 1995 for further development. In order to prepare the selection, a Blanket Concept Selection Exercise (BCSE) has been inititated by the participating associations under the auspices of the European Commission. This BCSE has been performed in 14 working groups which, in a comparative evaluation of the four blanket concepts, addressed specific fields. The working group safety addressed the safety implications. This report describes the methodology adopted, the safety issues identified, their comparative evaluation for the four concepts, and the results and conclusions of the working group to be entered into the overall evaluation. There, the results from all 14 working groups have been combined to yield a final ranking as a basis for the selection. In summary, the safety assessment showed that the four European blanket concepts can be considered as equivalent in terms of the safety rating adopted, each concept, however, rendering safety concerns of different quality in different areas which are substantiated in this report. (orig.) [de

  10. Neutronics design aspects of reference ARIES-I fusion blanket

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1990-12-01

    A SiC composite blanket concept was recently conceived for a deuterium-tritium burning, 1000 MW(e) tokamak fusion reactor design, ARIES-I. SiC composite structural material was chosen due to its very low activation features. High blanket nuclear performance and thermal efficiency, adequate tritium breeding, and a low level of activation are important design requirements for the ARIES-I reactor. The major approaches, other than using SiC as structural material, in meeting these design requirements, are to employ beryllium, the only low activation neutron multiplying material, and isotopically tailored Li 2 ZrO 3 , a tritium breeding material stable at high temperature, as blanket materials. 5 refs., 4 figs., 2 tabs

  11. Status of fusion reactor blanket evaluation studies in France

    International Nuclear Information System (INIS)

    Carre, F.; Chevereau, G.; Gervaise, F.; Proust, E.

    1985-03-01

    In the frame of recent CEA studies aiming at the evaluation and at the comparison of various candidate blanket concepts in moderate power conditions (Psub(n) approximately 2 MW/m 2 ), the present work examines the neutronic and thermomechanical performances of a water cooled Li 17 Pb 83 tubular blanket and those of a helium cooled canister blanket taking advantage of the excellent breeding capability of composite Beryllium/LiAlO 2 (85/15%) breeder elements. The purpose of the following discussion is to justify the impetus for these reference concepts and to summarize the state of their evaluation studies updated by the continuous assimilation of calculations and experiments in progress

  12. Progress and achievements of the ITER L-4 blanket project

    International Nuclear Information System (INIS)

    Daenner, W.; Toschi, R.; Cardella, A.

    1999-01-01

    The L-4 Blanket Project embraces the R and D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R and D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined.The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification tests. (author)

  13. Ceramic BOT type blanket with poloidal helium cooling

    International Nuclear Information System (INIS)

    Cardella, A.; Daenenr, W.; Iseli, M.; Ferrari, M.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.

    1989-01-01

    This paper briefly describes the work done and results achieved over the past two years on the ceramic breeder BOT blanket with poloidal helium cooling. A conclusive remark on the brick/plate option described previously is followed by short descriptions of the low and high performance pebble bed options elaborated as alternatives for both NET and DEMO. The results show, togethre with those about the poloidal cooling of the First Wall, good prospects for this blanket type provided that the questions connected wiht an extensive use of beryllium find a satisfactor answer. (author). 5 refs.; 7 figs.; 1 tab

  14. Progress and achievements of the ITER L-4 blanket project

    International Nuclear Information System (INIS)

    Daenner, W.; Toschi, R.; Cardella, A.

    2001-01-01

    The L-4 Blanket Project embraces the R and D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R and D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined. The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification tests. (author)

  15. Effect of reactor size on the breeding economics of LMFBR blankets

    International Nuclear Information System (INIS)

    Tagishi, A.; Driscoll, M.J.

    1975-02-01

    The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

  16. 2000 GUIDELINES FOR ADVANCEMENT AND EXCEPTIONAL PERFORMANCE AWARDS

    CERN Document Server

    2000-01-01

    1. GeneralAs indicated in last year's guidelines, studies on CERN's career structures are continuing in the context of the work of the Tripartite Forum on Employment Conditions (TREF) and should be completed during this year's 5-Yearly Remuneration Review for the application of any changes in 2001. In the meantime, the 2000 annual advancement exercise will be conducted on similar basic lines to those of previous years, i.e. within the procedures of the Administrative Circular N° 26 (Rev. 2), thus promoting the aspect of individual merit in advancement.The present guidelines continue to aim at the best possible degree of flexibility of application for each division within the present Staff Rules and Regulations in order to take full account of both the different levels of individual performance and also the different staff profiles.All advancement must be justified on the grounds of correctly judged merit: exceptional advancement (i.e. second step or entering the exceptional advancement grade) for par...

  17. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-11-01

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  18. Pulsed activation analyses of the ITER blanket design options considered in the blanket trade-off study

    International Nuclear Information System (INIS)

    Wang, Q.; Henderson, D.L.

    1995-01-01

    Pulsed activation calculations have been performed on two blanket options considered as part of the ITER home team blanket trade-off study. The objective was to compare the activity, afterheat and waste disposal rating (WDR) results of a composite blanket-shield design for the continuous operation approximation to a pulsed operation case to determine whether the differences are at most the duty factor as predicted by the two nuclide chain model. Up to a cooling period of 100 years, the pulsed activity and afterheat values were below the continuous oepration results and well within (except for one afterheat value) the maximum deviation predicted by the two nuclide chain model. No differences in the WDR values were noted as they are, to a large extent, based on long-lived nuclides which are insensitive to short-term changes in the operation history. (orig.)

  19. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  20. Thermohydraulics design and thermomechanics analysis of two European breeder blanket concepts for DEMO. Pt. 1 and Pt. 2. Pt. 1: BOT helium cooled solid breeding blanket. Pt. 2: Dual coolant self-cooled liquid metal blanket

    International Nuclear Information System (INIS)

    Norajitra, P.

    1995-06-01

    Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li 4 SiO 4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.) [de

  1. Advanced fuel technology and performance: Current status and trends

    International Nuclear Information System (INIS)

    1990-11-01

    During the last years the Nuclear Fuel Cycle and Waste Management Division of the IAEA has been giving great attention to the collection, analysis and exchange of information in the field of reactor fuel technology. Most of these activities are being conducted in the framework of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). The purpose of this Advisory Group Meeting on Advanced Fuel Technology and Performance was to update and to continue the previous work, and to review the experience of advanced fuel technology, its performance with regard to all types of reactors and to outline the future trends on the basis of national experience and discussions during the meeting. As a result of the meeting a Summary Report was prepared which reflected the status of the advanced nuclear fuel technology up to 1990. The 10 papers presented by participants of this meeting are also published here. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  2. Evaluation of potential blanket concepts for a Demonstration Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Chapin, D.L.; Chi, J.W.H.; Kelly, J.L.

    1978-01-01

    An evaluation has been made of several different blanket concepts for use in a near-term Demonstration Tokamak Hybrid Reactor (DTHR), whose main objective would be to produce a significant amount of fissile fuel while demonstrating the feasibility of the tokamak hybrid reactor concept. The desirability of a simple design using proven technology plus a proliferation resistant fuel cycle led to the selection of a low temperature and pressure water-cooled, zircaloy clad ThO 2 blanket concept to breed 233 U. The nuclear performance and thermal-hydraulics characteristics of the blanket were evaluated to arrive at a consistent design. The blanket was found to be feasible for producing a significant amount of fissile fuel even with the limited operating conditions and blanket coverage in the DTHR

  3. MIT LMFBR blanket physics project progress report No. 7, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1976-01-01

    Work during the period was devoted primarily to a range of analytical/numerical investigations, including evaluation of means to improve external blanket designs, beneficial attributes of the use of internal blankets, improved methods for the calculation of heterogeneous self-shielding and parametric studies of calculated spectral indices. Experimental work included measurements of the ratio of U-238 captures to U-235 fissions in a standard blanket mockup, and completion of development work on the radiophotoluminescent readout of LiF thermoluminescent detectors. The most significant findings were that there is very little prospect for substantial improvement in the breeding performance of external blankets, but internal blankets continue to show promise, particularly if they are used in such a way as to increase the volume fraction of fuel inside the core envelope. An improved equivalence theorem was developed which may allow use of fast reactor methods to calculate heterogeneously self-shielded cross sections in both fast and thermal reactors

  4. Emission and thermal performance upgrade through advanced control backfit

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, A.K. [Stone & Webster Engineering Corporation, Boston, MA (United States)

    1994-12-31

    Reducing emission and improving thermal performance of currently operating power plants is a high priority. A majority of these power plants are over 20 years old with old control systems. Upgrading the existing control systems with the latest technology has many benefits, the most cost beneficial are the reduction of emission and improving thermal performance. The payback period is usually less than two years. Virginia Power is installing Stone & Webster`s NO{sub x} Emissions Advisor and Advanced Steam Temperature Control systems on Possum Point Units 3 and 4 to achieve near term NO{sub x} reductions while maintaining high thermal performance. Testing has demonstrated NO{sub x} reductions of greater than 20 percent through the application of NO{sub x} Emissions Advisor on these units. The Advanced Steam Temperature Control system which has been operational at Virginia Power`s Mt. Storm Unit 1 has demonstrated a signification improvement in unit thermal performance and controllability. These control systems are being combined at Units 3 and 4 to reduce NO{sub x} emissions and achieve improved unit thermal performance and control response with the existing combustion hardware. Installation has been initiated and is expected to be completed by the spring of 1995. Possum Point Power Station Units 3 and 4 are pulverized coal, tangentially fired boilers producing 107 and 232 MW and have a distributed control system and a PC based performance monitoring system. The installation of the advanced control and automation system will utilize existing control equipment requiring the addition of several PCs and PLC.

  5. The TFTR lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bertone, P.C.; Creedon, R.L.; File, J.; Graumann, D.W.

    1985-01-01

    The Lithium Blanket Module (LBM) is an approximately 80X80X80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li 2 O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li 2 O pellets with satisfactory reproducibility were developed using purified Li 2 O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1-1nCi/g)

  6. Performance Measurement of Advanced Stirling Convertors (ASC-E3)

    Science.gov (United States)

    Oriti, Salvatore M.

    2013-01-01

    NASA Glenn Research Center (GRC) has been supporting development of the Advanced Stirling Radioisotope Generator (ASRG) since 2006. A key element of the ASRG project is providing life, reliability, and performance testing data of the Advanced Stirling Convertor (ASC). The latest version of the ASC (ASC-E3, to represent the third cycle of engineering model test hardware) is of a design identical to the forthcoming flight convertors. For this generation of hardware, a joint Sunpower and GRC effort was initiated to improve and standardize the test support hardware. After this effort was completed, the first pair of ASC-E3 units was produced by Sunpower and then delivered to GRC in December 2012. GRC has begun operation of these units. This process included performance verification, which examined the data from various tests to validate the convertor performance to the product specification. Other tests included detailed performance mapping that encompassed the wide range of operating conditions that will exist during a mission. These convertors were then transferred to Lockheed Martin for controller checkout testing. The results of this latest convertor performance verification activity are summarized here.

  7. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

    The objectives of the Blanket Comparison and Selection Study (BCSS) can be stated as follows: (1) Define a small number (approx. 3) of blanket design concepts that should be the focus of the blanket R and D program. A design concept is defined by the selection of all materials (e.g., breeder, coolant, structure and multiplier) and other major characteristics that significantly influence the R and D requirements. (2) Identify and prioritize the critical issues for the leading blanket concepts. (3) Provide the technical input necessary to develop a blanket R and D program plan. Guidelines for prioritizing the R and D requirements include: (a) critical feasibility issues for the leading blanket concepts will receive the highest priority, and (b) for equally important feasibility issues, higher R and D priority will be given to those that require minimum cost and short time

  8. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

    Malang, S.; Casal, V.; Arheidt, K.; Fischer, U.; Link, W.; Rust, K.

    1986-01-01

    A blanket concept for NET using liquid lithium-lead both as breeder material and as coolant is described. The need for inboard breeding is avoided by using beryllium as neutron multiplier in the outboard blanket. Novel flow channel inserts are employed in all poloidal ducts to reduce the MHD pressure drop. The concept offers a simple mechanical design and a higher tritium breeding ratio compared to water- and gas-cooled blankets. (author)

  9. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

    Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Usher, J.L.

    1980-04-01

    Definitions are given of 10 generic blanket types and the specific blanket chosen to be analyzed in detail from each of the 10 types. Dimensions, compositions, energy depositions and breeding ratios (where applicable) are presented for each of the 10 designs. Ultimately, based largely on neutronics and thermal hyraulics results, breeding an nonbreeding blanket options are selected for further design analysis and integration with a suitable power conversion subsystem

  10. Trade-off study of liquid-metal self-cooled blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

    A trade-off study of liquid-metal self-cooled blankets was carried out to define the performance of these blankets with respect to the main functions in a fusion reactor, and to determine the potential to operate at the maximum possible values of the performance parameters. The main purpose is to improve the reactor economics by maximizing the blanket energy multiplication factor, reduce the capital cost of the reactor, and satisfy the design requirements. The main parameters during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the 6 Li enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, the impact of different reactor design choices on the performance parameters was analyzed. The effect of the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, the coolant choice for the nonbreeding inboard blanket, and the neutron source distribution were part of the trade-off study. In addition, tritium breeding benchmark calculations were performed to study the impact of the use of different transport codes and nuclear data libraries. The importance and the negative effect of high TBR on the energy multiplication motivated the benchmark calculations

  11. Advanced non-destructive methods for an efficient service performance

    International Nuclear Information System (INIS)

    Rauschenbach, H.; Clossen-von Lanken Schulz, M.; Oberlin, R.

    2015-01-01

    Due to the power generation industry's desire to decrease outage time and extend inspection intervals for highly stressed turbine parts, advanced and reliable Non-destructive methods were developed by Siemens Non-destructive laboratory. Effective outage performance requires the optimized planning of all outage activities as well as modern Non-destructive examination methods, in order to examine the highly stressed components (turbine rotor, casings, valves, generator rotor) reliably and in short periods of access. This paper describes the experience of Siemens Energy with an ultrasonic Phased Array inspection technique for the inspection of radial entry pinned turbine blade roots. The developed inspection technique allows the ultrasonic inspection of steam turbine blades without blade removal. Furthermore advanced Non-destructive examination methods for joint bolts will be described, which offer a significant reduction of outage duration in comparison to conventional inspection techniques. (authors)

  12. Design development and manufacturing sequence of the European water-cooled Pb-17Li test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Futterer, M.A.; Bielak, B.; Deffain, J.P.; Giancarli, L.; Li Puma, A.; Salavy, J.F.; Szczepanski, J. [CEA Saclay, Gif-sur-Yvette (France). FDRN/DMT/SERMA; Dellis, C. [CEA Grenoble, DTA-CEREM/SGM, Grenoble (France); Nardi, C. [ENEA Frascati, ERG-FUS-TECN-MEC, Frascati (Italy); Schleisiek, K. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit

    1998-09-01

    In 1996, the European Community started the development of a water-cooled Pb17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the basic performance phase prior to D-T operation. The test module is designed to be a representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined. (orig.) 11 refs.

  13. First wall and blanket module safety enhancement by material selection and design decision

    International Nuclear Information System (INIS)

    Merrill, B.J.

    1980-01-01

    A thermal/mechanical study has been performed which illustrates the behavior of a fusion reactor first wall and blanket module during a loss of coolant flow event. The relative safety advantages of various material and design options were determined. A generalized first wall-blanket concept was developed to provide the flexibility to vary the structural material (stainless steel vs titanium), coolant (helium vs water), and breeder material (liquid lithium vs solid lithium aluminate). In addition, independent vs common first wall-blanket cooling and coupled adjacent module cooling design options were included in the study. The comparative analyses were performed using a modified thermal analysis code to handle phase change problems

  14. Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

    Science.gov (United States)

    Xiaokang, ZHANG; Songlin, LIU; Xia, LI; Qingjun, ZHU; Jia, LI

    2017-11-01

    The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR). Some updating of neutronics analyses was needed, because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket, including the optimization of radial build-up and customized structure for each blanket module. A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses. The tritium breeding capability, nuclear heating power, radiation damage, and decay heat were calculated by the MCNP and FISPACT code. The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency. The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW. The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60, respectively. The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module #3. The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time. The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.

  15. Effect of technological advances on cochlear implant performance in adults.

    Science.gov (United States)

    Lenarz, Minoo; Joseph, Gert; Sönmez, Hasibe; Büchner, Andreas; Lenarz, Thomas

    2011-12-01

    To evaluate the effect of technological advances in the past 20 years on the hearing performance of a large cohort of adult cochlear implant (CI) patients. Individual, retrospective, cohort study. According to technological developments in electrode design and speech-processing strategies, we defined five virtual intervals on the time scale between 1984 and 2008. A cohort of 1,005 postlingually deafened adults was selected for this study, and their hearing performance with a CI was evaluated retrospectively according to these five technological intervals. The test battery was composed of four standard German speech tests: Freiburger monosyllabic test, speech tracking test, Hochmair-Schulz-Moser (HSM) sentence test in quiet, and HSM sentence test in 10 dB noise. The direct comparison of the speech perception in postlingually deafened adults, who were implanted during different technological periods, reveals an obvious improvement in the speech perception in patients who benefited from the recent electrode designs and speech-processing strategies. The major influence of technological advances on CI performance seems to be on speech perception in noise. Better speech perception in noisy surroundings is strong proof for demonstrating the success rate of new electrode designs and speech-processing strategies. Standard (internationally comparable) speech tests in noise should become an obligatory part of the postoperative test battery for adult CI patients. Copyright © 2011 The American Laryngological, Rhinological, and Otological Society, Inc.

  16. Performance assessment of advanced engineering workstations for fuel management applications

    International Nuclear Information System (INIS)

    Turinsky, P.J.

    1989-07-01

    The purpose of this project was to assess the performance of an advanced engineering workstation [AEW] with regard to applications to incore fuel management for LWRs. The attributes of most interest to us that define an AEW are parallel computational hardware and graphics capabilities. The AEWs employed were super microcomputers manufactured by MASSCOMP, Inc. These computers utilize a 32-bit architecture, graphics co-processor, multi-CPUs [up to six] attached to common memory and multi-vector accelerators. 7 refs., 33 figs., 4 tabs

  17. Advanced Modified High Performance Synthetic Jet Actuator with Curved Chamber

    Science.gov (United States)

    Xu, Tian-Bing (Inventor); Su, Ji (Inventor); Jiang, Xiaoning (Inventor)

    2014-01-01

    The advanced modified high performance synthetic jet actuator with optimized curvature shape chamber (ASJA-M) is a synthetic jet actuator (SJA) with a lower volume reservoir or chamber. A curved chamber is used, instead of the conventional cylinder chamber, to reduce the dead volume of the jet chamber and increase the efficiency of the synthetic jet actuator. The shape of the curvature corresponds to the maximum displacement (deformation) profile of the electroactive diaphragm. The jet velocity and mass flow rate for the ASJA-M will be several times higher than conventional piezoelectric actuators.

  18. Proceedings of the fifteenth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    Tanigawa, Hisashi; Enoeda, Mikio

    2010-03-01

    This report is the Proceedings of 'the Fifteenth International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors. This workshop was held in Sapporo, Japan on 3-4, Sept. 2009. Twenty six participants from EU, Japan, India, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket development. By this workshop, advance of key technologies for solid breeder blanket development was shared among the participants. Also, desired direction of further investigation and development was recognized. The 20 of the presented papers are indexed individually. (J.P.N.)

  19. Proceedings of the fifteenth international workshop on ceramic breeder blanket interactions

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hisashi; Enoeda, Mikio [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki (Japan)

    2010-03-15

    This report is the Proceedings of 'the Fifteenth International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors. This workshop was held in Sapporo, Japan on 3-4, Sept. 2009. Twenty six participants from EU, Japan, India, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket development. By this workshop, advance of key technologies for solid breeder blanket development was shared among the participants. Also, desired direction of further investigation and development was recognized. The 20 of the presented papers are indexed individually. (J.P.N.)

  20. A development of user-friendly graphical interface for a blanket simulator

    International Nuclear Information System (INIS)

    Lee, Young-Seok; Yoon, Seok-Heun; Han, Jung-Hoon

    2010-01-01

    A web-based user-friendly graphical interface (GUI) system, named GUMBIS (Graphical User-friendly Monte-Carlo-Application Blanket-Design Interface System), was developed to cut down the efforts of the researchers and practitioners who study tokamak blanket designs with the Monte Carlo MCNP/MCNPX codes. GUMBIS was also aimed at supporting them to use the codes for their study without having through understanding on the complex menus and commands of the codes. Developed on the web-based environment, GUMBIS provides task sharing capability on a network. GUMBIS, applicable for both blanket design and neutronics analysis, could facilitate not only advanced blanket R and D but also the education and training of the researchers in the R and D.

  1. Irradiation performance updates on Korean advanced fuels for PWRs

    International Nuclear Information System (INIS)

    Jang, Y.K.; Jeon, K.L.; Kim, Y.H.; Yoo, J.S.; Kim, J.I.; Shin, J.C.; Chung, J.G.; Park, J.R.; Chung, S.K.; Kim, T.W.; Yoon, Y.B.; Park, K.M.; Yoo, M.J.; Kim, M.S.; Lee, T.H.

    2010-01-01

    The developments of advanced nuclear fuels for PWRs were started in 1999 and in 2001, respectively: PLUS7 TM for eight operating optimized power reactors of 1000 MWe class (OPR1000) and four advanced power reactors of 1400 MWe class (APR1400) under construction, and 16ACE7 TM and 17ACE7 TM for an operating 16x16 Westinghouse type plant and six operating 17x17 Westinghouse type plants. The design targets were as follows: batch average burnup up to 55 GWD/MTU, over 10% thermal margin increase, improvement of the mechanical integrity of higher seismic capability, higher debris or grid fretting wear performance, higher control rod insertion capability, increase of neutron economy, improvement of manufacturability, solving incomplete rod insertion (IRI) issue and top nozzle screw failure issue, etc. in comparison of the existing nuclear fuels. The irradiation tests using each four LTAs (Lead Test Assemblies) during 3 cycles were completed in three Korean nuclear reactors until 2009. The eight irradiation performance items which are assembly growth, rod growth, grid width growth, assembly bow, rod bow, assembly twist, rod diameter and cladding oxidation were examined in pool-side after each cycle and evaluated. The irradiation tests could be continued by expecting the good performances for next cycle from the previous cycle. After 2 cycle irradiations, the region implementation could be started in 15 nuclear power plants. Even though the verifications using the LTAs were completed, each surveillance program was launched and the irradiation performance data were being updated during region implementation. In addition to pool-side examinations (PSEs) by assembly-wise during irradiation tests, six rod-wise performance items were also examined in pool-side using each LTA after discharge. All performance items met their design criteria as a result of the evaluation. Even though the interesting ones among the irradiation performance parameters were assembly and grid growths

  2. Role of Fabrication on Materials Compatibility in APT Target/Blanket

    International Nuclear Information System (INIS)

    Iyer, N.; Louthan, M.R. Jr.; Dunn, K.; Fisher, D.L.

    1998-09-01

    This paper summarizes several of the options associated with the fabrication of selected target/blanket components. In addition, the materials characterization technologies required to validate these components performance is presented

  3. Design requirement on KALIMER blanket fuel assembly duct

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

    1998-03-01

    This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs

  4. On the conditions of existence of cold-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-12-01

    An extende analysis of the partially ionized boundary layer of a magnetized plasma has been performed, leading to the following results: (i) In a first approximation the ion density at the inner ''edge'' of the layer becomes related to the wall-near neutral gas density, in a way being independent of the spatial distribution of the ionization rate. (ii) The particle and momentum balance equations, and the associated impermeability condition of the plasma with respect to neutral gas penetration, are not sufficient to specify a cold-blanket state, but have to be combined with considerations of the heat blance. This leads to lower and upper power input limits, thus defining conditions for the existence of a cold-blanket state. At decreasing beta values , or increasing radiation losses, there are situations where such a state cannot exist at all. (iii) It should become possible to fulfill the cold-blanket conditions in full-scale reactors as well as in certain model experiments. Probably these conditions can also be satisfied in large tokamaks like JET, and by fast gas injection in devices such as Alcator, but not in medium-size tokamaks being operated at moderately high ion densities. (iv) A strong ''boundary layer stabilization'' mechanism due to the joint viscosity-resistivity-pressure effects is available under cold-blanket conditions. (author)

  5. Progress in blanket designs using SiCf/SiC composites

    International Nuclear Information System (INIS)

    Giancarli, L.; Golfier, H.; Nishio, S.; Raffray, R.; Wong, C.; Yamada, R.

    2002-01-01

    This paper summarizes the most recent design activities concerning the use of SiC f /SiC composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the TAURO blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium-lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and ARIES-AT blankets are essentially formed by a SiC f /SiC box acting as a container for the lithium-lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. The DREAM blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li 2 O (or other lithium ceramics) as breeder material and of SiC as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R and D on SiC f /SiC

  6. Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

    2016-11-01

    Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

  7. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  8. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  9. The application of advanced rotor (performance) methods for design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Bussel, G.J.W. van [Delft Univ. of Technology, Inst. for Wind Energy, Delft (Netherlands)

    1997-08-01

    The calculation of loads and performance of wind turbine rotors has been a topic for research over the last century. The principles for the calculation of loads on rotor blades with a given specific geometry, as well as the development of optimal shaped rotor blades have been published in the decades that significant aircraft development took place. Nowadays advanced computer codes are used for specific problems regarding modern aircraft, and application to wind turbine rotors has also been performed occasionally. The engineers designing rotor blades for wind turbines still use methods based upon global principles developed in the beginning of the century. The question what to expect in terms of the type of methods to be applied in a design environment for the near future is addressed here. (EG) 14 refs.

  10. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    Directory of Open Access Journals (Sweden)

    Kooyman Timothée

    2017-01-01

    Full Text Available Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long- and short-term neutron and gamma source is carried out whereas in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

  11. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    Science.gov (United States)

    Kooymana, Timothée; Buiron, Laurent; Rimpault, Gérald

    2017-09-01

    Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long and short term neutron and gamma source is carried out while in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

  12. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    International Nuclear Information System (INIS)

    Kooyman, T.; Buiron, L.; Rimpault, G.

    2017-01-01

    Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long- and short-term neutron and gamma source is carried out whereas in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing. (authors)

  13. Performance evaluation of scientific programs on advanced architecture computers

    International Nuclear Information System (INIS)

    Walker, D.W.; Messina, P.; Baille, C.F.

    1988-01-01

    Recently a number of advanced architecture machines have become commercially available. These new machines promise better cost-performance then traditional computers, and some of them have the potential of competing with current supercomputers, such as the Cray X/MP, in terms of maximum performance. This paper describes an on-going project to evaluate a broad range of advanced architecture computers using a number of complete scientific application programs. The computers to be evaluated include distributed- memory machines such as the NCUBE, INTEL and Caltech/JPL hypercubes, and the MEIKO computing surface, shared-memory, bus architecture machines such as the Sequent Balance and the Alliant, very long instruction word machines such as the Multiflow Trace 7/200 computer, traditional supercomputers such as the Cray X.MP and Cray-2, and SIMD machines such as the Connection Machine. Currently 11 application codes from a number of scientific disciplines have been selected, although it is not intended to run all codes on all machines. Results are presented for two of the codes (QCD and missile tracking), and future work is proposed

  14. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

    A mechanical configuration of ITER integrated primary first wall/shield blanket module were developed focusing on the welded attachment of its support leg to the back plate. A 100 mm x 150 mm space between the legs of adjacent modules was incorporated for the working space of welding/cutting tools. A concept of coolant branch pipe connection to accommodate deformation due to the leg welding and differential displacement of the module and the manifold/back plate during operation was introduced. Two-dimensional FEM analyses showed that thermal stresses in Cu-alloy (first wall) and stainless steel (first wall coolant tube and shield block) satisfied the stress criteria following ASME code for ITER BPP operation. On the other hand, three-dimensional FEM analyses for overall in-vessel structures exhibited excessive primary stresses in the back plate and its support structure to the vacuum vessel under VDE disruption load and marginal stresses in the support leg of module No.4. Fabrication procedure of the integrated primary first wall/shield blanket module was developed based on single step solid HIP for the joining of Cu-alloy/Cu-alloy, Cu-alloy/stainless steel, and stainless steel/stainless steel. (author)

  15. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

    The NET machine has three horizontal ports reserved for testing tritium breeding blanket designs during the physics phase and possibly five during the technology phase. The design of the ports and test blankets are modular to accept a range of blanket options, provide radiation shielding and allow routine replacement. Radiation levels during replacement or maintenance require that all operations must be carried out remotely. The paper describes the problems overcome in providing a port design which includes attachment to the vacuum vessel with double vacuum seals, an integrated cooled first wall and support guides for the test blanket module. The method selected to remotely replace the test module whilst controlling the spread of contamination is also adressed. The paper concludes that the provisions of a test blanket facility based on the NET machine design is feasible. (orig.)

  16. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

    This paper presents an overview of the critical materials issues that must be considered in the development of a tritium breeding blanket for a tokamak fusion reactor that operates on the D-T-Li fuel cycle. The primary requirements of the blanket system are identified and the important criteria that must be considered in the development of blanket technology are summarized. The candidate materials are listed for the different blanket components, e.g., breeder, coolant, structure and neutron multiplier. Three blanket concepts that appear to offer the most potential are: (1) liquid-metal breeder/coolant, (2) liquid-metal breeder/separate coolant, and (3) solid breeder/separate coolant. The major uncertainties associated with each of the design concepts are discussed and the key materials R and D requirements for each concept are identified

  17. Preliminary Analysis for K-DEMO Water Cooled Breeding Blanket Using MARS-KS

    International Nuclear Information System (INIS)

    Lee, Jeong-Hun; Kim, Geon-Woo; Park, Goon-Cherl; Cho, Hyoung-Kyu; Im, Kihak

    2014-01-01

    the pressure drops in them and transfers the flow rates to each DLL file as inlet flow rate boundary conditions. The methodology was applied for the two blanket module simulation and its performance was successfully validated. The multiple blanket module analysis method will be extended to the entire system analysis following the present work

  18. Feasibility study of incore fission chamber application for neutron flux measurements on the NET blanket

    International Nuclear Information System (INIS)

    Bertalot, L.

    1987-01-01

    A feasibility study has been carried out on the use of in-core fission chambers as neutron diagnostic tools to perform neutron flux measurements on the blanket component of NET. The high neutron and gamma fluxes and the severe thermal-mechanical and magnetic conditions of the blanket structure have been taken into account in this analysis. Preliminary design criteria and specifications of an in-core detector are presented for NET application. A research and development programme is outlined which aims to obtain more information on the tecnological constraints arising from the severe conditions of the NET blanket

  19. Fuel balance in nuclear power with fast reactors without a uranium blanket

    International Nuclear Information System (INIS)

    Naumov, V.V.; Orlov, V.V.; Smirnov, V.S.

    1994-01-01

    General aspects related to replacing the uranium blanket of a lead-cooled fast reactor burning uranium-plutonium nitride fuel with a more efficient lead reflector are briefly discussed in the article. A study is very briefly summarized, which showed that a breeding ratio of about 1 and electric power of about 300 MW were achievable. A nuclear fuel balance is performed to estimate the increased consumption of uranium to produce power and the gains achievable by eliminating the uranium blanket. Elimination of the uranium blanket has the advantages of simplifying and improving the fast reactor and eliminating the production of weapons quality plutonium. 3 figs

  20. Required momentum, heat, and mass transport experiments for liquid-metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Sze, D.K.; Abdou, M.A.

    1986-01-01

    Through the effects on fluid flow, many aspects of blanket behavior are affected by magnetohydrodynamic (MHD) effects, including pressure drop, heat transfer, mass transfer, and structural behavior. In this paper, a set of experiments is examined that could be performed in order to reduce the uncertainties in the highly related set of issues dealing with momentum, heat, and mass transport under the influence of a strong magnetic field (i.e., magnetic transport phenomena). By improving our basic understanding and by providing direct experimental data on blanket behavior, these experiments will lead to improved designs and an accurate assessment of the attractiveness of liquid-metal blankets

  1. Experimental investigation on streaming due to a gap between blanket modules in ITER

    International Nuclear Information System (INIS)

    Konno, Chikara; Maekawa, Fujio; Oyama, Yukio; Uno, Yoshitomo; Kasugai, Yoshimi; Maekawa, Hiroshi; Ikeda, Yujiro; Wada, Masayuki

    2000-01-01

    A gap streaming experiment was performed by using a D-T neutron source at FNS/JAERI as the ITER/EDA R and D Task T-218, in order to examine the streaming effects due to gap between shield blanket modules in ITER. The experiment had three phases. The first one defined neutron source characteristics (Source Characterization Experiment), the second (Experiment-l ) aimed at shield for welding part between shield blanket and back plate and the third (Experiment-2) focused on the influence that the gap between shield blanket modules would have on superconducting magnet. The effects of gap streaming were examined in detail experimentally. (author)

  2. Development and testing of a zero stitch MLI blanket using plastic pins for space use

    Science.gov (United States)

    Hatakenaka, Ryuta; Miyakita, Takeshi; Sugita, Hiroyuki; Saitoh, Masanori; Hirai, Tomoyuki

    2014-11-01

    New types of MLI blanket have been developed to achieve high thermal performance while maintaining production and assembly workability equivalent to the conventional type. Tag-pins, which are widely used in commercial applications to hook price tags to products, are used to fix the films in place and the pin material is changed to polyetheretherketone (PEEK) for use in space. Thermal performance is measured by using a boil-off calorimeter, in which a rectangular liquid nitrogen tank is used to evaluate the degradation at the bending corner and joint of the blanket. Zero-stitch- and multi-blanket-type MLIs show significantly improved thermal performance (ɛeff is smaller than 0.0050 at room temperature) despite having the same fastener interface as traditional blankets, while the venting design and number of tag-pins are confirmed as appropriate in a depressurization test.

  3. Thermal Model Predictions of Advanced Stirling Radioisotope Generator Performance

    Science.gov (United States)

    Wang, Xiao-Yen J.; Fabanich, William Anthony; Schmitz, Paul C.

    2014-01-01

    This paper presents recent thermal model results of the Advanced Stirling Radioisotope Generator (ASRG). The three-dimensional (3D) ASRG thermal power model was built using the Thermal Desktop(trademark) thermal analyzer. The model was correlated with ASRG engineering unit test data and ASRG flight unit predictions from Lockheed Martin's (LM's) I-deas(trademark) TMG thermal model. The auxiliary cooling system (ACS) of the ASRG is also included in the ASRG thermal model. The ACS is designed to remove waste heat from the ASRG so that it can be used to heat spacecraft components. The performance of the ACS is reported under nominal conditions and during a Venus flyby scenario. The results for the nominal case are validated with data from Lockheed Martin. Transient thermal analysis results of ASRG for a Venus flyby with a representative trajectory are also presented. In addition, model results of an ASRG mounted on a Cassini-like spacecraft with a sunshade are presented to show a way to mitigate the high temperatures of a Venus flyby. It was predicted that the sunshade can lower the temperature of the ASRG alternator by 20 C for the representative Venus flyby trajectory. The 3D model also was modified to predict generator performance after a single Advanced Stirling Convertor failure. The geometry of the Microtherm HT insulation block on the outboard side was modified to match deformation and shrinkage observed during testing of a prototypic ASRG test fixture by LM. Test conditions and test data were used to correlate the model by adjusting the thermal conductivity of the deformed insulation to match the post-heat-dump steady state temperatures. Results for these conditions showed that the performance of the still-functioning inboard ACS was unaffected.

  4. Advanced multistage turbine blade aerodynamics, performance, cooling, and heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Fleeter, S.; Lawless, P.B. [Purdue Univ., West Lafayette, IN (United States)

    1995-10-01

    The gas turbine has the potential for power production at the highest possible efficiency. The challenge is to ensure that gas turbines operate at the optimum efficiency so as to use the least fuel and produce minimum emissions. A key component to meeting this challenge is the turbine. Turbine performance, both aerodynamics and heat transfer, is one of the barrier advanced gas turbine development technologies. This is a result of the complex, highly three-dimensional and unsteady flow phenomena in the turbine. Improved turbine aerodynamic performance has been achieved with three-dimensional highly-loaded airfoil designs, accomplished utilizing Euler or Navier-Stokes Computational Fluid Dynamics (CFD) codes. These design codes consider steady flow through isolated blade rows. Thus they do not account for unsteady flow effects. However, unsteady flow effects have a significant impact on performance. Also, CFD codes predict the complete flow field. The experimental verification of these codes has traditionally been accomplished with point data - not corresponding plane field measurements. Thus, although advanced CFD predictions of the highly complex and three-dimensional turbine flow fields are available, corresponding data are not. To improve the design capability for high temperature turbines, a detailed understanding of the highly unsteady and three-dimensional flow through multi-stage turbines is necessary. Thus, unique data are required which quantify the unsteady three-dimensional flow through multi-stage turbine blade rows, including the effect of the film coolant flow. This requires experiments in appropriate research facilities in which complete flow field data, not only point measurements, are obtained and analyzed. Also, as design CFD codes do not account for unsteady flow effects, the next logical challenge and the current thrust in CFD code development is multiple-stage analyses that account for the interactions between neighboring blade rows.

  5. Operation performance and granule characterization of upflow anaerobic sludge blanket (UASB) reactor treating wastewater with starch as the sole carbon source.

    Science.gov (United States)

    Lu, Xueqin; Zhen, Guangyin; Estrada, Adriana Ledezma; Chen, Mo; Ni, Jialing; Hojo, Toshimasa; Kubota, Kengo; Li, Yu-You

    2015-03-01

    Long-term performance of a lab-scale UASB reactor treating starch wastewater was investigated under different hydraulic retention times (HRT). Successful start-up could be achieved after 15days' operation. The optimal HRT was 6h with organic loading rate (OLR) 4g COD/Ld at COD concentration 1000mg/L, attaining 81.1-98.7% total COD removal with methane production rate of 0.33L CH4/g CODremoved. Specific methane activity tests demonstrated that methane formation via H2-CO2 and acetate were the principal degradation pathways. Vertical characterizations revealed that main reactions including starch hydrolysis, acidification and methanogenesis occurred at the lower part of reactor ("main reaction zone"); comparatively, at the up converting acetate into methane predominated ("substrate-shortage zone"). Further reducing HRT to 3h caused volatile fatty acids accumulation, sludge floating and performance deterioration. Sludge floating was ascribed to the excess polysaccharides in extracellular polymeric substances (EPS). More efforts are required to overcome sludge floating-related issues. Copyright © 2015 Elsevier Ltd. All rights reserved.

  6. Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system

    International Nuclear Information System (INIS)

    Davidson, J.W.; Battat, M.E.

    1993-01-01

    Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies

  7. Development of simulator for remote handling system of ITER blanket

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakanhira, Masataka; Matsumoto, Yasuhiro; Shibanuma, K.

    2007-01-01

    The maintenance activity in the ITER has to be performed remotely because 14 MeV neutron caused by fusion reaction induces activation of structural material and emission of gamma ray. In general, it is one of the most critical issues to avoid collision between the remote maintenance system and in-vessel components. Therefore, the visual information in the vacuum vessel is required strongly to understand arrangement of these devices and components. However, there is a limitation of arrangement of viewing cameras in the vessel because of high intensity of gamma ray. It is expected that enough numbers of cameras and lights are not available because of arrangement restriction. Furthermore, visibility of the interested area such as the contacting part is frequently disturbed by the devices and components, thus it is difficult to recognize relative position between the devices and components only by visual information even if enough cameras and lights are equipped. From these reasons, the simulator to recognize the positions of each devices and components is indispensable for remote handling systems in fusion reactors. The authors have been developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robot simulation software ''ENVISION''. The simulator is connected to the control system of the manipulator which was developed as a part of the blanket maintenance system in the EDA and can reconstruct the positions of the manipulator and the blanket module using the position data of the motors through the LAN. In addition, it can provide virtual visual information, such as the connecting operation behind the blanket module with making the module transparent on the screen. It can be used also for checking the maintenance sequence before the actual operation. The developed simulator will be modified further adding other necessary functions and finally completed as a prototype of the actual simulator for the blanket remote handling system

  8. Present development status of EUROFER and ODS-EUROFER for application in blanket concepts

    Energy Technology Data Exchange (ETDEWEB)

    Lindau, R. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: rainer.lindau@imf.fzk.de; Moeslang, A. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany); Rieth, M. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany); Klimiankou, M. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany); Materna-Morris, E. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany); Alamo, A. [CEA-Saclay, SRMA/SMPX, 91191 Gif-sur-Yvette Cedex (France); Tavassoli, A.-A. F. [CEA-Saclay, SRMA/SMPX, 91191 Gif-sur-Yvette Cedex (France); Cayron, C. [CEA-Grenoble, DRT/DTEN/SMP/LS2M, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France); Lancha, A.-M. [CIEMAT, Avda. Complutense no. 22, 28040 Madrid (Spain); Fernandez, P. [CIEMAT, Avda. Complutense no. 22, 28040 Madrid (Spain); Baluc, N. [CRPP-EPFL, 5232 Villigen PSI (Switzerland); Schaeublin, R. [CRPP-EPFL, 5232 Villigen PSI (Switzerland); Diegele, E. [EFDA Close Support Unit, Boltzmannstr. 2, 85748 Garching (Germany); Filacchioni, G. [ENEA CR Casaccia, Via Anguillarese 301, 00100 S. Maria di Galeria, Rome (Italy); Rensman, J.W. [NRG, MM and I, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands); Schaaf, B. van der [NRG, MM and I, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands); Lucon, E. [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Dietz, W. [MECS, Schoenenborner Weg 15, 51789 Lindlar (Germany)

    2005-11-15

    Within the European Union, the two major breeding blanket concepts presently being developed are the helium cooled pebble bed (HCPB), and the helium cooled lithium lead (HCLL) blankets. For both concepts, different conceptual designs are being discussed with temperature windows in the range 250-550 deg. C for conservative approaches based on reduced activation ferritic-martensitic (RAFM) steels, and in the range 250-650 deg. C for more advanced versions, taking into account oxide dispersion strengthened (ODS) steels. As a final result of a systematic development of RAFM-steels in Europe, the 9% CrWVTa alloy EUROFER was specified and produced in an industrial scale with a variety of product forms. A large characterisation program is being performed including irradiation in materials test reactors between 60 and 450 deg. C ({<=}15 dpa), and in a fast breeder reactor at 330 deg. C up to 30 dpa. EUROFER is resistant to high temperature ageing, and the existing creep-rupture data ({approx}30,000 h, 450-600 deg. C) indicate long-term stability and predictability. The ODS variant of EUROFER shows superior tensile and creep properties compared to EUROFER. Applying a new production route has diminished the problem of lower ductility and inferior impact properties. A reliable joining technique for ODS and RAFM steels employing diffusion welding was successfully developed.

  9. Performance Evaluation and Robustness Testing of Advanced Oscilloscope Triggering Schemes

    Directory of Open Access Journals (Sweden)

    Shakeb A. KHAN

    2010-01-01

    Full Text Available In this paper, performance and robustness of two advanced oscilloscope triggering schemes is evaluated. The problem of time period measurement of complex waveforms can be solved using the algorithms, which utilize the associative memory network based weighted hamming distance (Whd and autocorrelation based techniques. Robustness of both the advanced techniques, are then evaluated by simulated addition of random noise of different levels to complex test signals waveforms, and minimum value of Whd (Whd min and peak value of coefficient of correlation(COCmax are computed over 10000 cycles of the selected test waveforms. The distance between mean value of second lowest value of Whd and Whd min and distance between second highest value of coefficient of correlation (COC and COC max are used as parameters to analyze the robustness of considered techniques. From the results, it is found that both the techniques are capable of producing trigger pulses efficiently; but correlation based technique is found to be better from robustness point of view.

  10. Development of Human Performance Analysis and Advanced HRA Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Won Dea; Park, Jin Kyun; Kim, Jae Whan; Kim, Seong Whan; Kim, Man Cheol; Ha, Je Joo

    2007-06-15

    The purpose of this project is to build a systematic framework that can evaluate the effect of human factors related problems on the safety of nuclear power plants (NPPs) as well as develop a technology that can be used to enhance human performance. The research goal of this project is twofold: (1) the development of a human performance database and a framework to enhance human performance, and (2) the analysis of human error with constructing technical basis for human reliability analysis. There are three kinds of main results of this study. The first result is the development of a human performance database, called OPERA-I/II (Operator Performance and Reliability Analysis, Part I and Part II). In addition, a standard communication protocol was developed based on OPERA to reduce human error caused from communication error in the phase of event diagnosis. Task complexity (TACOM) measure and the methodology of optimizing diagnosis procedures were also finalized during this research phase. The second main result is the development of a software, K-HRA, which is to support the standard HRA method. Finally, an advanced HRA method named as AGAPE-ET was developed by combining methods MDTA (misdiagnosis tree analysis technique) and K-HRA, which can be used to analyze EOC (errors of commission) and EOO (errors of ommission). These research results, such as OPERA-I/II, TACOM, a standard communication protocol, K-HRA and AGAPE-ET methods will be used to improve the quality of HRA and to enhance human performance in nuclear power plants.

  11. Development of Human Performance Analysis and Advanced HRA Methodology

    International Nuclear Information System (INIS)

    Jung, Won Dea; Park, Jin Kyun; Kim, Jae Whan; Kim, Seong Whan; Kim, Man Cheol; Ha, Je Joo

    2007-06-01

    The purpose of this project is to build a systematic framework that can evaluate the effect of human factors related problems on the safety of nuclear power plants (NPPs) as well as develop a technology that can be used to enhance human performance. The research goal of this project is twofold: (1) the development of a human performance database and a framework to enhance human performance, and (2) the analysis of human error with constructing technical basis for human reliability analysis. There are three kinds of main results of this study. The first result is the development of a human performance database, called OPERA-I/II (Operator Performance and Reliability Analysis, Part I and Part II). In addition, a standard communication protocol was developed based on OPERA to reduce human error caused from communication error in the phase of event diagnosis. Task complexity (TACOM) measure and the methodology of optimizing diagnosis procedures were also finalized during this research phase. The second main result is the development of a software, K-HRA, which is to support the standard HRA method. Finally, an advanced HRA method named as AGAPE-ET was developed by combining methods MDTA (misdiagnosis tree analysis technique) and K-HRA, which can be used to analyze EOC (errors of commission) and EOO (errors of ommission). These research results, such as OPERA-I/II, TACOM, a standard communication protocol, K-HRA and AGAPE-ET methods will be used to improve the quality of HRA and to enhance human performance in nuclear power plants

  12. An advanced probabilistic structural analysis method for implicit performance functions

    Science.gov (United States)

    Wu, Y.-T.; Millwater, H. R.; Cruse, T. A.

    1989-01-01

    In probabilistic structural analysis, the performance or response functions usually are implicitly defined and must be solved by numerical analysis methods such as finite element methods. In such cases, the most commonly used probabilistic analysis tool is the mean-based, second-moment method which provides only the first two statistical moments. This paper presents a generalized advanced mean value (AMV) method which is capable of establishing the distributions to provide additional information for reliability design. The method requires slightly more computations than the second-moment method but is highly efficient relative to the other alternative methods. In particular, the examples show that the AMV method can be used to solve problems involving non-monotonic functions that result in truncated distributions.

  13. IMPULSE---an advanced, high performance nuclear thermal propulsion system

    International Nuclear Information System (INIS)

    Petrosky, L.J.; Disney, R.K.; Mangus, J.D.; Gunn, S.A.; Zweig, H.R.

    1993-01-01

    IMPULSE is an advanced nuclear propulsion engine for future space missions based on a novel conical fuel. Fuel assemblies are formed by stacking a series of truncated (U, Zr)C cones with non-fueled lips. Hydrogen flows radially inward between the cones to a central plenum connected to a high performance bell nozzle. The reference IMPULSE engine rated at 75,000 lb thrust and 1800 MWt weighs 1360 kg and is 3.65 meters in height and 81 cm in diameter. Specific impulse is estimated to be 1000 for a 15 minute life at full power. If longer life times are required, the operating temperature can be reduced with a concomitant decrease in specific impulse. Advantages of this concept include: well defined coolant paths without outlet flow restrictions; redundant orificing; very low thermal gradients and hence, thermal stresses, across the fuel elements; and reduced thermal stresses because of the truncated conical shape of the fuel elements

  14. Biocatalysis conversion of methanol to methane in an upflow anaerobic sludge blanket (UASB) reactor: Long-term performance and inherent deficiencies.

    Science.gov (United States)

    Lu, Xueqin; Zhen, Guangyin; Chen, Mo; Kubota, Kengo; Li, Yu-You

    2015-12-01

    Long-term performance of methanol biocatalysis conversion in a lab-scale UASB reactor was evaluated. Properties of granules were traced to examine the impact of methanol on granulation. Methanolic wastewater could be stably treated during initial 240d with the highest biogas production rate of 18.6 ± 5.7 L/Ld at OLR 48 g-COD/Ld. However, the reactor subsequently showed severe granule disintegration, inducing granule washout and process upsets. Some steps (e.g. increasing influent Ca(2+) concentration, etc.) were taken to prevent rising dispersion, but no clear improvement was observed. Further characterizations in granules revealed that several biotic/abiotic factors all caused the dispersion: (1) depletion of extracellular polymeric substances (EPS) and imbalance of protein/polysaccharide ratio in EPS; (2) restricted formation of hard core and weak Ca-EPS bridge effect due to insufficient calcium supply; and (3) simplification of species with the methanol acclimation. More efforts are required to solve the technical deficiencies observed in methanolic wastewater treatment. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. Performance of a transmutation advanced device for sustainable energy application

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, C.; Rosales, J.; Garcia, L. [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Perez-Navarro, A.; Escriva, A. [Universidad Politecnica de Valencia, Valencia (Spain). Inst. de Ingenieria Energetica; Abanades, A. [Universidad Politecnica de Madrid (Spain). Grupo de Modelizacion de Sistemas Termoenergeticos

    2009-07-01

    Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

  16. Performance of a transmutation advanced device for sustainable energy application

    International Nuclear Information System (INIS)

    Garcia, C.; Rosales, J.; Garcia, L.; Perez-Navarro, A.; Escriva, A.; Abanades, A.

    2009-01-01

    Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

  17. MHD generator performance analysis for the Advanced Power Train study

    Science.gov (United States)

    Pian, C. C. P.; Hals, F. A.

    1984-01-01

    Comparative analyses of different MHD power train designs for early commercial MHD power plants were performed for plant sizes of 200, 500, and 1000 MWe. The work was conducted as part of the first phase of a planned three-phase program to formulate an MHD Advanced Power Train development program. This paper presents the results of the MHD generator design and part-load analyses. All of the MHD generator designs were based on burning of coal with oxygen-enriched air preheated to 1200 F. Sensitivities of the MHD generator design performance to variations in power plant size, coal type, oxygen enrichment level, combustor heat loss, channel length, and Mach number were investigated. Basd on these sensitivity analyses, together with the overall plant performance and cost-of-electricity analyses, as well as reliability and maintenance considerations, a recommended MHD generator design was selected for each of the three power plants. The generators for the 200 MWe and 500 MWe power plant sizes are supersonic designs. A subsonic generator design was selected for the 1000 MWe plant. Off-design analyses of part-load operation of the supersonic channel selected for the 200 MWe power plant were also conductd. The results showed that a relatively high overall net plant efficiency can be maintained during part-laod operation with a supersonic generator design.

  18. APT Blanket Detailed Bin Model Based on Initial Plate-Type Design -3D FLOWTRAN-TF Model

    International Nuclear Information System (INIS)

    Hamm, L.L.

    1998-01-01

    This report provides background information for a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report for the APT. This report gives a brief description of the FLOWTRAN-TF code which was used for detailed blanket bin modeling

  19. Corrosion performance of advanced structural materials in sodium.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Momozaki, Y.; Li, M.; Rink, D.L. (Nuclear Engineering Division)

    2012-05-16

    Test Facility, and Clinch River Breeder Reactor. Among the nonmetallic elements discussed, oxygen is deemed controllable and its concentration in sodium can be maintained in sodium for long reactor life by using cold-trap method. It was concluded that among the cold-trap and getter-trap methods, the use of cold trap is sufficient to achieve oxygen concentration of the order of 1 part per million. Under these oxygen conditions in sodium, the corrosion performance of structural materials such as austenitic stainless steels and ferritic steels will be acceptable at a maximum core outlet sodium temperature of {approx}550 C. In the current sodium compatibility studies, the oxygen concentration in sodium will be controlled and maintained at {approx}1 ppm by controlling the cold trap temperature. The oxygen concentration in sodium in the forced convection sodium loop will be controlled and monitored by maintaining the cold trap temperature in the range of 120-150 C, which would result in oxygen concentration in the range of 1-2 ppm. Uniaxial tensile specimens are being exposed to flowing sodium and will be retrieved and analyzed for corrosion and post-exposure tensile properties. Advanced materials for sodium exposure include austenitic alloy HT-UPS and ferritic-martensitic steels modified 9Cr-1Mo and NF616. Among the nonmetallic elements in sodium, carbon was assessed to have the most influence on structural materials since carbon, as an impurity, is not amenable to control and maintenance by any of the simple purification methods. The dynamic equilibrium value for carbon in sodium systems is dependent on several factors, details of which were discussed in the earlier report. The current sodium compatibility studies will examine the role of carbon concentration in sodium on the carburization-decarburization of advanced structural materials at temperatures up to 650 C. Carbon will be added to the sodium by exposure of carbon-filled iron tubes, which over time will enable

  20. Corrosion performance of advanced structural materials in sodium

    International Nuclear Information System (INIS)

    Natesan, K.; Momozaki, Y.; Li, M.; Rink, D.L.

    2012-01-01

    Clinch River Breeder Reactor. Among the nonmetallic elements discussed, oxygen is deemed controllable and its concentration in sodium can be maintained in sodium for long reactor life by using cold-trap method. It was concluded that among the cold-trap and getter-trap methods, the use of cold trap is sufficient to achieve oxygen concentration of the order of 1 part per million. Under these oxygen conditions in sodium, the corrosion performance of structural materials such as austenitic stainless steels and ferritic steels will be acceptable at a maximum core outlet sodium temperature of ∼550 C. In the current sodium compatibility studies, the oxygen concentration in sodium will be controlled and maintained at ∼1 ppm by controlling the cold trap temperature. The oxygen concentration in sodium in the forced convection sodium loop will be controlled and monitored by maintaining the cold trap temperature in the range of 120-150 C, which would result in oxygen concentration in the range of 1-2 ppm. Uniaxial tensile specimens are being exposed to flowing sodium and will be retrieved and analyzed for corrosion and post-exposure tensile properties. Advanced materials for sodium exposure include austenitic alloy HT-UPS and ferritic-martensitic steels modified 9Cr-1Mo and NF616. Among the nonmetallic elements in sodium, carbon was assessed to have the most influence on structural materials since carbon, as an impurity, is not amenable to control and maintenance by any of the simple purification methods. The dynamic equilibrium value for carbon in sodium systems is dependent on several factors, details of which were discussed in the earlier report. The current sodium compatibility studies will examine the role of carbon concentration in sodium on the carburization-decarburization of advanced structural materials at temperatures up to 650 C. Carbon will be added to the sodium by exposure of carbon-filled iron tubes, which over time will enable carbon to diffuse through iron and

  1. Materials issues in the design of the ITER first wall, blanket, and divertor

    International Nuclear Information System (INIS)

    Mattas, R.F.; Smith, D.L.; Wu, C.H.; Shatalov, G.

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

  2. Research on Elemental Technology of Advanced Nuclear Fuel Performance Verification

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Dong Uk; Jean, Sang Hwan; Koo, Min

    2003-04-01

    Most of current properties models and fuel performance models used in the performance evaluation codes are based on the in-pile data up to 33,000 MWd/MtU. Therefore, international experts are investigating the properties changes and developing advanced prediction models for high burn-up application. Current research is to develop high burn-up fission gas release model for the code and to support the code development activities by collecting data and models, reviewing/assessing the data and models together, and benchmarking the selected models against the appropriate in-pile data. For high burn-up applications, two stage two step fission gas release model is developed based on the real two diffusion process in the grain lattice and grain boundaries of the fission gases and the observation of accelerated release rate in the high burn-up. It is found that the prediction of this model is in excellent agreement with the in-pile measurement results, not only in the low burn-up but also in the high burn-up. This research is found that the importance of thermal conductivity of oxide fuel, especially in the high burn-up, is focused again. It is found that even the temperature dependent models differ from one to another and most of them overestimate the conductivity in the high burn-up. An in-pile data benchmarking of high LHGR fuel rod shows that the difference can reach 30%∼40%, which predicts 400 .deg. C lower than the real fuel centerline temperature. Recent models on the thermal expansion and heat capacity of oxide fuel are found to be well-defined. Irradiation swelling of the oxide fuel are now well-understood that in most cases in LWRs solid fission product swelling is dominant. Thus, the accumulation of in-pile data can enhance the accuracy of the model prediction, rather than theoretical modeling works. Thermo-physical properties of Zircaloy cladding are also well-defined and well-understood except the thermal expansion. However, it turns out that even the

  3. Advanced digital technology - improving nuclear power plant performance through maintainability

    International Nuclear Information System (INIS)

    Ford, J.L.; Senechal, R.R.; Altenhein, G.D.; Harvey, R.P.

    1998-01-01

    In today's energy sector there is ever increasing pressure on utilities to operate power plants at high capacity factors. To ensure nuclear power is competitive into the next century, it is imperative that strategic design improvements be made to enhance the performance of nuclear power plants. There are a number of factors that affect a nuclear power plant's performance; lifetime maintenance is one of the major contributors. The maturing of digital technology has afforded ABB the opportunity to make significant design improvements in the area of maintainability. In keeping with ABB's evolutionary advanced nuclear plant design approach, digital technology has systematically been incorporated into the control and protection systems of the most recent Korean nuclear units in operation and under construction. One example of this was the multi-functional design team approach that was utilized for the development of ABB's Digital Plant Protection System (DPPS). The design team consisted of engineers, maintenance technicians, procurement specialists and manufacturing personnel in order to provide a complete perspective on all facets of the design. The governing design goals of increased reliability and safety, simplicity of design, use of off-the-shelf products and reduced need for periodic surveillance testing were met with the selection of proven ABB-Advant Programmable Logic Controllers (PLCs) as the heart of the DPPS. The application of digital PLC technology allows operation for extended periods without requiring routine maintenance or re-calibration. A well documented commercial dedication program approved by the United States Nuclear Regulatory Commission (US NRC) as part of the System 80+ TM Advanced Light Water Reactor Design Certification Program, allowed the use of off-the shelf products in the design of the safety protection system. In addition, a number of mechanical and electrical improvements were made which support maintainability. The result is a DPPS

  4. Performance evaluation of cognitive radio in advanced metering infrastructure communication

    Science.gov (United States)

    Hiew, Yik-Kuan; Mohd Aripin, Norazizah; Din, Norashidah Md

    2016-03-01

    Smart grid is an intelligent electricity grid system. A reliable two-way communication system is required to transmit both critical and non-critical smart grid data. However, it is difficult to locate a huge chunk of dedicated spectrum for smart grid communications. Hence, cognitive radio based communication is applied. Cognitive radio allows smart grid users to access licensed spectrums opportunistically with the constraint of not causing harmful interference to licensed users. In this paper, a cognitive radio based smart grid communication framework is proposed. Smart grid framework consists of Home Area Network (HAN) and Advanced Metering Infrastructure (AMI), while AMI is made up of Neighborhood Area Network (NAN) and Wide Area Network (WAN). In this paper, the authors only report the findings for AMI communication. AMI is smart grid domain that comprises smart meters, data aggregator unit, and billing center. Meter data are collected by smart meters and transmitted to data aggregator unit by using cognitive 802.11 technique; data aggregator unit then relays the data to billing center using cognitive WiMAX and TV white space. The performance of cognitive radio in AMI communication is investigated using Network Simulator 2. Simulation results show that cognitive radio improves the latency and throughput performances of AMI. Besides, cognitive radio also improves spectrum utilization efficiency of WiMAX band from 5.92% to 9.24% and duty cycle of TV band from 6.6% to 10.77%.

  5. UCIMS: Advances in geotechnical construction and performance monitoring

    Directory of Open Access Journals (Sweden)

    Raphael Siebenmann

    2015-04-01

    Full Text Available The Crossrail project currently under construction in Central London has been described as “The Big Dig on Steroids”, obviously referencing the Central Artery/Tunnel project in Boston completed in 2007. To address the multiple demands for timely construction performance monitoring, Crossrail envisioned the underground construction information management system (UCIMS to monitor construction progress and structural health along the entire route, with a network of geotechnical instruments (i.e. slope inclinometers, extensometers, piezometers, etc. and tunnel boring machine (TBM position information. The UCIMS is a geospatially referenced relational database that was developed using an open source geographic information system (GIS that allowed all stakeholders near immediate feedback of construction performance. The purpose of this article is to provide a brief history of geotechnical and structural monitoring software, to describe the structure and operation of the UCIMS, and to demonstrate how the functionality afforded by this system provided the requisite feedback to the stakeholders. Examples will be given regarding how the data management and visualization concepts incorporated into the UCIMS advanced the geotechnical construction industry.

  6. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    Werme, L.

    1991-12-01

    A near-field performance evaluation of an advanced cold process canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie local-iced corrosion for the steel or copper canisters can be dismissed as a failure mechanism; The evaluation of the effects of processed outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. For completeness also evaluation of post-failure behaviour was carried out. Analyses were focussed on low probability phenomena from faults in canisters. Some items were identified where further research is justified in order to increase knowledge of the phenomena and thus strengthen the confidence of safety margins. However, it can be concluded that the risks of these scenarios can be judged to be acceptable. This is due to the fact that firstly, the probability of occurrence of most of these scenarios can be controlled to a large extent through technical measures. Secondly, these analyses indicated that the consequences would not be severe

  7. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    Werme, L.

    1990-09-01

    A near-field performance evaluation of an Advanced Cold Process Canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. The canister design was originally proposed by TVO. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. Throughout the analysis, present day underground conditions has been assumed to persist during the service life of the canister. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie localized corrosion for the steel or copper canisters can be dismissed as a failure mechanism. The evaluation of the effects of processes outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. This factor will ensure the safety of the concept. (orig.)

  8. CONCERNING THE ADVANCED SCIENCE IN HIGH PERFORMANCE SPORT

    Directory of Open Access Journals (Sweden)

    Gagea Adrian

    2010-04-01

    Full Text Available The advanced sciences are based on the most recent huge increasing of technology and on interdisciplinary commencement of great interest topics, as top sport is considering. The main problem in top sport seems to be the obtaining high sport’s performance in as short as possible time, having great efficiency and minimum risks.The cell-engineering domain, in which the author of this paper has a modest contribution, is a means of genetic control for human performance, including sport, gene expression, molecular interactions within the cell, intracellular signalling, cell mechanics and motility etc.The domain of Psyche, of controlling feelings and manifestations, is also, on the focus of top sport interest, especially for the reason that, from inside of this domain, is feasible to accede at the biological reserves unavoidable in normal conditions, but avoidable in emergency or surviving situations. The new knowledge about energetic metabolism, about the rotation of ATP molecules, or coming out from scientifically experiments of association of nutrients or of reconsidering the recovery stimulants after effort, are providing, also, very useful information for top sport practitioners.It is not to disregard the contribution of the new information about the human physical limits, biomechanics, tactics of doing and controls the physical effort by means of sensorial biofeedback or theperformance’s advantages coming from new high-minded techniques and materials of sport accessories

  9. Transmutation blanket design for a Tokamak system

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A. Fortini; Costa, Antonella L.

    2011-01-01

    Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

  10. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

    Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-10-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

  11. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

    1998-01-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

  12. Beryllium R&D for blanket application

    Science.gov (United States)

    Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

    1998-10-01

    The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

  13. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1978-01-01

    The thermal and mechanical aspects of hybrid reactor blanket design considerations are discussed. This paper is intended as a companion to that of J. D. Lee of Lawrence Livermore Laboratory on the nuclear aspects of hybrid reactor blanket design. The major design characteristics of hybrid reactor blankets are discussed with emphasis on the areas of difference between hybrid reactors and standard fusion or fission reactors. Specific examples are used to illustrate the design tradeoffs and choices that must be made in hybrid reactor design. These examples are drawn from the work on the Mirror Hybrid Reactor

  14. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Young, J.R.

    1975-01-01

    Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

  15. RAMI analysis for DEMO HCPB blanket concept cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

    2015-10-15

    Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

  16. Heat Loads Due To Small Penetrations In Multilayer Insulation Blankets

    Science.gov (United States)

    Johnson, W. L.; Heckle, K. W.; E Fesmire, J.

    2017-12-01

    The main penetrations (supports and piping) through multilayer insulation systems for cryogenic tanks have been previously addressed by heat flow measurements. Smaller penetrations due to fasteners and attachments are now experimentally investigated. The use of small pins or plastic garment tag fasteners to ease the handling and construction of multilayer insulation (MLI) blankets goes back many years. While it has long been understood that penetrations and other discontinuities degrade the performance of the MLI blanket, quantification of this degradation has generally been lumped into gross performance multipliers (often called degradation factors or scale factors). Small penetrations contribute both solid conduction and radiation heat transfer paths through the blanket. The conduction is down the stem of the structural element itself while the radiation is through the hole formed during installation of the pin or fastener. Analytical models were developed in conjunction with MLI perforation theory and Fourier’s Law. Results of the analytical models are compared to experimental testing performed on a 10 layer MLI blanket with approximately 50 small plastic pins penetrating the test specimen. The pins were installed at ∼76-mm spacing inches in both directions to minimize the compounding of thermal effects due to localized compression or lateral heat transfer. The testing was performed using a liquid nitrogen boil-off calorimeter (Cryostat-100) with the standard boundary temperatures of 293 K and 78 K. Results show that the added radiation through the holes is much more significant than the conduction down the fastener. The results are shown to be in agreement with radiation theory for perforated films.

  17. Probabilistic safety assessment of the dual-cooled waste transmutation blanket for the FDS-I

    International Nuclear Information System (INIS)

    Hu, L.; Wu, Y.

    2006-01-01

    The subcritical dual-cooled waste transmutation (DWT) blanket is one of the key components of fusion-driven subcritical system (FDS-I). The probabilistic safety assessment (PSA) can provide valuable information on safety characteristics of FDS-I to give recommendations for the optimization of the blanket concepts and the improvement of the design. Event tree method has been adopted to probabilistically analyze the safety of the DWT blanket for FDS-I using the home-developed PSA code RiskA. The blanket melting frequency has been calculated and compared with the core melting frequencies of PWRs and a fast reactor. Sensitivity analysis of the safety systems has been performed. The results show that the current preliminary design of the FDS-I is very attractive in safety

  18. Self-cooled blanket concepts using Pb-17Li as liquid breeder and coolant

    International Nuclear Information System (INIS)

    Malang, S.; Deckers, H.; Fischer, U.; John, H.; Meyder, R.; Norajitra, P.; Reimann, J.; Reiser, H.; Rust, K.

    1991-01-01

    A blanket design concept using Pb-17Li eutectic alloy as both breeder material and coolant is described. Such a self-cooled blanket for the boundary conditions of a DEMO-reactor is under development at the Kernforschungszentrum Karlsruhe (KfK) in the frame of the European blanket development program. Results of investigations in the areas of design, neutronics, magneto-hydrodynamics, thermo-mechanics, ancillary loop systems, and safety are reported. Based on recent progress, it can be concluded that the boundary conditions of a DEMO-reactor can be met, tritium self-sufficiency can be obtained without using beryllium as an additional neutron multiplier, and tritium inventory and permeation are acceptably low. However, to complete judge the feasibility of the proposed concept, further studies are necessary to obtain a better understanding of the magneto-hydrodynamic phenomena and their effects on the thermal-hydraulic performance of a fusion reactor blanket. (orig.)

  19. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

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

    1976-01-01

    The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)

  20. Blanket design for imploding liner systems

    International Nuclear Information System (INIS)

    Schaffer, M. J.

    1980-01-01

    The blanket design comprises hot, molten, rotating liquid vortex systems suitable for rapidly compressing confined plasmas, in which stratified immiscible liquid layers having successively greater mass densities outwardly of the axis of rotation are provided

  1. Performance of advanced oxide fuel pins in EBR-II

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Jensen, S.M.; Hales, J.W.; Karnesky, R.A.; Makenas, B.J.

    1986-05-01

    The effects of design and operating parameters on mixed-oxide fuel pin irradiation performance were established for the Hanford Engineering Development Laboratory (HEDL) advanced oxide EBR-II test series. Fourteen fuel pins breached in-reactor with reference 316 SS cladding. Seven of the breaches are attributed to FCMI. Of the remaining seven breached pins, three are attributed to local cladding over-temperatures similar to the breach mechanism for the reference oxide pins irradiated in EBR-II. FCCI was found to be a contributing factor in two high burnup, i.e., 11.7 at. % breaches. The remaining two breaches were attributed to mechanical interaction of UO 2 fuel and fission products accumulated in the lower cladding insulator gap, and a loss of cladding ductility possibly due to liquid metal embrittlement. Fuel smear density appears to have the most significant impact on lifetime. Quantitative evaluations of cladding diameter increases attributed to FCMI, established fuel smear density, burnup, and cladding thickness-to-diameter ratio as the major parameters influencing the extent of cladding strain

  2. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    Ranta-Puska, K.; Pihlatie, M.

    2001-01-01

    The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85 Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

  3. The fusion blanket program at Chalk River

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-03-01

    Work on the Fusion Blanket Program commenced at Chalk River in 1984 June. Co-funded by Canadian Fusion Fuels Technology Project and Atomic Energy of Canada Limited, the Program utilizes Chalk River expertise in instrumented irradiation testing, ceramics, tritium technology, materials testing and compound chemistry. This paper gives highlights of studies to date on lithium-based ceramics, leading contenders for the fusion blanket

  4. HIGH PERFORMANCE ADVANCED TOKAMAK REGIMES FOR NEXT-STEP EXPERIMENTS

    International Nuclear Information System (INIS)

    GREENFIELD, C.M.; MURAKAMI, M.; FERRON, J.R.; WADE, M.R.; LUCE, T.C.; PETTY, C.C.; MENARD, J.E; PETRIE, T.W.; ALLEN, S.L.; BURRELL, K.H.; CASPER, T.A; DeBOO, J.C.; DOYLE, E.J.; GAROFALO, A.M; GORELOV, Y.A; GROEBNER, R.J.; HOBIRK, J.; HYATT, A.W; JAYAKUMAR, R.J; KESSEL, C.E; LA HAYE, R.J; JACKSON, G.L; LOHR, J.; MAKOWSKI, M.A.; PINSKER, R.I.; POLITZER, P.A.; PRATER, R.; STRAIT, E.J.; TAYLOR, T.S; WEST, W.P.

    2003-01-01

    OAK-B135 Advanced Tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic (MHD) stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half-radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding (ELMing) H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results of recent experiment and simulation efforts

  5. Experimental program for the Fast Breeder Blanket Facility, FBBF

    International Nuclear Information System (INIS)

    Ott, K.O.; Clikeman, F.M.; Johnson, R.H.; Borg, R.C.

    1976-01-01

    The work performed in the reporting period was primarily concerned with the development of the experimental program (Task A) and with the pre-analysis of future loadings and the impact upon the permanent loading of the two converter regions, which contain 4.8 percent enriched UO 2 rods. It appears necessary that a neutron poison (B 4 C) be placed in the converter (transformer) regions in order to hold, also for future loadings, the k/sub eff/ of a hypothetically flooded FBBF well below 1. Since it is planned to use the same welded converter regions for all experiments, the required B 4 C loading needs to be determined prior to the first blanket loading. Further the equipment needs have been identified (Task D), the 252 Cf-source has been requested on a loan basis (Task E). First discussions with ANL on blanket experiments have been initiated

  6. The State of the Art Report on the Development and Manufacturing Technology of Test Blanket Module

    International Nuclear Information System (INIS)

    Lee, J. S.; Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Kim, K. H.

    2006-07-01

    The main objective of the present R and D on breeder blanket is the development of test blanket modules (TBMs) to be installed and tested in International Thermonuclear Experimental Reactor (ITER). In the program of the blanket development, a blanket module test in the ITER is scheduled from the beginning of the ITER operation, and the performance test of TBM in ITER is the most important milestone for the development of the DEMO blanket. The fabrication of TBMs has been required to test the basic performance of the DEMO blanket, i.e., tritium production/recovery, high-grade heat generation and radiation shielding. Therefore, the integration of the TBM systems into ITER has been investigated with the aim to check the safety, reliability and compatibility under nuclear fusion state. For this reason, in the Test Blanket Working Group (TBWG) as an activity of the International Energy Association (IEA), a variety of ITER TBMs have been proposed and investigated by each party: helium-cooled ceramic (WSG-1), helium-cooled LiPb (WSG-2), water-cooled ceramic (WSG-3), self-cooled lithium (WSG-4) and self-cooled molten salt (WSG-5) blanket systems. Because we are still deficient in investigation of TBM development, the need of development became pressing. In this report, for the development of TBM sub-module and mock-up, it is necessary to analyze and examine the state of the art on the development of manufacturing technology of TBM in other countries. And we will be applied as basic data to establish a manufacturing technology

  7. Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

    2016-11-15

    Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

  8. The EC conceptual design proposal of a water-cooled convertible blanket for ITER

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Baraer, L.; Bielak, B.; Raepsaet, X.; Salavy, J.F.; Sedano, L.; Szczepanski, J.; Quintric-Bossy, J.; Severi, Y.

    1993-01-01

    For several years the EC laboratories have developed breeding blankets for DEMO. From this experience, it has been derived a proposal of tritium breeding blanket for the Extended Performance Phase (EPP) of ITER. The general basic ideas are the following: (i) the switch from the shielding blanket used during the BPP to the breeding blanket for the EPP should not require segments replacement ('convertible' blanket): (ii) its use should not have significant impact on the Basic Performance Phase (BPP); (iii) design and used materials should assure good safety standards and acceptable public perception; (iv) the blanket coolant should be compatible with the coolant required in the high heat-flux components (e.g. divertor, etc.; (v) the required R and D should fit with the ITER time schedule; (vi) the blanket should be able to withstand large power excursions and to accept long downtimes. The proposed design consists of a water-cooled liquid metal blanket, using the eutectic Pb-17Li during the EPP and a non-breeding Pb-alloy (Pb-18Mg or Pb-50Bi) during the BPP. Each segment is basically formed by a box containing the alloy, cooled by an array of poloidal hairpin-type cooling tubes and reinforced by toroidal and radial stiffeners. The coolant tubes are double-walled tubes allowing leak detections. The selected First Wall (FW) is a toroidally-drilled steel plate with brazed water-cooling U-tube. The structural material is austenitic stainless steel (316L(N)) which limits the maximum acceptable neutron fluence to about 1 MWa/m 2 . The advantages of using other structural materials requiring longer leadtimes, such as ferritic/martensitic steels, are also briefly discussed

  9. Workshop on cold-blanket research

    International Nuclear Information System (INIS)

    1977-05-01

    The objective of the workshop was to identify and discuss cold-plasma blanket systems. In order to minimize the bombardment of the walls by hot neutrals the plasma should be impermeable. This requires a density edge-thickness product of nΔ > 10 15 cm -2 . An impermeable cold plasma-gas blanket surrounding a hot plasma core reduces the plasma wall/limiter interaction. Accumulation of impurities in this blanket can be expected. Fuelling from a blanket may be possible as shown by experimental results, though not fully explained by classical transport of neutrals. Refuelling of a reacting plasma had to be ensured by inward diffusion. Experimental studies of a cold impermeable plasma have been done on the tokamak-like Ringboog device. Simulation calculations for the next generation of large tokamaks using a particular transport model, indicate that the plasma edge profile can be controlled to reduce the production of sputtered impurities to an acceptable level. Impurity control requires a small fraction of the radial space to accomodate the cold-plasma layer. The problem of exhaust is, however, more complicated. If the cold-blanket scheme works as predicted in the model calculations, then α-particles generated by fusion will be transported to the cold outside layer. The Communities' experimental programme of research has been discussed in terms of the tokamaks which are available and planned. Two options present themselves for the continuation of cold-blanket research

  10. Liquid metal blanket module testing and design for ITER/TIBER II

    International Nuclear Information System (INIS)

    Mattas, R.F.; Cha, Y.; Finn, P.A.; Majumdar, S.; Picologlou, B.; Stevens, H.; Turner, L.

    1988-05-01

    A major goal for ITER is the testing of nuclear components to demonstrate the integrated performance of the most attractive concepts that can lead to a commercial fusion reactor. As part of the ITER/TIBER II study, the test program and design of test models were examined for a number of blanket concepts. The work at Argonne National Laboratory focused on self-cooled liquid metal blankets. A test program for liquid metal blankets was developed based upon the ITER/TIBER II operating schedule and the specific data needs to resolve the key issues for liquid metals. Testing can begin early in reactor operation with liquid metal MHD tests to confirm predictive capability. Combined heat transfer/MHD tests can be performed during initial plasma operation. After acceptable heat transfer performance is verified, tests to determine the integrated high temperature performance in a neutron environment can begin. During the high availability phase operation, long term performance and reliability tests will be performed. It is envisioned that a companion test program will be conducted outside ITER to determine behavior under severe accident conditions and upper performance limits. A detailed design of a liquid metal test module and auxiliary equipment was also developed. The module followed the design of the TPSS blanket. Detailed analysis of the heat transfer and tritium systems were performed, and the overall layout of the systems was determined. In general, the blanket module appears to be capable of addressing most of the testing needs. 8 refs., 27 figs., 11 tabs

  11. Liquid lithium blanket processing studies

    International Nuclear Information System (INIS)

    Talbot, J.B.; Clinton, S.D.

    1979-01-01

    The sorption of tritium on yttrium from flowing molten lithium and the subsequent release of tritium from yttrium for regeneration of the metal sorbent were investigated to evaluate the feasibility of such a tritium-recovery process for a fusion reactor blanket of liquid lithium. In initial experiments with the forced convection loop, yttrium samples were contacted with lithium at 300 0 C. A mass transfer coefficient of 2.5 x 10 - cm/sec, which is more than an order of magnitude less than the value measured in earlier static experiments, was determined for the flowing lithium system. Rates of tritium release from yttrium samples were measured to evaluate possible thermal regeneration of the sorbent. Values for diffusion coefficients at 505, 800, and 900 0 C were estimated to be 1.1 x 10 -13 , 4.9 x 10 -12 , and 9.3 x 10 -10 cm 2 /sec, respectively. Tritium release from yttrium was investigated at higher temperatures and with hydrogen added to the argon sweep gas to provide a reducing atmosphere

  12. Effects of fertile blanket on 600 MWth gas-cooled fast reactors: reactor and fuel cycle model

    International Nuclear Information System (INIS)

    Choi, Hang Bok

    2002-07-01

    A physics study has been performed to search for an optimum size of blanket for a 600 MWth gas-cooled fast reactor under fixed fuel and core specifications. The variables considered in this study are the reflector material, reflector thickness and blanket volume. The parametric calculations have shown that a positive breeding gain can be obtained by deploying 8 m 3 natural uranium blanket on the axial and radial boundaries of the core, surrounded by 40 cm Zr 3 Si 2 reflector. However the blanket core has disadvantages compared to the no-blanket core from the viewpoints of fuel fabrication cost and proliferation risk. On the other hand, the no-blanket core has large uncertainties in the possibility of achieving a positive breeding gain. Therefore further studies are recommended for the no-blanket option to improve the breeding gain and achieve a fissile self-sufficient fuel cycle, which is also proliferation-resistant. As an alternative, the blanket option can be considered, that ensures a positive breeding gain

  13. Preconceptual design of a packed fluidized bed blanket for a fission suppressed thorium-fueled CTHR

    International Nuclear Information System (INIS)

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

    1981-01-01

    This paper describes a thorium-fueled PFB blanket concept for a Commercial Tokamak Hybrid Reactor. A preliminary mechanical concept is presented and the results of neutronics, thermal-hydraulics and economics analyses are discussed. Futher work needed to design and advance the concept is recommended

  14. Conceptual design and testing strategy of a dual functional lithium-lead test blanket module in ITER and EAST

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

    A dual functional lithium-lead (DFLL) test blanket module (TBM) concept has been proposed for testing in the International Thermonuclear Experimental Reactor (ITER) and the Experimental Advanced Superconducting Tokamak (EAST) in China to demonstrate the technologies of the liquid lithium-lead breeder blankets with emphasis on the balance between the risks and the potential attractiveness of blanket technology development. The design of DFLL-TBM concept has the flexibility of testing both the helium-cooled quasi-static lithium-lead (SLL) blanket concept and the He/PbLi dual-cooled lithium-lead (DLL) blanket concept. This paper presents an effective testing strategy proposed to achieve the testing target of SLL and DLL DEMO blankets relevant conditions, which includes three parts: materials R and D and small-scale out-of-pile mockups testing in loops, middle-scale TBMs pre-testing in EAST and full-scale consecutive TBMs testing corresponding to different operation phases of ITER during the first 10 years. The design of the DFLL-TBM concept and the testing strategy ability to test TBMs for both blanket concepts in sequence and or in parallel for both ITER and EAST are discussed

  15. High Performance Thrusters for Advanced Green Monopropellants, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The development of an advanced green monopropellant propulsion system could have significant benefits to a wide range of NASA space missions, from deep space...

  16. Hydraulic performance of a multistage array of advanced centrifugal contactors

    International Nuclear Information System (INIS)

    Hodges, M.E.

    1984-01-01

    The hydraulic characteristics of an advanced design centrifugal contactor array have been determined at the Savannah River Laboratory (SRL). The advanced design utilizes couette mixing (Taylor vortices) in the annulus between the rotating and stationary bowls. Excellent phase separation over a wide range of flow conditions was obtained. Interfaces within an entire eight-stage array were controlled with a single weir air pressure. 2 references, 5 figures

  17. Adaptation of the HCPB DEMO TBM as breeding blanket for ITER : Neutronic and thermal analyses

    International Nuclear Information System (INIS)

    Aquaro, D.; Morellini, D.; Cerullo, N.

    2006-01-01

    Two breeding blanket are presently developed in Europe for the DEMO reactor: the first one, the Helium Cooled Lithium Lead (HCLL) uses a liquid breeder while the other , the Helium Cooled Pebble Bed (HCPB), uses a solid breeder in form of pebble bed. The modules of these blankets, called Test Blanket Modules (TBM) will be located in correspondence of the equatorial ports of ITER in order to be tested. ITER FEAT was designed with shielding blankets, therefore in the final stage of the experiment, in the foreseen tritium -deuterium operation phase, the tritium will be supplied to the reactor and not produced inside it. Since the production of tritium is of main importance for the feasibility of a nuclear fusion reactor, perhaps in the ITER final stage, the shielding blanket could be substituted by means of a breeding blanket. The geometry and composition of this breeding blanket would be, of course, similar to that of TBM which demonstrated to have the best performances. This paper illustrates a neutronic and thermal analysis of an hypothetical triziogen blanket for ITER FEAT made similar to a HCPB test module. The main aims of the performed analyses are to determine the Tritium Breeding Ratio (TBR) considering different solid breeders (Li 4 SiO 4 and Li 2 TiO 3 ) with different enrichment in 6 Li and different structural materials (a 9%CRWVTa reduced activation ferritic martensitic steel (EUROFER) or ceramic matrix composites like SiCf/SiC). The breeding blanket design is compared considering the highest value of TBR and the verification of the temperature constraints ( 550 o C for the steel, 950 o C for the breeder and 650 o C for the Beryllium). The neutronic analyses have been performed by means of MCNP-4C code and the thermal analyses using the MSC-MARC code. A TBR about equal 1 was obtained with a SiCf/SiC structural material and a Li 4 SiO 4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give

  18. Electrical connectors for blanket modules in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I., E-mail: poddubnyyii@nikiet.ru [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Khomiakov, S.; Kolganov, V. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Sadakov, S.; Calcagno, B.; Chappuis, Ph.; Roccella, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Danilov, I.; Leshukov, A.; Strebkov, Y. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Ulrickson, M. [Sandia National Laboratories MS-1129, PO Box 5800, Albuquerque, NM 87185 (United States)

    2014-10-15

    Highlights: • Analysis of static and cyclic strength for L-shaped and Z-shaped ES has been performed. • Analysis results do show that for L-shaped ES static and cyclic strength criteria are not satisfied. • Static and cyclic strength criteria are met well by ES with Z-shaped elastic elements. • ES with Z-shaped elastic elements has been adopted as a new baseline design for ITER. - Abstract: Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak. The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was

  19. New concepts for controlled fusion reactor blanket design

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Avci, H.; El-Maghrabi, M.

    1975-01-01

    Several new concepts for fusion reactor blanket design based on the idea of shifting, or tailoring, the neutron spectrum incident on the first structural wall are presented. The spectral shifter is a nonstructural element which can be made of graphite, silicon carbide, or three dimensionally woven carbon fibers (and containing other materials as appropriate) placed between the neutron source and the first structural wall. The softened neutron spectrum incident on the structural components leads to lower gas production and atom displacement rates than in more standard fusion blanket designs. In turn, this results in longer anticipated lifetimes for the structural materials and can significantly reduce radioactivity and afterheat levels. In addition, the neutron spectrum in the first structural wall can be made to approach the flux shape in fast breeder reactors. Such spectral softening means that existing radiation facilities may be more profitably used to provide relevant materials radiation damage data for the structural materials in these fusion blanket designs. This general class of blanket concepts are referred to as internal spectral shifter and energy converter, or ISSEC concepts. These specific design concepts fall into three main categories: ISSEC/EB concepts based on utilizing existing designs which breed tritium behind the first structural wall; ISSEC/IB concepts based on breeding tritium inside the first vacuum wall; and ISSEC/Bu concepts based on using boron, carbon, and perhaps, beryllium to obtain an energy multiplier and converter design that does not attempt to breed tritium or utilize lithium. The detailed analyses relate specifically to the nuclear performance of ISSEC systems and to a discussion of materials radiation damage problems in the structural material.(U.S.)

  20. Space environment durability of beta cloth in LDEF thermal blankets

    Science.gov (United States)

    Linton, Roger C.; Whitaker, Ann F.; Finckenor, Miria M.

    1993-01-01

    Beta cloth performance for use on long-term space vehicles such as Space Station Freedom (S.S. Freedom) requires resistance to the degrading effects of the space environment. The major issues are retention of thermal insulating properties through maintaining optical properties, preserving mechanical integrity, and generating minimal particulates for contamination-sensitive spacecraft surfaces and payloads. The longest in-flight test of beta cloth's durability was on the Long Duration Exposure Facility (LDEF), where it was exposed to the space environment for 68 months. The LDEF contained 57 experiments which further defined the space environment and its effects on spacecraft materials. It was deployed into low-Earth orbit (LEO) in Apr. 1984 and retrieved Jan. 1990 by the space shuttle. Among the 10,000 plus material constituents and samples onboard were thermal control blankets of multilayer insulation with a beta cloth outer cover and Velcro attachments. These blankets were exposed to hard vacuum, thermal cycling, charged particles, meteoroid/debris impacts, ultraviolet (UV) radiation, and atomic oxygen (AO). Of these space environmental exposure elements, AO appears to have had the greatest effect on the beta cloth. The beta cloth analyzed in this report came from the MSFC Experiment S1005 (Transverse Flat-Plate Heat Pipe) tray oriented approximately 22 deg from the leading edge vector of the LDEF satellite. The location of the tray on LDEF and the placement of the beta cloth thermal blankets are shown. The specific space environment exposure conditions for this material are listed.

  1. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

    Schuller, M.J.

    1985-01-01

    Numerical and experimental studies were carried out to assess the feasibility of using an aqueous slurry, with lithium in its solid component, to meet the tritium breeding, cooling, and shielding requirements of a controlled thermonuclear reactor (CTR). The numerical studies were designed to demonstrate the theoretical ability of a conceptual slurry blanket to breed adequate tritium to sustain the CTR. The experimental studies were designed to show that the tritium retention characteristics of likely solid components for the slurry were conducive to adequate tritium recovery without the need for isotopic separation. The numerical portion of this work consisted in part of using ANISN, a one-dimensional finite difference neutron transport code, to model the neutronic performance of the slurry blanket concept. The parameters governing tritium production and retention in a slurry were computed and used to modify the results of the ANISN computer runs. The numerical work demonstrated that the slurry blanket was only marginally capable of breeding sufficient tritium without the aid of a neutron multiplying region. The experimental portion of this work consisted of several neutron irradiation experiments, which were designed to determine the retention abilities of LiF particles

  2. Neutronic investigation and activation calculation for CFETR HCCB blankets

    Science.gov (United States)

    Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

    2017-12-01

    The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10-4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

  3. Progress in fusion reactors blanket analysis and evaluation at CEA

    International Nuclear Information System (INIS)

    Proust, E.; Gervaise, F.; Carre, F.; Chevereau, G.; Doutriaux, D.

    1986-09-01

    In the frame of the recent CEA studies aiming at the development, evaluation and comparison of solid breeder blanket concepts in view of their adaptation to NET, the evaluation of specific questions related to the first wall design, the present paper examines first the performances of a helium cooled toroidal blanket design for NET, based on innovative Beryllium/Ceramics breeder rod elements. Neutronic and thermo-mechanical optimisation converges on a concept featured by a breeding capability in excess of 1.2, a reasonnable pumping power of 1% and a narrow breeder temperature range (470+-30 deg C of the breeder), the latter being largely independent of the power level. This design proves naturally adapted to ceramic breeder assigned to very strict working conditions, and provides for any change in the thermal and heat transfer characteristics over the blanket lifetime. The final section of the paper is devoted to the evaluation of the heat load poloidal distribution and to the irradiation effects on first wall structural materials

  4. Thermo-mechanical characterization of ceramic pebbles for breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lo Frano, Rosa, E-mail: rosa.lofrano@ing.unipi.it; Aquaro, Donato; Scaletti, Luca

    2016-11-01

    Highlights: • Experimental activities to characterize the Li{sub 4}SiO{sub 4}. • Compression tests of pebbles. • Experimental evaluation of thermal conductivity of pebbles bed at different temperatures. • Experimental test with/without compression load. - Abstract: An open issue for fusion power reactor is to design a suitable breeding blanket capable to produce the necessary quantity of the tritium and to transfer the energy of the nuclear fusion reaction to the coolant. The envisaged solution called Helium-Cooled Pebble Bed (HCPB) breeding blanket foresees the use of lithium orthosilicate (Li{sub 4}SiO{sub 4}) or lithium metatitanate (Li{sub 2}TiO{sub 3}) pebble beds. The thermal mechanical properties of the candidate pebble bed materials are presently extensively investigated because they are critical for the feasibility and performances of the numerous conceptual designs which use a solid breeder. This study is aimed at the investigation of mechanical properties of the lithium orthosilicate and at the characterization of the main chemical, physical and thermo-mechanical properties taking into account the production technology. In doing that at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa adequate experiments were carried out. The obtained results may contribute to characterize the material of the pebbles and to optimize the design of the envisaged fusion breeding blankets.

  5. Sensisivity and Uncertainty analysis for the Tritium Breeding Ratio of a DEMO Fusion reactor with a Helium cooled pebble bed blanket

    OpenAIRE

    Nunnenmann, Elena; Fischer, Ulrich; Stieglitz, Robert

    2016-01-01

    An uncertainty analysis was performed for the tritium breeding ratio (TBR) of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB), currently under development in the European Power Plant Physics and Technology (PPPT) programme for a fusion power demonstration reactor (DEMO). A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design c...

  6. Analysis instruments for the performance of Advanced Practice Nursing.

    Science.gov (United States)

    Sevilla-Guerra, Sonia; Zabalegui, Adelaida

    2017-11-29

    Advanced Practice Nursing has been a reality in the international context for several decades and recently new nursing profiles have been developed in Spain as well that follow this model. The consolidation of these advanced practice roles has also led to of the creation of tools that attempt to define and evaluate their functions. This study aims to identify and explore the existing instruments that enable the domains of Advanced Practice Nursing to be defined. A review of existing international questionnaires and instruments was undertaken, including an analysis of the design process, the domains/dimensions defined, the main results and an exploration of clinimetric properties. Seven studies were analysed but not all proved to be valid, stable or reliable tools. One included tool was able to differentiate between the functions of the general nurse and the advanced practice nurse by the level of activities undertaken within the five domains described. These tools are necessary to evaluate the scope of advanced practice in new nursing roles that correspond to other international models of competencies and practice domains. Copyright © 2017 Elsevier España, S.L.U. All rights reserved.

  7. Passive Solar still: Recent advancement in design and related Performance.

    Science.gov (United States)

    Awasthi, Anuradha; Kumari, Kanchan; Panchal, Hitesh; Sathyamurthy, Ravishankar

    2018-05-31

    Present review paper mainly focuses on different varieties of solar stills and highlights mostly the passive solar still with advanced modifications in the design and development of material, single and multi-effect solar still with augmentation of different materials, energy absorbing, insulators, mechanisms of heat and mass transfer to improve the loss of heat and enhance the productivity of solar still. The cost-benefit analysis along with the progressive advancement for solar stills is the major highlights of this review. To increase the output of solar still nowadays, applications of advance modifications is one of the promising tools, and it is anticipated that shortly more vigor will be added in this area with the modifications in designs of solar stills.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  9. Liquid metal flows in insulating elements of self-cooled blankets

    International Nuclear Information System (INIS)

    Molokov, S.

    1995-01-01

    Liquid metal flows in insulating rectangular ducts in strong magnetic fields are considered with reference to poloidal concepts of self-cooled blankets. Although the major part of the flow in poloidal blanket concepts is close to being fully developed, manifolds, expansions, contractions, elbows, etc., which are necessary elements in blanket designs, cause three-dimensional effects. The present investigation demonstrates the flow pattern in basic insulating geometries for actual and more advanced liquid metal blanket concepts and discusses the ways to avoid pressure losses caused by flow redistribution. Flows in several geometries, such as symmetric and non-symmetric 180 turns with and without manifolds, sharp and linear expansions with and without manifolds, etc., have been considered. They demonstrate the attractiveness of poloidal concepts of liquid metal blankets, since they guarantee uniform conditions for heat transfer. If changes in the duct cross-section occur in the plane perpendicular to the magnetic field (ideally a coolant should always flow in the radial-poloidal plane), the disturbances are local and the slug velocity profile is reached roughly at a distance equivalent to one duct width from the manifolds, expansions, etc. The effects of inertia in these flows are unimportant for the determination of the pressure drop and velocity profiles in the core of the flow but may favour heat transfer characteristics via instabilities and strongly anisotropic turbulence. (orig.)

  10. Advanced handling-systems with enhanced performance flexibility

    International Nuclear Information System (INIS)

    1986-04-01

    This report describes the results of a project related to future applications and requirements for advanced handling systems. This report consists of six chapters. Following the description of the aims the tools for setting up the requirements for the handling systems including the experience during the data acquisition process is described. Furthermore some information is given about the current state of the art of robotics and manipulators. Of paramount importance are the descriptions of applications and related concepts in the following chapters leading to specific categories of advanced handling units. The paper closes with the description of the first concepts for realization. (orig./HP) [de

  11. Activation analysis and waste management for blanket materials of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

    International Nuclear Information System (INIS)

    Jiang, Jieqiong; Yuan, Baoxin; Zou, Jun; Wu, Yican

    2014-01-01

    The preliminary studies of the activation analysis and waste management for blanket materials of the multi-functional experimental fusion–fission hybrid reactor, i.e. Multi-Functional eXperimental Fusion Driven Subcritical system named FDS-MFX, were performed. The neutron flux of the FDS-MFX blanket was calculated using VisualBUS code and Hybrid Evaluated Nuclear Data Library (HENDL) developed by FDS Team. Based on these calculated neutron fluxes, the activation properties of blanket materials were analyzed by the induced radioactivity, the decay heat and the contact dose rate for different regions of the FDS-MFX blanket. The safety and environment assessment of fusion power (SEAFP) strategy, which was developed in Europe, was applied to FDS-MFX blanket for the management of activated materials. Accordingly, the classification and management strategy of activated materials after different cooling time were proposed for FDS-MFX blanket

  12. Advances in passive cooling design and performance analysis

    International Nuclear Information System (INIS)

    Woodcock, J.

    1994-01-01

    The Third International Conference on Containment Design and Operation continues the trend of rapidly extending the state of the art in containment methodology, joining other conferences, OECD-sponsored International Standard Problem exercises, and vendor licensing submittals. Methodology developed for use on plants with passive features is under increasing scrutiny for advanced designs, since the passive features are often the only deviation from existing operating base of the past 30 years of commercial nuclear power. This session, 'Containment Passive Safety Systems Design and Operation,' offers papers on a wide range of topics, with authors from six organizations from around the world, dealing with general passive containments, Westinghouse AP600, large (>1400 MWe) passive plants, and the AECL advanced CANDU reactor. This level and variety of participation underscores the high interest and accelerated methods development associated with advanced passive containment heat removal. The papers presented in this session demonstrate that significant contributions are being made to the advancement of technology necessary for building a new generation of safer, more economical nuclear plants. (author)

  13. Performance of Uplink Carrier Aggregation in LTE-Advanced Systems

    DEFF Research Database (Denmark)

    Wang, Hua; Rosa, Claudio; Pedersen, Klaus

    2010-01-01

    Carrier aggregation (CA) has been proposed to aggregate two or more component carriers (CCs) to support a much wider transmission bandwidth for LTE-Advanced systems. With carrier aggregation, it is possible to schedule a user equipment (UE) on multiple component carriers simultaneously. In this p...

  14. Measure of manufacturing performance in advanced manufacturing systems

    NARCIS (Netherlands)

    Ron, de A.J.

    1995-01-01

    Because of the financial risks as a result of the high investments, decisions concerning investing in advanced manufacturing systems are difficult. The difficulty to decide is gained by the lack of a well-defined measure to support decisions and alarming messages from the industry concerning inverse

  15. 2001 GUIDELINES FOR ADVANCEMENT AND EXCEPTIONAL PERFORMANCE AWARDS

    CERN Multimedia

    Human Resources Division

    2001-01-01

    1. General Following the recent Council decision that 1 September 2001 will be the date of entry into force of the new Career Structure conditions, the 2001 annual advancement exercise will be conducted on the same basis as in previous years. The practical details of the implementation of the new Career Structure conditions from 1 September 2001 and for future annual advancement exercises are presently subject to concertation between the Management and the Staff Association and will be communicated to the staff in due course. In the meantime, therefore, this year's exercise will take place within the present Staff Rules and Regulations, and following the procedures of the present Administrative Circular No. 26 (rev. 2). The effective date of decisions will be 1 July 2001. The Guidelines will be as described last year in CERN Bulletin 10/2000, of which the main aspects are summarised below. 2. Advancement guidelines The budget allocation for advancement in 2001 as a percentage of the basic salary cost is the s...

  16. Recent Advances in High-Performance Direct Methanol Fuel Cells

    Science.gov (United States)

    Narayanan, S. R.; Chun, W.; Valdez, T. I.; Jeffries-Nakamura, B.; Frank, H.; Surumpudi, S.; Halpert, G.; Kosek, J.; Cropley, C.; La Conti, A. B.; hide

    1996-01-01

    Direct methanol fuel cells for portable power applications have been advanced significantly under DARPA- and ARO-sponsored programs over the last five years. A liquid-feed, direct methanol fuel cell developed under these programs, employs a proton exchange membrane as electrolyte and operates on aqueous solutions of methanol with air or oxygen as the oxidant.

  17. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1994-11-01

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.) [de

  18. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Palmer, B.J.F.

    1987-11-01

    There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

  19. The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse.

    1988-03-01

    We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs

  20. Evaluation of heat transfer characteristics of a sphere-packed pipe for Flibe blanket

    International Nuclear Information System (INIS)

    Watanabe, Atsushi; Ebara, Shinji; Sagara, Akio; Hashizume, Hidetoshi

    2013-01-01

    A Flibe blanket has been proposed to be used in FFHR. Since Flibe has poor heat transfer performance, heat transfer promoter is required, and a sphere-packed pipe (SPP) has been proposed to enhance the heat transfer performance in the Flibe blanket. In this paper, the fluid flow and heat transfer characteristics in the SPP is evaluated numerically using a k–ε turbulent model for the flow field and an algebraic model for the thermal field. As a result, it was shown that bypass flows in the SPP play a significant role in heat transfer. Also it is thought that the turbulent energy can strongly affect heat transfer performance

  1. Neutronics analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Jiang Jieqiong; Wang Minghuang; Chen Zhong; Qiu Yuefeng; Liu Jinchao; Bai Yunqing; Chen Hongli; Hu Yanglin

    2010-01-01

    Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1 GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33 000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication.

  2. Blanket/first wall challenges and required R&D on the pathway to DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, Mohamed, E-mail: abdou@fusion.ucla.edu; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-11-15

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  3. Blanket/first wall challenges and required R&D on the pathway to DEMO

    International Nuclear Information System (INIS)

    Abdou, Mohamed; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-01-01

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  4. Epoxy blanket protects milled part during explosive forming

    Science.gov (United States)

    1966-01-01

    Epoxy blanket protects chemically milled or machined sections of large, complex structural parts during explosive forming. The blanket uniformly covers all exposed surfaces and fills any voids to support and protect the entire part.

  5. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.

    1981-01-01

    The Tandem Mirror Reactor, with its cylindrical central cell, has led to numerous blanket designs taking advantage of the simple geometry. Also many new applications for fusion neutrons are now being considered. To the pure fusion electricity producers and hybrids producing fissile fuel, we are adding studies of synthetic fuel producers and fission-suppressed hybrids. The three blanket concepts presented are new ideas and should be considered illustrative of the breadth of Livermore's application studies. They are not meant to imply fully analyzed designs

  6. Fusion blanket high-temperature heat transfer

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1983-01-01

    Deep penetration of 14 MeV neutrons makes two-temperature region blankets feasible. A relatively low-temperature (approx. 300 0 C) metallic structure is the vacuum/coolant pressure boundary, while the interior of the blanket, which is a simple packed bed of nonstructural material, operates at very high temperatures (>1000 0 C). The water-cooled shell structure is thermally insulated from the steam-cooled interior. High-temperature steam can dramatically increase the efficiency of electric power generation, as well as produce hydrogen and oxygen-based synthetic fuels at high-efficiency

  7. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

    Tritium release from the candidate ceramic materials, Li 2 O, LiA10 2 , Li 2 SiO 3 , Li 4 SiO 4 and Li 2 ZrO 3 , is being investigated in many blanket programs. Factors that affect tritium release from the ceramic into the helium sweep gas stream include operating temperature, ceramic microstructure, tritium transport and solubility in the solid. A review is presented of the material properties studied and of the irradiation programs and the results are summarized. The ceramic breeder blanket concept is briefly reviewed

  8. Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, A. [Univ. of California, Berkeley, CA (United States); Fratoni, M. [Univ. of California, Berkeley, CA (United States)

    2015-09-22

    , low electrical conductivity and therefore low MHD pressure drop, low chemical reactivity, and extremely low tritium inventory; the addition of sodium (FLiNaBe) has been considered because it retains the properties of FliBe but also lowers the melting point. Although many of these blanket concepts are promising, challenges still remain. The limited amount of beryllium available poses a problem for ceramic breeders such as the HCPB. FLiBe and FLiNaBe are highly viscous and have a low thermal conductivity. Lithium lead possesses a poor thermal conductivity which can cause problems in both DCLL and LiPb blankets. Additionally, the tritium permeation from these two blankets into plant components can be a problem and must be reduced. Consequently, Lawrence Livermore National Laboratory (LLNL) is attempting to develop a lithium-based alloy—most likely a ternary alloy—which maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) while reducing overall flammability concerns for use in the blanket of an inertial fusion energy (IFE) power plant. The LLNL concept employs inertial confinement fusion (ICF) through the use of lasers aimed at an indirect-driven target composed of deuterium-tritium fuel. The fusion driver/target design implements the same physics currently experimented at the National Ignition Facility (NIF). The plant uses lithium in both the primary coolant and blanket; therefore, lithium-related hazards are of primary concern. Although reducing chemical reactivity is the primary motivation for the development of new lithium alloys, the successful candidates will have to guarantee acceptable performance in all their functions. The scope of this study is to evaluate the neutronics performance of a large number of lithium-based alloys in the blanket of the IFE engine and assess their properties upon activation. This manuscript is organized as follows: Section 12 presents the models and methodologies used for the analysis; Section

  9. Effects of Different Practice Conditions on Advanced Instrumentalists' Performance Accuracy.

    Science.gov (United States)

    Rosenthal, Roseanne K.; And Others

    1988-01-01

    Examines the relative effects of the five practice conditions of modeling, singing, silent analysis, free practice, and control on instrumentalists' performance. Finds significant differences in subjects' performance of correct rhythms, phrasing or dynamics, and tempo, and insignificant differences among performances of correct notes and…

  10. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  11. Processing and waste disposal needs for fusion breeder blankets system

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1988-01-01

    We evaluated the waste disposal and recycling requirements for two types of fusion breeder blanket (solid and liquid). The goal was to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under U.S. Nuclear Regulatory Commission regulations. Described in this paper are the radionuclides expected in fusion blanket materials, plans for reprocessing and disposal of blanket components, and estimates for the operating costs involved in waste disposal. (orig.)

  12. Tritium inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Reiter, F.

    1990-01-01

    This report reviews studies of the transport of hydrogen isotopes in the DEMO relevant water-cooled Pb-17Li blanket to be tested in NET and in a self-cooled blanket which uses Pb-17Li or Flibe as a liquid breeder material and V or Fe as a first wall material. The time dependences of tritium inventory and permeation in these blankets and of deuterium and tritium recycling in the self-cooled blanket are presented and discussed

  13. Nuclear, thermo-mechanical and tritium release analysis of ITER breeding blanket

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Miki, Nobuharu; Akiba, Masato

    2003-06-01

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li 2 TiO 3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li 2 TiO 3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328

  14. Critical and shielding parametric studies with the Monte Carlo code TRIPOLI to identify the key points to take into account during the transportation of blanket assemblies with high ratio of americium

    International Nuclear Information System (INIS)

    Gosmain, Cecile-Aline

    2011-01-01

    In the framework of French research program on Generation IV sodium cooled fast reactor, one possible option consists in burning minor actinides in this kind of Advanced Sodium Technological Reactor. Two types of transmutation mode are studied in the world : the homogeneous mode of transmutation where actinides are scattered with very low enrichment ratio in fissile assemblies and the heterogeneous mode where fissile core is surrounded by blanket assemblies filled with minor actinides with ratio of incorporated actinides up to 20%. Depending on which element is considered to be burnt and on its content, these minor actinides contents imply constraints on assemblies' transportation between Nuclear Power Plants and fuel cycle facilities. In this study, we present some academic studies in order to identify some key constraints linked to the residual power and neutron/gamma load of such kind of blanket assemblies. To simplify the approach, we considered a modeling of a 'model cask' dedicated to the transportation of a unique irradiated blanket assembly loaded with 20% of Americium and basically inspired from an existent cask designed initially for the damaged fissile Superphenix assembly transport. Thermal calculations performed with EDF-SYRTHES code have shown that due to thermal limitations on cladding temperature, the decay time to be considered before transportation is 20 years. This study is based on explicit 3D representations of the cask and the contained blanket assembly with the Monte Carlo code TRIPOLI/JEFF3.1.1 library and concludes that after such a decay time, the transportation of a unique Americium radial blanket is feasible only if the design of our model cask is modified in order to comply with the dose limitation criterion. (author)

  15. Neutronics and activation of the preliminary reaction chamber of HiPER reactor based in a SCLL blanket

    Energy Technology Data Exchange (ETDEWEB)

    Juárez, Rafael, E-mail: rafael.juarez@upm.es [Instituto de Fusión Nuclear, UPM, Madrid (Spain); Escuela Técnica Superior de Ingenieros Industriales, UNED, Madrid (Spain); Sanz, Javier; Lopez-Revelles, A.J. [Escuela Técnica Superior de Ingenieros Industriales, UNED, Madrid (Spain); Perlado, José Manuel [Instituto de Fusión Nuclear, UPM, Madrid (Spain)

    2013-10-15

    Highlights: • Neutronic study of a proposal of a reaction chamber for HiPER reactor. • Two options for the blanket size, thin and thick, are studied and compared. • The thin blanket performs better than the thick blanket. • The proposed Vacuum Vessel is unviable as lifetime component in both cases. • Likely solutions for the Vacuum Vessel lifetime extension are explored. -- Abstract: The HiPER reactor design is exploring different reaction chambers. In this study, we tackle the neutronics and activation studies of a preliminary reaction chamber based in the following technologies: unprotected dry wall for the First Wall, self-cooled lead lithium blanket, and independent low activation steel Vacuum Vessel. The most critical free parameter in this stage is the blanket thickness, as a function of the {sup 6}Li enrichment. After a parametric study, we select for study both a “thin” and “thick” blanket, with “high” and “low” {sup 6}Li enrichment respectively, to reach a TBR = 1.1. To help to make a choice, we compute, for both blanket options, in addition to the TBR, the energy amplification factor, the tritium partial pressure, the {sup 203}Hg and {sup 210}Po total activity in the LiPb loop, and the Vacuum Vessel thickness required to guarantee the reweldability during its lifetime. The thin blanket shows a superior performance in the safety related issues and structural viability, but it operates at higher {sup 6}Li enrichment. It is selected for further improvements. The Vacuum Vessel shows to be unviable in both cases, with the thickness varying between 39 and 52 cm. Further chamber modifications, such as the introduction of a neutron reflector, are required to exploit the benefits of the thin blanket with a reasonable Vacuum Vessel.

  16. Proceedings of the eleventh international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2004-07-01

    This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li 2 TiO 3 and so on, fabrication technology developments and characterization of the Li 2 TiO 3 and Li 4 SiO 4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li 2 TiO 3 and Li 4 SiO 4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

  17. Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Aktaa, J., E-mail: jarir.aktaa@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Kecskés, S.; Pereslavtsev, P.; Fischer, U.; Boccaccini, L.V. [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-10-15

    Highlights: • First non-linear structural analysis for the European Helium Cooled Pebble Bed Blanket Module taking into account high dose irradiation. • Most critical areas were identified and analyzed with regard to the effect of irradiation on predicted damage at these areas. • Despite the extensive computing time 100 cycles were simulated by using the sub-modelling technique investigating damage at most critical area. • The results show a positive effect of irradiation on calculated damage which is mainly attributed to the irradiation induced hardening. - Abstract: For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design. For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of

  18. Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

    International Nuclear Information System (INIS)

    Aktaa, J.; Kecskés, S.; Pereslavtsev, P.; Fischer, U.; Boccaccini, L.V.

    2014-01-01

    Highlights: • First non-linear structural analysis for the European Helium Cooled Pebble Bed Blanket Module taking into account high dose irradiation. • Most critical areas were identified and analyzed with regard to the effect of irradiation on predicted damage at these areas. • Despite the extensive computing time 100 cycles were simulated by using the sub-modelling technique investigating damage at most critical area. • The results show a positive effect of irradiation on calculated damage which is mainly attributed to the irradiation induced hardening. - Abstract: For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design. For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of

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

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, Kenneth Mitchell [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1994-04-01

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

  20. Technical evaluation of major candidate blanket systems for fusion power reactor

    International Nuclear Information System (INIS)

    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 2 O/H 2 O/Be, Mo-alloy/Li 2 O/He/Be, Mo-alloy/LiAlO 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. (author)

  1. Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

    1980-09-01

    The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design

  2. Phase-IIC experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1992-12-01

    Neutronics experiments on two types of heterogeneous blankets have been performed as the Phase-IIC experiment of JAERI/USDOE collaborative program on fusion blanket neutronics. The experimental system was used in the same geometry as the previous Phase-IIA series which was a closed geometry using neutron source enclosure of lithium carbonate. The heterogeneous blankets selected here are the beryllium edge-on and the water coolant channel assemblies. In the former the beryllium and lithium-oxide layers are piled up alternately in the front part of test blanket. In the latter, the three simulated water cooling channels are settled in the Li 2 O blanket. These are producing steep gradient of neutron flux around material boundary. The calculation accuracy and measurement method for these features is a key of interest in the experiments. The measurements were performed for tritium production rate and the other nuclear parameters as well as the previous experiments. This report describes the experimental detail and the results enough to use for the benchmark data for testing the data and method of design calculation of fusion reactors. (author)

  3. Lead cooled heterogeneous accelerator driven molten-fluoride blanket for incineration of long-lived radioactive wastes

    International Nuclear Information System (INIS)

    Lopatkin, A.V.; Matyushechkin, V.M.; Tretyakov, I.T.; Blagovolin, P.P.; Kazaritsky, V.D.

    1997-01-01

    This paper presents a tentative design description and evaluation of the basic parameters of a lead cooled heterogeneous accelerator driven molten fluoride blanket. The proton beam of a 1 GeV accelerator strikes the blanket from below and generates spallation neutrons in the flow of lead, which serves as a target. These neutrons leave the target zone and get into a heterogeneous blanket with separated volumes of molten salts and lead. Fissile materials are dissolved in the salt. On getting into the molten salt volume the neutrons cause fission (transmutation) of the actinides, the produced heat being removed by circulation of molten lead. Two versions of the blanket design are examined. The first version: molten salt circulates in the fuel channels, while lead cools the channels flowing through the interchannel space (the salt channel design). The second version: it is lead that circulates in the channels, while molten salt takes up the interchannel space (the lead channel design). A preliminary blanket design study showed that both blanket designs possess a potential for improving performance. At present time the blanket design, mentioned above as the salt channel design, seems to be more promising. 1 ref., 2 figs., 2 tabs

  4. Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

    2016-11-01

    Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

  5. First wall fusion blanket temperature variation - slab geometry

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1978-01-01

    The first wall of a fusion blanket is approximated by a slab, with the surface facing the plasma subjected to an applied heat flux, while the rear surface is convectively cooled. The relevant parameters affecting the heat transfer during the early phases of heating as well as for large times are established. Analytical solutions for the temperature variation with time and space are derived. Numerical calculations for an aluminum and stainless steel slab are performed for a wall loading of 1 MW(th)/m 2 . Both helium and water cooling are considered. (Auth.)

  6. Conception of divertorless tokamak reactor with turbulent plasma blanket

    International Nuclear Information System (INIS)

    Nedospasov, A.V.; Tokar, M.Z.

    1980-01-01

    The results of the calculations presented here demonstrate that, with technically reasonable degree of the magnetic field stochastisation, the turbulent plasma blanket can take the place of a divertor. It performs the three main functions of the divertor: (a) the exhaust of the helium and unburned fuel; (b) weakening of the fast particle flux to the wall surface; and (c) essential reduction of the impurity content in the active zone of the reactor. Taking into account that plasma flows to the first wall along field lines, we may figuratively say that the first wall plays the role of a divertor in our conception. (orig.)

  7. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

    Tokuhiro, A.

    1991-11-01

    A review of recent research activities in liquid metal magnetohydrodynamics (LM-MHDs) is presented in this article. Two major reserach areas are discussed. The first topic involves the thermomechanical design issues in a proposed tokamak fusion reactor. The primary concerns are in the magneto-thermal-hydraulic performance of a self-cooled liquid metal blanket. The second topic involves the application of MHD in material processing in the metallurgical and semiconductor industries. The two representative applications are electromagnetic stirring (EMS) of continuously cast steel and the Czochralski (CZ) method of crystal growth in the presence of a magnetic field. (author) 24 figs., 10 tabs., 136 refs

  8. Sensitivity and uncertainty analysis of NET/ITER shielding blankets

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Gruppelaar, H.; Verschuur, K.A.

    1990-09-01

    Results are presented of sensitivity and uncertainty calculations based upon the European fusion file (EFF-1). The effect of uncertainties in Fe, Cr and Ni cross sections on the nuclear heating in the coils of a NET/ITER shielding blanket has been studied. The analysis has been performed for the total cross section as well as partial cross sections. The correct expression for the sensitivity profile was used, including the gain term. The resulting uncertainty in the nuclear heating lies between 10 and 20 per cent. (author). 18 refs.; 2 figs.; 2 tabs

  9. Connecting Performance Analysis and Visualization to Advance Extreme Scale Computing

    Energy Technology Data Exchange (ETDEWEB)

    Bremer, Peer-Timo [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mohr, Bernd [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Schulz, Martin [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Pasccci, Valerio [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Gamblin, Todd [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brunst, Holger [Dresden Univ. of Technology (Germany)

    2015-07-29

    The characterization, modeling, analysis, and tuning of software performance has been a central topic in High Performance Computing (HPC) since its early beginnings. The overall goal is to make HPC software run faster on particular hardware, either through better scheduling, on-node resource utilization, or more efficient distributed communication.

  10. INTOR first wall/blanket/shield activity

    International Nuclear Information System (INIS)

    Gohar, Y.; Billone, M.C.; Cha, Y.S.; Finn, P.A.; Hassanein, A.M.; Liu, Y.Y.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.

    1986-01-01

    The main emphasis of the INTOR first wall/blanket/shield (FWBS) during this period has been upon the tritium breeding issues. The objective is to develop a FWBS concept which produces the tritium requirement for INTOR operation and uses a small fraction of the first wall surface area. The FWBS is constrained by the dimensions of the reference design and the protection criteria required for different reactor components. The blanket extrapolation to commercial power reactor conditions and the proper temperature for power extraction have been sacrificed to achieve the highest possible local tritium breeding ratio (TBR). In addition, several other factors that have been considered in the blanket survey study include safety, reliability, lifetime fluence, number of burn cycles, simplicity, cost, and development issues. The implications of different tritium supply scenarios were discussed from the cost and availability for INTOR conditions. A wide variety of blanket options was explored in a preliminary way to determine feasibility and to see if they can satisfy the INTOR conditions. This survey and related issues are summarized in this report. Also discussed are material design requirements, thermal hydraulic considerations, structure analyses, tritium permeation through the first wall into the coolant, and tritium inventory

  11. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Anzidei, L.

    1994-01-01

    There are four breeding blanket concepts for a fusion DEMO reactor under development within the framework of the fusion technology programme of the European Union (EU). This paper describes the design of these concepts, the accompanying R + D programme and the status of the development. (authors). 8 figs., 1 tab

  12. Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

    Energy Technology Data Exchange (ETDEWEB)

    Pitulski, R.H.

    1979-10-01

    A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U/sup 233/ in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U/sup 233/, Pu/sup 239/, and H/sup 3/ production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m/sup -2/) exposure period. Although the results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids.

  13. Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

    International Nuclear Information System (INIS)

    Pitulski, R.H.

    1979-10-01

    A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U 233 in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U 233 , Pu 239 , and H 3 production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m -2 ) exposure period. Although the results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids

  14. Advanced Communications Technology Satellite (ACTS): Four-Year System Performance

    Science.gov (United States)

    Acosta, Roberto J.; Bauer, Robert; Krawczyk, Richard J.; Reinhart, Richard C.; Zernic, Michael J.; Gargione, Frank

    1999-01-01

    The Advanced Communications Technology Satellite (ACTS) was conceived at the National Aeronautics and Space Administration (NASA) in the late 1970's as a follow-on program to ATS and CTS to continue NASA's long history of satellite communications projects. The ACTS project set the stage for the C-band satellites that started the industry, and later the ACTS project established the use of Ku-band for video distribution and direct-to-home broadcasting. ACTS, launched in September 1993 from the space shuttle, created a revolution in satellite system architecture by using digital communications techniques employing key technologies such as a fast hopping multibeam antenna, an on-board baseband processor, a wide-band microwave switch matrix, adaptive rain fade compensation, and the use of 900 MHz transponders operating at Ka-band frequencies. This paper describes the lessons learned in each of the key ACTS technology areas, as well as in the propagation investigations.

  15. Transport modelling and gyrokinetic analysis of advanced high performance discharges

    International Nuclear Information System (INIS)

    Kinsey, J.E.; Imbeaux, F.; Staebler, G.M.; Budny, R.; Bourdelle, C.; Fukuyama, A.; Garbet, X.; Tala, T.; Parail, V.

    2005-01-01

    Predictive transport modelling and gyrokinetic stability analyses of demonstration hybrid (HYBRID) and advanced tokamak (AT) discharges from the International Tokamak Physics Activity (ITPA) profile database are presented. Both regimes have exhibited enhanced core confinement (above the conventional ITER reference H-mode scenario) but differ in their current density profiles. Recent contributions to the ITPA database have facilitated an effort to study the underlying physics governing confinement in these advanced scenarios. In this paper, we assess the level of commonality of the turbulent transport physics and the relative roles of the transport suppression mechanisms (i.e. E x B shear and Shafranov shift (α) stabilization) using data for select HYBRID and AT discharges from the DIII-D, JET and AUG tokamaks. GLF23 transport modelling and gyrokinetic stability analysis indicate that E x B shear and Shafranov shift stabilization play essential roles in producing the improved core confinement in both HYBRID and AT discharges. Shafranov shift stabilization is found to be more important in AT discharges than in HYBRID discharges. We have also examined the competition between the stabilizing effects of E x B shear and Shafranov shift stabilization and the destabilizing effects of higher safety factors and parallel velocity shear. Linear and nonlinear gyrokinetic simulations of idealized low and high safety factor cases reveal some interesting consequences. A low safety factor (i.e. HYBRID relevant) is directly beneficial in reducing the transport, and E x B shear stabilization can dominate parallel velocity shear destabilization allowing the turbulence to be quenched. However, at low-q/high current, Shafranov shift stabilization plays less of a role. Higher safety factors (as found in AT discharges), on the other hand, have larger amounts of Shafranov shift stabilization, but parallel velocity shear destabilization can prevent E x B shear quenching of the turbulent

  16. Transport modeling and gyrokinetic analysis of advanced high performance discharges

    International Nuclear Information System (INIS)

    Kinsey, J.; Imbeaux, F.; Bourdelle, C.; Garbet, X.; Staebler, G.; Budny, R.; Fukuyama, A.; Tala, T.; Parail, V.

    2005-01-01

    Predictive transport modeling and gyrokinetic stability analyses of demonstration hybrid (HYBRID) and Advanced Tokamak (AT) discharges from the International Tokamak Physics Activity (ITPA) profile database are presented. Both regimes have exhibited enhanced core confinement (above the conventional ITER reference H-mode scenario) but differ in their current density profiles. Recent contributions to the ITPA database have facilitated an effort to study the underlying physics governing confinement in these advanced scenarios. In this paper, we assess the level of commonality of the turbulent transport physics and the relative roles of the transport suppression mechanisms (i.e. ExB shear and Shafranov shift (α) stabilization) using data for select HYBRID and AT discharges from the DIII-D, JET, and AUG tokamaks. GLF23 transport modeling and gyrokinetic stability analysis indicates that ExB shear and Shafranov shift stabilization play essential roles in producing the improved core confinement in both HYBRID and AT discharges. Shafranov shift stabilization is found to be more important in AT discharges than in HYBRID discharges. We have also examined the competition between the stabilizing effects of ExB shear and Shafranov shift stabilization and the destabilizing effects of higher safety factors and parallel velocity shear. Linear and nonlinear gyrokinetic simulations of idealized low and high safety factor cases reveals some interesting consequences. A low safety factor (i.e. HYBRID relevant) is directly beneficial in reducing the transport, and ExB shear stabilization can win out over parallel velocity shear destabilization allowing the turbulence to be quenched. However, at low-q/high current, Shafranov shift stabilization plays less of a role. Higher safety factors (as found in AT discharges), on the other hand, have larger amounts of Shafranov shift stabilization, but parallel velocity shear destabilization can prevent ExB shear quenching of the turbulent

  17. A study on the enhancement of the reliability in gravure offset roll printing with blanket swelling control

    International Nuclear Information System (INIS)

    Kim, Ga Eul; Woo, Kyoohee; Kang, Dongwoo; Jang, Yunseok; Lee, Taik-Min; Kwon, Sin; Choi, Young-Man; Lee, Moon G

    2016-01-01

    In roll-offset printing (patterning) technology with a PDMS blanket as a transfer medium, one of the major reliability issues is the occurrence of swelling, which involves absorption of the ink solvent in the printing blanket with repeated printing. This study developed a method to resolve blanket swelling in gravure offset roll printing and performed experiments for performance verification. The physical phenomena of mass and heat transfer were applied to fabricate a device based on convection drying. The proposed device managed to effectively control blanket swelling through drying by blowing air and additional temperature control. The experiments verified that printing quality (in particular the variation of the width of printed patterns) was maintained over 500 continuous printing. (paper)

  18. Tritium transport in HCLL and WCLL DEMO blankets

    Energy Technology Data Exchange (ETDEWEB)

    Candido, Luigi [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Utili, Marco [ENEA UTIS- C.R. Brasimone, Bacino del Brasimone, Camugnano, BO (Italy); Nicolotti, Iuri [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Zucchetti, Massimo, E-mail: massimo.zucchetti@polito.it [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy)

    2016-11-01

    Highlights: • Tritium inventories and tritium losses are the main output of the presented model for HCLL and WCLL. • A parametric study has been performed, to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and/or losses. • An improved design is needed, in order to reduce the radiological hazard related to tritium activity. According to test number 7, HCLL-BB could be able to have a tritium inventory of 33.05 g and losses of 19.55 Ci/d. • WCLL-BB shows a very low radiological risk, much lower than that suggested (inventory: 17.48 g, losses: 3.2 Ci/d). An ptimization study has been performed aiming to minimize the water flow rate for an upgraded design. • Both for HCLL and WCLL, the most critical parameters able to produce relevant variations in inventories and losses are the helium/water fraction, the CPS/WDS and the permeation reduction factors. - Abstract: The Helium-Cooled Lithium Lead (HCLL) and Water-Cooled Lithium Lead (WCLL) Breeding Blankets are two of the four blanket designs proposed for DEMO reactor. The study of tritium transport inside the blankets is fundamental to assess their preliminary design and safety features. A mathematical model has been derived, in a new form making makes easier to determine the most critical components as far as tritium losses and tritium inventories are concerned, and to model the tritium performance of the whole system. Two cases have been studied, the former with tritium generation rate constant in time and the latter considering a typical pulsed operation for a time span of 100 h. Tritium inventories and tritium losses are the main output of the model. Tritium concentrations, inventories and losses are initially calculated and compared for the two blankets, in a reference case without permeation barriers or cold traps. A parametric study to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and

  19. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2014-01-01

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM

  20. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tariq Siddique, M.; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-05-15

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM.

  1. Classification methods for noise transients in advanced gravitational-wave detectors II: performance tests on Advanced LIGO data

    International Nuclear Information System (INIS)

    Powell, Jade; Heng, Ik Siong; Torres-Forné, Alejandro; Font, José A; Lynch, Ryan; Trifirò, Daniele; Cuoco, Elena; Cavaglià, Marco

    2017-01-01

    The data taken by the advanced LIGO and Virgo gravitational-wave detectors contains short duration noise transients that limit the significance of astrophysical detections and reduce the duty cycle of the instruments. As the advanced detectors are reaching sensitivity levels that allow for multiple detections of astrophysical gravitational-wave sources it is crucial to achieve a fast and accurate characterization of non-astrophysical transient noise shortly after it occurs in the detectors. Previously we presented three methods for the classification of transient noise sources. They are Principal Component Analysis for Transients (PCAT), Principal Component LALInference Burst (PC-LIB) and Wavelet Detection Filter with Machine Learning (WDF-ML). In this study we carry out the first performance tests of these algorithms on gravitational-wave data from the Advanced LIGO detectors. We use the data taken between the 3rd of June 2015 and the 14th of June 2015 during the 7th engineering run (ER7), and outline the improvements made to increase the performance and lower the latency of the algorithms on real data. This work provides an important test for understanding the performance of these methods on real, non stationary data in preparation for the second advanced gravitational-wave detector observation run, planned for later this year. We show that all methods can classify transients in non stationary data with a high level of accuracy and show the benefits of using multiple classifiers. (paper)

  2. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

    Engineering data on neutron irradiation performance of tritium breeders are needed to design the breeding blanket of fusion reactor. In this study, tritium release experiments of the breeders were carried out to examine the effects of various parameters (such as sweep gas flow rate, hydrogen content in sweep gas, irradiation temperature and thermal neutron flux) on tritium generation and release behavior. Lithium titanate (Li 2 TiO 3 ) is considered as a candidate tritium breeder in the blanket design of International Thermonuclear Experimental Reactor (ITER). As for the shape of the breeder material, a small spherical form is preferred to reduce the thermal stress induced in the breeder. Li 2 TiO 3 pebbles of about 170g in total weight and with 0.3 and 2 mm in diameter were manufactured by a wet process, and an assembly packed with the binary Li 2 TiO 3 pebbles was irradiated in Japan Materials Testing Reactor (JMTR). The tritium was generated in the Li 2 TiO 3 pebble bed and released from the pebble bed, and was swept downstream using the sweep gas for on-line analysis of tritium content. Concentration of total tritium and gaseous tritium (HT or T 2 gas) released from the Li 2 TiO 3 pebble bed were measured by ionization chambers, and the ratio of (gaseous tritium)/(total tritium) was evaluated. The sweep gas flow rate was changed from 100 to 900cm 3 /min, and hydrogen content in the sweep gas was changed from 100 to 10000 ppm. Furthermore, thermal neutron flux was changed using a window made of hafnium (Hf) neutron absorber. The irradiation temperature at an outer region of the Li 2 TiO 3 pebble bed was held between 200 and 400degC. The main results of this experiment are summarized as follows. 1) When the temperature at the outside edge of the Li 2 TiO 3 pebble bed exceeded 100degC, the tritium release from the Li 2 TiO 3 pebble bed started. The ratio of the tritium release rate and the tritium generation rate (normalized tritium release rate: R/G) reached

  3. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.

    1986-01-01

    A new conceptual design of a fusion reactor blanket simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBR), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation methods because design deficiencies that could affect results are insignificant. Inelastic scattering cross sections are identified as a major source of these discrepancies. The conceptual design of this FBBR analog, the fusion reactor blanket facility (FRBF), is presented. Essential features are a cylindrical geometry and a distributed, cosine-shaped line source of 14-MeV neutrons. This source can be created by sweeping a deuteron beam over an elongated titanium-tritide target. To demonstrate that the design of the FRBF will not contribute significant deviations in experimental results, neutronics analyses were performed: results of comparisons of 2-dimensional to 1-dimensional predictions are reported for two blanket compositions. Expected deviations from 1-D predictions which are due to source anisotropy and blanket asymmetry are minimal. Thus, design of the FRBF allows simple and straightforward interpretation of the experimental results, without a need for coarse 3-D calculations

  4. Design and analysis of breeding blanket with helium cooled solid breeder for ITER-TBM

    International Nuclear Information System (INIS)

    Yuan Tao; Feng Kaiming; Chen Zhi; Wang Xiaoyu

    2007-01-01

    Test blanket module (TBM) is one of important components in ITER. Some of related blanket technologies of future fusion, such as tritium self-sufficiency, the exaction of high-grade heat, design criteria and safety requirements and environmental impacts, will be demonstrated in ITER-TBM. In ITER device, the three equatorial ports have allocated for TBM testing. China had proposed to develop independently the ITER-TBM with helium cooled solid breeder in 12th meeting of test blanket workgroup (TBWG-12). In this work, the preliminary design and analysis for Chinese HCSB TBM will be carried out. The TBM must be contains the function of the first wall, breeding blanket, shield and structure. Finally, in the period of preliminary investigation, HCSB TBM design adopt modularization concept which is helium as coolant and tritium purge gas, ferritic/martensitic steel as structural material, Lithium orthosilicate (Li 4 SiO 4 ) as tritium breeder, beryllium pebble as neutron multiplier. TBM is allocated in standard vertical frame port. HCSB TBM consist of first wall, backplate, breeding sub-modules, caps, grid and support plate, and breeding sub-modules is arranged by layout of 2 x 6 in blanket box. In this paper, main components of HCSB TBM will be described in detail, also performance analysis of main components have been completed. (authors)

  5. RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

    International Nuclear Information System (INIS)

    Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

    2004-01-01

    RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

  6. Conceptual design of two helium cooled fusion blankets (ceramic and liquid breeder) for INTOR

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Taczanowski, S.

    1983-08-01

    Neutronic and heat transfer calculations have been performed for two helium cooled blankets for the INTOR design. The neutronic calculations show that the local tritium breeding ratios, both for the ceramic blanket (Li 2 SiO 3 ) and for the liquid blanket (Li 17 Pb 83 ) solutions, are 1.34 for natural tritium and about 1.45 using 30% Li 6 enrichment. The heat transfer calculations show that it is possible to cool the divertor section of the torus (heat flux = 1.7 MW/m 2 ) with helium with an inlet pressure of 52 bar and an inlet temperature of 40 0 C. The temperature of the back face of the divertor can be kept at 130 0 C. With helium with the same inlet conditions it is possible to cool the first wall as well (heat flux = 0.136 MW/m 2 ) and keep the back-face of this wall at a temperature of 120 0 C. For the ceramic blanket we use helium with 52 bar inlet pressure and 400 0 C inlet temperature to ensure sufficiently high temperatures in the breeder material. The maximum temperature in the pressure tubes containing the blanket is 450 0 C, while the maximum breeder particle temperature is 476 0 C. (orig./RW) [de

  7. Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

    Science.gov (United States)

    Urgorri, F. R.; Moreno, C.; Carella, E.; Rapisarda, D.; Fernández-Berceruelo, I.; Palermo, I.; Ibarra, A.

    2017-11-01

    The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg h-1 while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

  8. The Effect of Advanced Management Accounting Practices on the Competitive Strategies and Performance

    OpenAIRE

    Rahmi Yücel; Kayhan Ahmetoğulları

    2015-01-01

    This study is based on a sample of 300 managers from the production industry of West Marmara Region. The goal of this study is to examine the interaction among advanced management accounting practices, competitive strategies and company performance. As a result of study, it is found that advanced accounting management practices have a positive effect on the company performance and competitive strategies. In addition, competitive strategies have an effect on the company performance positively....

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

    International Nuclear Information System (INIS)

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

    1980-09-01

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

  10. [Development of performance evaluation and management system on advanced schistosomiasis medical treatment].

    Science.gov (United States)

    Zhou, Xiao-Rong; Huang, Shui-Sheng; Gong, Xin-Guo; Cen, Li-Ping; Zhang, Cong; Zhu, Hong; Yang, Jun-Jing; Chen, Li

    2012-04-01

    To construct a performance evaluation and management system on advanced schistosomiasis medical treatment, and analyze and evaluate the work of the advanced schistosomiasis medical treatment over the years. By applying the database management technique and C++ programming technique, we inputted the information of the advanced schistosomiasis cases into the system, and comprehensively evaluated the work of the advanced schistosomiasis medical treatment through the cost-effect analysis, cost-effectiveness analysis, and cost-benefit analysis. We made a set of software formula about cost-effect analysis, cost-effectiveness analysis, and cost-benefit analysis. This system had many features such as clear building, easy to operate, friendly surface, convenient information input and information search. It could benefit the performance evaluation of the province's advanced schistosomiasis medical treatment work. This system can satisfy the current needs of advanced schistosomiasis medical treatment work and can be easy to be widely used.

  11. Attentional models of multitask pilot performance using advanced display technology.

    Science.gov (United States)

    Wickens, Christopher D; Goh, Juliana; Helleberg, John; Horrey, William J; Talleur, Donald A

    2003-01-01

    In the first part of the reported research, 12 instrument-rated pilots flew a high-fidelity simulation, in which air traffic control presentation of auditory (voice) information regarding traffic and flight parameters was compared with advanced display technology presentation of equivalent information regarding traffic (cockpit display of traffic information) and flight parameters (data link display). Redundant combinations were also examined while pilots flew the aircraft simulation, monitored for outside traffic, and read back communications messages. The data suggested a modest cost for visual presentation over auditory presentation, a cost mediated by head-down visual scanning, and no benefit for redundant presentation. The effects in Part 1 were modeled by multiple-resource and preemption models of divided attention. In the second part of the research, visual scanning in all conditions was fit by an expected value model of selective attention derived from a previous experiment. This model accounted for 94% of the variance in the scanning data and 90% of the variance in a second validation experiment. Actual or potential applications of this research include guidance on choosing the appropriate modality for presenting in-cockpit information and understanding task strategies induced by introducing new aviation technology.

  12. Advancing the use of performance evaluation in health care

    DEFF Research Database (Denmark)

    Traberg, Andreas; Jacobsen, Peter

    2014-01-01

    Purpose – The purpose of this paper is to develop a framework for health care performance evaluation that enables decision makers to identify areas indicative of corrective actions. The framework should provide information on strategic pro-/regress in an operational context that justifies the need...... unit, where operational decision makers have been struggling with extensive amounts of performance information. Research limitations/implications – The implementation of the framework in a single case in a public and highly political environment restricts the generalizing potential. The authors...

  13. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-01

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  14. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-15

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  15. Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

    International Nuclear Information System (INIS)

    Chapin, D.L.; Green, L.; Lee, A.Y.; Culbert, M.E.; Kelly, J.L.

    1979-09-01

    The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO 2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li 2 O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

  16. Measurement-Based Performance Evaluation of Advanced MIMO Transceiver Designs

    Directory of Open Access Journals (Sweden)

    Schneider Christian

    2005-01-01

    Full Text Available This paper describes the methodology and the results of performance investigations on a multiple-input multiple-output (MIMO transceiver scheme for frequency-selective radio channels. The method relies on offline simulations and employs real-time MIMO channel sounder measurement data to ensure a realistic channel modeling. Thus it can be classified in between the performance evaluation using some predefined channel models and the evaluation of a prototype hardware in field experiments. New aspects for the simulation setup are discussed, which are frequently ignored when using simpler model-based evaluations. Example simulations are provided for an iterative ("turbo" MIMO equalizer concept. The dependency of the achievable bit error rate performance on the propagation characteristics and on the variation in some system design parameters is shown, whereas the antenna constellation is of particular concern for MIMO systems. Although in many of the considered constellations turbo MIMO equalization appears feasible in real field scenarios, there exist cases with poor performance as well, indicating that in practical applications link adaptation of the transmitter and receiver processing to the environment is necessary.

  17. Advanced Certification Program for Computer Graphic Specialists. Final Performance Report.

    Science.gov (United States)

    Parkland Coll., Champaign, IL.

    A pioneer program in computer graphics was implemented at Parkland College (Illinois) to meet the demand for specialized technicians to visualize data generated on high performance computers. In summer 1989, 23 students were accepted into the pilot program. Courses included C programming, calculus and analytic geometry, computer graphics, and…

  18. Predicting timing performance of advanced mechatronics control systems

    NARCIS (Netherlands)

    Voeten, J.P.M.; Hendriks, T.; Theelen, B.D.; Schuddemat, J.; Tabingh Suermondt, W.; Gemei, J.; Kotterink, C.; Huet, van J.; Eichler, G.; Kuepper, A.; Schau, V.; Fouchal, H.; Unger, H.

    2011-01-01

    Embedded control is a key product technology differentiator for many high-tech industries, including ASML. The strong increase in complexity of embedded control systems, combined with the occurrence of late changes in control requirements, results in many timing performance problems showing up only

  19. Advanced Modulation Techniques for High-Performance Computing Optical Interconnects

    DEFF Research Database (Denmark)

    Karinou, Fotini; Borkowski, Robert; Zibar, Darko

    2013-01-01

    We experimentally assess the performance of a 64 × 64 optical switch fabric used for ns-speed optical cell switching in supercomputer optical interconnects. More specifically, we study four alternative modulation formats and detection schemes, namely, 10-Gb/s nonreturn-to-zero differential phase-...

  20. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    International Nuclear Information System (INIS)

    Li, Jia; Jiang, Kecheng; Zhang, Xiaokang; Nie, Xingchen; Zhu, Qinjun; Liu, Songlin

    2016-01-01

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  1. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jia, E-mail: lijia@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China); Nie, Xingchen [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Zhu, Qinjun; Liu, Songlin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2016-12-15

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  2. Activation and afterheat analyses for the HCPB test blanket

    International Nuclear Information System (INIS)

    Pereslavtsev, P.; Fischer, U.

    2007-01-01

    The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in the framework of the European Fusion Technology Programme for performance tests in ITER. The recent development programme focussed on the detailed engineering design of the Test Blanket Module (TBM) and associated systems including the assessment of safety and licensing related issues with the objective to prepare for a preliminary Safety Report. To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20 degree ITER torus sector with an integrated TBM of the HCPB PI (Plant Integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a three (calendar) years period. It was simulated by a continuous irradiation for 3 years minus the last month and a discontinuous irradiation with 250 pulses (420 s pulse length, 1200 s power-off in between) over the last month. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER according to the M-DRG-1 irradiation scenario with a total first wall fluence of 0.3 MWa/m 2 . For both irradiation scenarios the radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity and afterheat of the TBM, its constituting components and materials including their

  3. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    Science.gov (United States)

    Enoeda, M.; Kosaku, Y.; Hatano, T.; Kuroda, T.; Miki, N.; Honma, T.; Akiba, M.; Konishi, S.; Nakamura, H.; Kawamura, Y.; Sato, S.; Furuya, K.; Asaoka, Y.; Okano, K.

    2003-12-01

    This paper presents results of conceptual design activities and associated R&D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R&D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  4. Advances in exercise, fitness, and performance genomics in 2012.

    Science.gov (United States)

    Pérusse, Louis; Rankinen, Tuomo; Hagberg, James M; Loos, Ruth J F; Roth, Stephen M; Sarzynski, Mark A; Wolfarth, Bernd; Bouchard, Claude

    2013-05-01

    A small number of excellent articles on exercise genomics issues were published in 2012. A new PYGM knock-in mouse model will provide opportunities to investigate the exercise intolerance and very low activity level of people with McArdle disease. New reports on variants in ACTN3 and ACE have increased the level of uncertainty regarding their true role in skeletal muscle metabolism and strength traits. The evidence continues to accumulate on the positive effects of regular physical activity on body mass index or adiposity in individuals at risk of obesity as assessed by their FTO genotype or by the number of risk alleles they carry at multiple obesity-susceptibility loci. The serum levels of triglycerides and the risk of hypertriglyceridemia were shown to be influenced by the interactions between a single nucleotide polymorphism (SNP) in the NOS3 gene and physical activity level. Allelic variation at nine SNPs was shown to account for the heritable component of the changes in submaximal exercise heart rate induced by the HERITAGE Family Study exercise program. SNPs at the RBPMS, YWHAQ, and CREB1 loci were found to be particularly strong predictors of the changes in submaximal exercise heart rate. The 2012 review ends with comments on the importance of relying more on experimental data, the urgency of identifying panels of genomic predictors of the response to regular exercise and particularly of adverse responses, and the exciting opportunities offered by recent advances in our understanding of the global architecture of the human genome as reported by the Encyclopedia of DNA Elements project.

  5. Advances in Exercise, Fitness, and Performance Genomics in 2012

    Science.gov (United States)

    Pérusse, Louis; Rankinen, Tuomo; Hagberg, James M.; Loos, Ruth J. F.; Roth, Stephen M.; Sarzynski, Mark A.; Wolfarth, Bernd; Bouchard, Claude

    2013-01-01

    A small number of excellent papers on exercise genomics issues have been published in 2012. A new PYGM knock-in mouse model will provide opportunities to investigate the exercise intolerance and very low activity level of people with McArdle disease. New reports on variants in ACTN3 and ACE have increased the level of uncertainty regarding their true role in skeletal muscle metabolism and strength traits. The evidence continues to accumulate on the positive effects of regular physical activity on body mass index (BMI) or adiposity in individuals at risk of obesity as assessed by their FTO genotype or by the number of risk alleles they carry at multiple obesity-susceptibility loci. Serum levels of triglycerides and the risk of hypertriglyceridemia were shown to be influenced by the interactions between a single nucleotide polymorphism (SNP) in the NOS3 gene and physical activity level. Allelic variation at nine SNPs was shown to account for the heritable component of the changes in submaximal exercise heart rate induced by the HERITAGE Family Study exercise program. SNPs at the RBPMS, YWHAQ, and CREB1 loci were found to be particularly strong predictors of the changes in submaximal exercise heart rate. The 2012 review ends with comments on the importance of relying more on experimental data, the urgency of identifying panels of genomic predictors of the response to regular exercise and particularly of adverse responses, and the exciting opportunities offered by recent advances in our understanding of the global architecture of the human genome as reported by the ENCODE project. PMID:23470294

  6. Advanced thorium cycles in LWRs and HWRs

    International Nuclear Information System (INIS)

    Radkowsky, A.

    The main aspects of advanced thorium cycles in LWRs and HWRs are reviewed. New concepts include the seed blanket close packed heavy water breeder, the light water seed blanket thorium burner and self-induced thorium cycle in CANDU type reactors. (author)

  7. Pebble bed blanket design for deuterium burning tandem mirror reactors

    International Nuclear Information System (INIS)

    Grotz, S.P.; Dhir, V.K.

    1983-01-01

    The UCLA tandem mirror reactor, SATYR, was developed around the capability of tandem mirrors with thermal barriers to burn deuterium at reasonable efficiency levels. The pebble bed concept has been incorporated into our blanket design for the following reasons: 1) Large area-to-volume ratio for purposes of heat removal; 2) Large volume of structure for high thermal capacity thus increasing the safety margin during off-normal incidents; 3) Relatively inexpensive manufacturing costs because of large acceptable tolerances and lack of exotic materials (i.e., lithium). A simplified stress analysis of the blanket module was performed to optimize and simplify the design. The pre-specified stress intensity limitations used were based upon a 30-year predicted lifetime for each module. Along with stress analysis of the vessel a detailed thermal hydraulic analysis of the pebble bed has been completed. Parameters affecting the pebble bed design are fluidization velocity, pressure drop, heat transfer coefficient, thermally induced stress in the spheres and spatial variation of the power density. Although reasonable gross thermal efficiencies of the 2 designs has been achieved (28% for H 2 O and 39% for He) the high net recirculating power fraction for heating and neutral beams results in relatively low net plant efficiencies (21% and 27%). The results show that a blanket can be designed with good thermal efficiency and a relative-ly simple configuration. However, application of this concept to the high Q deuterium-tritium fuel cycle would have difficulties resulting from the need for continuous removal of the tritium. (orig./HP)

  8. Tritium inventory and permeation in the ITER breeding blanket

    International Nuclear Information System (INIS)

    Violante, V.; Tosti, S.; Sibilia, C.; Felli, F.; Casadio, S.; Alvani, C.

    2000-01-01

    A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5x10 -3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5x10 -5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity

  9. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    International Nuclear Information System (INIS)

    Bohm, Tim; Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul; Ulrickson, Michael; Bullock, James

    2015-01-01

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  10. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

    2015-10-15

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  11. Design and R and D activities on ceramic breeder blanket for fusion experimental reactors in JAERI

    International Nuclear Information System (INIS)

    Kurasawa, T.; Takatsu, H.; Sato, S.; Nakahira, M.; Furuya, K.; Hashimoto, T.; Kawamura, H.; Kuroda, T.; Tsunematsu, T.; Seki, M.

    1995-01-01

    Design and R and D activities on ceramic breeder blanket of a fusion experimental reactor have been progressed in JAERI. A layered pebble bed type ceramic breeder blanket with water cooling is a prime candidate concept. Design activities have been concentrated on improvement of the design by conducting detailed analyses and also by fabrication procedure consideration based on the current technologies. A wide variety of R and Ds have also been conducted in accordance with the design activities. Development of fabrication technology of the blanket box structure and its mechanical testing, elementary testing on thermal performances of the pebble bed, and engineering-oriented material tests of breeder and beryllium pebbles are the main achievements during the last two years. (orig.)

  12. Neutronic optimization of a LiAlO2 solid breeder blanket

    International Nuclear Information System (INIS)

    Levin, P.; Ghoniem, N.M.

    1986-02-01

    In this report, a pressurized lobular blanket configuration is neutronically optimized. Among the features of this blanket configuration are the use of beryllium and LiAlO 2 solid breeder pins in a cross-flow configuration in a helium coolant. One-dimensional neutronic optimization calculations are performed to maximize the tritium breeding ratio (TER). The procedure involves spatial allocations of Be, LiAlO 2 , 9-C (ferritic steel), and He; in such a way as to maximize the TBR subject to several material, engineering and geometrical constraints. A TBR of 1.17 is achieved for a relatively thin blanket (approx. = 43 cm depth), and consistency with all imposed constraints

  13. Performance modelling for product development of advanced window systems

    DEFF Research Database (Denmark)

    Appelfeld, David

    properties (chapter 5) of CFSs. The last chapter concludes the thesis and the individual investigations. It is complicated to holistically evaluate the performance of a prototyped system, since simulation programs evaluate standardised products such as aluminium venetian blinds. State-of-the-art tools......The research presented in this doctoral thesis shows how the product development (PD) of Complex Fenestration Systems (CFSs) can be facilitated by computer-based analysis to improve the energy efficiency of fenestration systems as well as to improve the indoor environment. The first chapter defines...... the hypothesis and objectives of the thesis, which is followed by an extended introduction and background. The third chapter briefly suggests the PD framework which is suitable for CFSs. The fourth and fifth chapter refer to the detailed performance modelling of thermal properties (chapter 4) and optical...

  14. Aquifer storage and recovery: recent hydrogeological advances and system performance.

    Science.gov (United States)

    Maliva, Robert G; Guo, Weixing; Missimer, Thomas M

    2006-12-01

    Aquifer storage and recovery (ASR) is part of the solution to the global problem of managing water resources to meet existing and future freshwater demands. However, the metaphoric "ASR bubble" has been burst with the realization that ASR systems are more physically and chemically complex than the general conceptualization. Aquifer heterogeneity and fluid-rock interactions can greatly affect ASR system performance. The results of modeling studies and field experiences indicate that more sophisticated data collection and solute-transport modeling are required to predict how stored water will migrate in heterogeneous aquifers and how fluid-rock interactions will affect the quality of stored water. It has been well-demonstrated, by historic experience, that ASR systems can provide very large volumes of storage at a lesser cost than other options. The challenges moving forward are to improve the success rate of ASR systems, optimize system performance, and set expectations appropriately.

  15. Enhancing performance and reducing stress in sports technological advances

    CERN Document Server

    Ivancevic, Tijana; Greenberg, Ronald

    2015-01-01

    This book is designed to help athletes and individuals interested in high sports performance in their journey towards the perfection of human sports abilities and achievements. It has two main goals: accelerating the acquisition of motor skills and preparing and vigilantly reducing the recovery time after training and competition. The Diamond Sports Protocol (DSP) presents state-of-the-art techniques for current sport and health technologies, particularly neuromuscular electrical stimulation (Sports Wave), oxygen infusion (Oxy Sports), infrared (Sports Infrared Dome) and lactic acid cleaning (Turbo Sports). The book suggest DSP as an essential part of every future athlete's training, competition and health maintenance. The book is for everyone interested in superior sports performance, fast and effective rehabilitation from training and competition and sports injury prevention.

  16. The RISMC approach to perform advanced PRA analyses - 15332

    International Nuclear Information System (INIS)

    Mandelli, D.; Smith, C.; Riley, T.; Nielsen, J.; Alfonsi, A.; Rabiti, C.; Cogliati, J.

    2015-01-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power up-rates. In order to evaluate the impact of these two factors on the safety of the plant, the RISMC (Risk Informed Safety Margin Characterization) Pathway aims to develop simulation-based tools and methods to assess risks for existing nuclear power plants in order to optimize safety. This pathway, by developing new methods, is extending the state-of-the-practice methods that have been traditionally based on logic structures such as Event-Trees and Fault-Trees. These static types of models mimic system response in an inductive and deductive way respectively, yet are restrictive in the ways they can represent spatial and temporal constructs. RISMC analyses are performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool (RAVEN)currently under development at the Idaho National Laboratory. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power up-rate of a boiling water reactor system during a station blackout accident scenario. We employ the system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Our analysis is in fact performed by: 1) sampling values of a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the set of simulation runs. Results obtained give a detailed investigation of the issues associated with a plant power up-rate including the effects of station blackout accident scenarios. We are able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management

  17. Use of advanced simulations in fuel performance codes

    International Nuclear Information System (INIS)

    Van Uffelen, P.

    2015-01-01

    The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

  18. Advanced stellar compass - Onboard autonomous orbit determination, preliminary performance

    DEFF Research Database (Denmark)

    Betto, Maurizio; Jørgensen, John Leif; Jørgensen, Peter Siegbjørn

    2004-01-01

    and the cost of deep space missions. From past experience, it appears that navigation is the Achilles heel of deep space missions. Performed on ground, this imposes considerable constraints on the entire system and limits operations. This makes it is very expensive to execute, especially when the mission lasts...... be implemented into an ASC without degrading the attitude measurements; and (4) to identify the areas of development and consolidation. The results obtained are very encouraging....

  19. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  20. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  1. A Cross-National Comparison of Reported Effort and Mathematics Performance in TIMSS Advanced

    Science.gov (United States)

    Eklöf, Hanna; Pavešic, Barbara Japelj; Grønmo, Liv Sissel

    2014-01-01

    The purpose of the study was to measure students' reported test-taking effort and the relationship between reported effort and performance on the Trends in International Mathematics and Science Study (TIMSS) Advanced mathematics test. This was done in three countries participating in TIMSS Advanced 2008 (Sweden, Norway, and Slovenia), and the…

  2. One- and two-dimensional heating analyses of fusion synfuel blankets

    International Nuclear Information System (INIS)

    Tsang, J.S.K.; Lazareth, O.W.; Powell, J.R.

    1979-01-01

    Comparisons between one- and two-dimensional neutronics and heating analyses were performed on a Brookhaven designed fusion reactor blanket featuring synthetic fuel production. In this two temperature region blanket design, the structural shell is stainless steel. The interior of the module is a packed ball of high temperature ceramic material. The low temperature shell and the high temperature ceramic interior are separately cooled. Process steam (approx. 1500 0 C) is then produced in the ceramic core for the producion of H 2 and H 2 -based synthetic fuels by a high temperature electrolysis (HTE) process

  3. Cost study of the ESPRESSO blanket for a Tandem Mirror Reactor

    International Nuclear Information System (INIS)

    Raffray, A.R.; Hoffman, M.A.; Gaskins, T.

    1986-02-01

    A detailed cost study of the ESPRESSO blanket concept for the Tandem Mirror Fusion Reactor (TMR) has been performed to complement the thermal-hydraulic parametric study and to help narrow down the choice of parameters for the final design. The ESPRESSO blanket consists of a number of structurally independent ring modules. Each ring module is made up of a number of mutually pressure-supporting canisters containing arrays of breeder tubes. Two separate helium coolant flows are used: a main flow to cool the tube bank and a cooler first wall flow

  4. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    International Nuclear Information System (INIS)

    Daigo Tsuru; Mikio Enoeda; Masato Akiba

    2006-01-01

    Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

  5. Radiolysis and corrosion aspects of the aqueous self-cooled blanket concept

    International Nuclear Information System (INIS)

    Bruggeman, A.; Snykers, M.; Bogaerts, W.F.; Waeben, R.; Embrechts, M.J.; Steiner, D.

    1989-01-01

    Corrosion and radiolysis aspects of the Aqueous Self-Cooled Blanket concept, proposed as a potential shielding breeding blanket for near term fusion devices and fusion reactors, have been investigated. On the basis of preliminary results for selected aqueous solutions of lithium compounds, no particular corrosion problems have been revealed for the low-temperature concept envisaged for NET and radiolysis effects might be controlled by appropriate countermeasures. For the reactor-relevant high-temperature concept particular attention has to be paid to intergranular stress-corrosion and to the synergistic radiolysis-corrosion effects. Further information is needed from tests performed in relevant operational conditions. (orig.)

  6. Advanced analysis of finger-tapping performance: a preliminary study.

    Science.gov (United States)

    Barut, Cağatay; Kızıltan, Erhan; Gelir, Ethem; Köktürk, Fürüzan

    2013-06-01

    The finger-tapping test is a commonly employed quantitative assessment tool used to measure motor performance in the upper extremities. This task is a complex motion that is affected by external stimuli, mood and health status. The complexity of this task is difficult to explain with a single average intertap-interval value (time difference between successive tappings) which only provides general information and neglects the temporal effects of the aforementioned factors. This study evaluated the time course of average intertap-interval values and the patterns of variation in both the right and left hands of right-handed subjects using a computer-based finger-tapping system. Cross sectional study. Thirty eight male individuals aged between 20 and 28 years (Mean±SD = 22.24±1.65) participated in the study. Participants were asked to perform single-finger-tapping test for 10 seconds of test period. Only the results of right-handed (RH) 35 participants were considered in this study. The test records the time of tapping and saves data as the time difference between successive tappings for further analysis. The average number of tappings and the temporal fluctuation patterns of the intertap-intervals were calculated and compared. The variations in the intertap-interval were evaluated with the best curve fit method. An average tapping speed or tapping rate can reliably be defined for a single-finger tapping test by analysing the graphically presented data of the number of tappings within the test period. However, a different presentation of the same data, namely the intertap-interval values, shows temporal variation as the number of tapping increases. Curve fitting applications indicate that the variation has a biphasic nature. The measures obtained in this study reflect the complex nature of the finger-tapping task and are suggested to provide reliable information regarding hand performance. Moreover, the equation reflects both the variations in and the general

  7. Performance characteristics of the DIII-D advanced divertor cryopump

    International Nuclear Information System (INIS)

    Menon, M.M.; Maingi, R.; Wade, M.R.; Baxi, C.B.; Campbell, G.L.; Holtrop, K.L.; Hyatt, A.W.; Laughon, G.J.; Makariou, C.C.; Mahdavi, M.A.; Reis, E.E.; Schaffer, M.J.; Schaubel, K.M.; Scoville, J.T.; Smith, J.P.; Stambaugh, R.D.

    1993-10-01

    A cryocondensation pump, cooled by forced flow of two-phase helium, has been installed for particle exhaust from the divertor region of the DIII-D tokamak. The Inconel pumping surface is of coaxial geometry, 25.4 mm in outer diameter and 11.65 m in length. Because of the tokamak environment, the pump is designed to perform under relatively high pulsed heat loads (300 Wm -2 ). Results of measurements made on the pumping characteristics for D 2 , H 2 , and Ar are discussed

  8. Strategy Guideline: Advanced Construction Documentation Recommendations for High Performance Homes

    Energy Technology Data Exchange (ETDEWEB)

    Lukachko, A.; Gates, C.; Straube, J.

    2011-12-01

    As whole house energy efficiency increases, new houses become less like conventional houses that were built in the past. New materials and new systems require greater coordination and communication between industry stakeholders. The Guideline for Construction Documents for High Performance Housing provides advice to address this need. The reader will be presented with four changes that are recommended to achieve improvements in energy efficiency, durability and health in Building America houses: create coordination drawings, improve specifications, improve detail drawings, and review drawings and prepare a Quality Control Plan.

  9. Advanced Numerical Techniques of Performance Evaluation. Volume 1

    Science.gov (United States)

    1990-06-01

    34Speedup": Performance Analysis of Parallel Programs. Technical Report ANL-87-7, Mathematics and Computer Science Division, Argonne National Laboratory...and dequeue- pilicessoms a shared bus, and a write-through cache coherency4 hras anocu i amlc.wih a rotm sig if roocl[ Lvel & hakr 9U.Th Smety a atie...James M. Boyle. Beyond "Speedup": Parlor- mance Analysis of Parallel Progrms. Technical Report ANL-J7-7. Mathematics and Computer Science Division, Aronn

  10. Development of the advanced nuclear materials -Materials performance evaluation-

    International Nuclear Information System (INIS)

    Kim, Woo Chul; Noh, Kye Hoh; Han, Jung Hoh; Jung, Han Sub; Kim, Hong Pyo; Lee, Duk Hyun; Lee, Eun Heui; Hwang, Sung Sik; Huh, Doh Haeng

    1995-07-01

    The software for ACPD was modified to use multi-specimens and multi-frequency. The stress corrosion cracking resistance test of Alloy 600 in Pb contained water was performed by slow strain rate tester. The corrosion fatigue test machine was installed, and an autoclave for this test was purchased. The fatigue test was conducted in air. The stability for the long term test in DCPD was evaluated, and the improvement of current source and the revision of potential drop difference according to temperature variation increased the detection accuracy. A Ag/AgCl reference electrode and electrode support were assembled and the performance test was carried out at high temperature under high pressure. The zirconia pH electrode was assembled. The specimens with SUS304, Zr-2.5Nb were machined for irradiation assisted degradation test. The erosion/corrosion for the selected secondary side piping of Kori-1 was evaluated by CHECKMATE code. The chemical analysis and metallurgical inspection of the secondary piping of Kori-1 were conducted, and the erosion/corrosion test loop was made. 29 figs, 12 tabs, 11 refs. (Author)

  11. Tritium Management In HCLL-PPCS Model AB Blanket

    International Nuclear Information System (INIS)

    Ricapito, I.; Aiello, A.; Benamati, G.; Utili, M.; Ciampichetti, A.; Zucchetti, M.

    2006-01-01

    One the main issues in the HCLL blanket development for a prototype fusion reactor is the technical feasibility of the bred tritium processing system. The basis of such concern lies in the very low tritium-Pb17Li Sieverts' constant, as measured by different scientists in the past years. In the PPCS reactor 650 g/d of tritium must be generated in the breeding blanket while less than 1 g/y of tritium has to be released to the environment through the secondary cooling circuit. As a consequence, CPS (Coolant Purification System) plays a fundamental role because it has to keep at an acceptable level the tritium partial pressure in the primary HCS (Helium Cooling Circuit) limiting, therefore, the tritium environmental release through leakage and permeation into the secondary cooling circuit. On the other hand, the He mass flow-rate to be processed by CPS is linear with the tritium permeation rate from the breeder into HCS. Therefore, with the above mentioned low Sieverts' constant values and the consequent high tritium partial pressure in the liquid metal, the possibility to keep acceptable the CPS capacity depends on a highly efficient and stable performance of tritium permeation barriers, to be applied not only on the blanket cooling plates but also on the steam generator walls. However, the experimental results on the tritium permeation barriers under relevant operative conditions were so far quite disappointing. The new data on the Sieverts' constant achieved at ENEA CR Brasimone, one order of magnitude higher than those founding the past, have a big impact in relaxing the above mentioned requirements for the tritium management in PPCS model AB reactor. Besides presenting and discussing these recent experimental results, an updated assessment of the tritium permeation rate from the liquid breeder into HCS through the cooling plates and from HCS into the environment through the steam generators is given in this paper. The consequent new constraints in terms of tritium

  12. Performance Analysis of a Utility Helicopter with Standard and Advanced Rotors

    National Research Council Canada - National Science Library

    Yeo, Hyeonsoo; Bousman, William G; Johnson, Wayne

    2002-01-01

    Flight test measurements of the performance of the UH-60 Black Hawk helicopter with both standard and advanced rotors are compared with calculations obtained using the comprehensive helicopter analysis CAMRAD II...

  13. Determination of the Performance Parameters of a Spectrophotometer: An Advanced Experiment.

    Science.gov (United States)

    Cope, Virgil W.

    1978-01-01

    Describes an advanced analytical chemistry laboratory experiment developed for the determination of the performance parameters of a spectrophotometer. Among the parameters are the baseline linearity with wavelength, wavelength accuracy and respectability, stray light, noise level and pen response time. (HM)

  14. Very High-Performance Advanced Filter Bank Analog-to-Digital Converter (AFB ADC) Project

    National Research Council Canada - National Science Library

    Velazquez, Scott

    1999-01-01

    ... of the art by using a parallel array of individual commercial off the shelf converters. The significant performance improvements afforded by the Advanced Filter Bank Analog to Digital Converter (AFB ADC...

  15. Performance investigation of an advanced multi-effect adsorption desalination (MEAD) cycle

    KAUST Repository

    Thu, Kyaw; Kim, Young Deuk; Shahzad, Muhammad Wakil; Saththasivam, Jayaprakash; Ng, Kim Choon

    2015-01-01

    This article presents the development of an advanced adsorption desalination system with quantum performance improvement. The proposed multi-effect adsorption desalination (MEAD) cycle utilizes a single heat source i.e., low-temperature hot water

  16. E3 Success Story - Advancing Performance in Sustainability and Workforce Development

    Science.gov (United States)

    E3: North Carolina advances performance in sustainability and workforce development strategies for the state's manufacturers. The initiative helps communities and manufacturers address energy and sustainability challenges by leveraging expertise.

  17. Flibe blanket concept for transmuting transuranic elements and long lived fission products

    International Nuclear Information System (INIS)

    Gohar, Y.

    2000-01-01

    A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

  18. Mechanics of advanced materials analysis of properties and performance

    CERN Document Server

    Matveenko, Valery

    2015-01-01

    The last decades have seen a large extension of types of materials employed in various applications. In many cases these materials demonstrate mechanical properties and performance that vary significantly from those of their traditional counterparts. Such uniqueness is sought – or even specially manufactured – to meet increased requirements on modern components and structures related to their specific use. As a result, mechanical behaviors of these materials under different loading and environmental conditions are outside the boundaries of traditional mechanics of materials, presupposing development of new characterization techniques, theoretical descriptions and numerical tools. The book presents interesting examples of recent developments in this area. Among the studied materials are bulk metallic glasses, metamaterials, special composites, piezoelectric smart structures, nonwovens, etc.

  19. Advances in Exercise, Fitness, and Performance Genomics in 2015.

    Science.gov (United States)

    Sarzynski, Mark A; Loos, Ruth J F; Lucia, Alejandro; Pérusse, Louis; Roth, Stephen M; Wolfarth, Bernd; Rankinen, Tuomo; Bouchard, Claude

    2016-10-01

    This review of the exercise genomics literature encompasses the highest-quality articles published in 2015 across seven broad topics: physical activity behavior, muscular strength and power, cardiorespiratory fitness and endurance performance, body weight and adiposity, insulin and glucose metabolism, lipid and lipoprotein metabolism, and hemodynamic traits. One study used a quantitative trait locus for wheel running in mice to identify single nucleotide polymorphisms (SNPs) in humans associated with physical activity levels. Two studies examined the association of candidate gene ACTN3 R577X genotype on muscular performance. Several studies examined gene-physical activity interactions on cardiometabolic traits. One study showed that physical inactivity exacerbated the body mass index (BMI)-increasing effect of an FTO SNP but only in individuals of European ancestry, whereas another showed that high-density lipoprotein cholesterol (HDL-C) SNPs from genome-wide association studies exerted a smaller effect in active individuals. Increased levels of moderate-to-vigorous-intensity physical activity were associated with higher Matsuda insulin sensitivity index in PPARG Ala12 carriers but not Pro12 homozygotes. One study combined genome-wide and transcriptome-wide profiling to identify genes and SNPs associated with the response of triglycerides (TG) to exercise training. The genome-wide association study results showed that four SNPs accounted for all of the heritability of △TG, whereas the baseline expression of 11 genes predicted 27% of △TG. A composite SNP score based on the top eight SNPs derived from the genomic and transcriptomic analyses was the strongest predictor of ΔTG, explaining 14% of the variance. The review concludes with a discussion of a conceptual framework defining some of the critical conditions for exercise genomics studies and highlights the importance of the recently launched National Institutes of Health Common Fund program titled "Molecular

  20. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

  1. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

    The present report summarizes the design activities of the ITER first wall and shielding blanket conducted by the JA Home Team during this year (1994) in close contact with the JCT, and reported during the four Technical Meetings held at Garching ITER Co-center. These activities are based on the Task Agreement between the JCT and the JA Home Team. In the present report, a layered configuration composed of separate first walls, modular-type blanket modules and separate back plates has been proposed to realize reliable assembly and maintenance schemes as well as to realize reliable component designs under high surface heat loads, high neutron wall loading and electromagnetic loads during disruptions. Outline of the structural design, consideration on fabricability and maintainability, and the results of thermal, mechanical and electromagnetic analyses are described. (author)

  2. A Clinical Assessment Tool for Advanced Theory of Mind Performance in 5 to 12 Year Olds

    Science.gov (United States)

    O'Hare, Anne E.; Bremner, Lynne; Nash, Marysia; Happe, Francesca; Pettigrew, Luisa M.

    2009-01-01

    One hundred forty typically developing 5- to 12-year-old children were assessed with a test of advanced theory of mind employing Happe's strange stories. There was no significant difference in performance between boys and girls. The stories discriminated performance across the different ages with the lowest performance being in the younger…

  3. Fabrication of advanced military radiation detector sensor and performance evaluation

    International Nuclear Information System (INIS)

    Kang, Sin Yang

    2010-02-01

    shows a performance comparable to commercial ones. The beta ray sensor was manufactured using a BC-408 organic plastic scintillator. The main goal of this study was the measurement of beta - ray responses of the PMMA light guide attached to the sensor and the determination of window materials. Finally, the CsI(Tl) scintillator gamma ray sensor manufactured as engineering model and the shielding case was made of duralumin (side plane) and aluminium (window). The performance of the gamma - ray sensor was comparable to commercial high - sensitivity industrial sensors. These three types of scintillation sensors can be used directly as military equipment. And this study helps improving the protection of our country from nuclear threats

  4. Advanced Computational and Experimental Techniques for Nacelle Liner Performance Evaluation

    Science.gov (United States)

    Gerhold, Carl H.; Jones, Michael G.; Brown, Martha C.; Nark, Douglas

    2009-01-01

    The Curved Duct Test Rig (CDTR) has been developed to investigate sound propagation through a duct of size comparable to the aft bypass duct of typical aircraft engines. The axial dimension of the bypass duct is often curved and this geometric characteristic is captured in the CDTR. The semiannular bypass duct is simulated by a rectangular test section in which the height corresponds to the circumferential dimension and the width corresponds to the radial dimension. The liner samples are perforate over honeycomb core and are installed on the side walls of the test section. The top and bottom surfaces of the test section are acoustically rigid to simulate a hard wall bifurcation or pylon. A unique feature of the CDTR is the control system that generates sound incident on the liner test section in specific modes. Uniform air flow, at ambient temperature and flow speed Mach 0.275, is introduced through the duct. Experiments to investigate configuration effects such as curvature along the flow path on the acoustic performance of a sample liner are performed in the CDTR and reported in this paper. Combinations of treated and acoustically rigid side walls are investigated. The scattering of modes of the incident wave, both by the curvature and by the asymmetry of wall treatment, is demonstrated in the experimental results. The effect that mode scattering has on total acoustic effectiveness of the liner treatment is also shown. Comparisons of measured liner attenuation with numerical results predicted by an analytic model based on the parabolic approximation to the convected Helmholtz equation are reported. The spectra of attenuation produced by the analytic model are similar to experimental results for both walls treated, straight and curved flow path, with plane wave and higher order modes incident. The numerical model is used to define the optimized resistance and reactance of a liner that significantly improves liner attenuation in the frequency range 1900-2400 Hz. A

  5. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

    Wei, Jianghua; Song, Yuntao; Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong

    2015-01-01

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  6. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  7. A parametric study of rate of advance and area coverage rate performance of synthetic aperture radar.

    Energy Technology Data Exchange (ETDEWEB)

    Raynal, Ann Marie [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hensley, Jr., William H. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Burns, Bryan L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Doerry, Armin Walter [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2014-11-01

    The linear ground distance per unit time and ground area covered per unit time of producing synthetic aperture radar (SAR) imagery, termed rate of advance (ROA) and area coverage rate (ACR), are important metrics for platform and radar performance in surveillance applications. These metrics depend on many parameters of a SAR system such as wavelength, aircraft velocity, resolution, antenna beamwidth, imaging mode, and geometry. Often the effects of these parameters on rate of advance and area coverage rate are non-linear. This report addresses the impact of different parameter spaces as they relate to rate of advance and area coverage rate performance.

  8. Swiss fusion blanket experiments: Final report, November 1, 1985-October 31, 1987

    International Nuclear Information System (INIS)

    Woodruff, G.L.

    1987-01-01

    The major thrust of this project related to the effort to transfer the Lithium Blanket Module (LBM) to the Nuclear Engineering Laboratory of the Swiss Institute of Technology at Lausanne, and to the subsequent support with analytical calculations of a variety of experiments performed with the LBM. 12 refs

  9. Two-dimensional cross-section sensitivity and uncertainty analysis for fusion reactor blankets

    International Nuclear Information System (INIS)

    Embrechts, M.J.

    1982-02-01

    A two-dimensional sensitivity and uncertainty analysis for the heating of the TF coil for the FED (fusion engineering device) blanket was performed. The uncertainties calculated are of the same order of magnitude as those resulting from a one-dimensional analysis. The largest uncertainties were caused by the cross section uncertainties for chromium

  10. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m 2 and a particle heat flux of 1 MW/m 2 . Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  11. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

  12. Fabrication of advanced military radiation detector sensor and performance evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Sin Yang

    2010-02-15

    efficiency: 41%). This alpha - particle sensor shows a performance comparable to commercial ones. The beta ray sensor was manufactured using a BC-408 organic plastic scintillator. The main goal of this study was the measurement of beta - ray responses of the PMMA light guide attached to the sensor and the determination of window materials. Finally, the CsI(Tl) scintillator gamma ray sensor manufactured as engineering model and the shielding case was made of duralumin (side plane) and aluminium (window). The performance of the gamma - ray sensor was comparable to commercial high - sensitivity industrial sensors. These three types of scintillation sensors can be used directly as military equipment. And this study helps improving the protection of our country from nuclear threats

  13. Study on compact design of remote handling equipment for ITER blanket maintenance

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi

    2006-03-01

    In the ITER, the neutrons created by D-T reactions activate structural materials, and thereby, the circumstance in the vacuum vessel is under intense gamma radiation field. Thus, the in-vessel components such as blanket are handled and replaced by remote handling equipment. The objective of this report is to study the compactness of the remote handling equipment (a vehicle/manipulator) for the ITER blanket maintenance. In order to avoid the interferences between the blanket and the equipment during blanket replacement in the restricted vacuum vessel, a compact design of the equipment is required. Therefore, the compact design is performed, including kinematic analyses aiming at the reduction of the sizes of the vehicle equipped with a manipulator handling the blanket and the rail for the vehicle traveling in the vacuum vessel. Major results are as follows: 1. The compact vehicle/manipulator is designed concentration on the reduction of the rail size and simplification of the guide roller mechanism as well as the reduction of the gear diameter for vehicle rotation around the rail. Height of the rail is reduced from 500 mm to 400 mm by a parameter survey for weight, stiffness and stress of the rail. The roller mechanism is divided into two simple functional mechanisms composed of rollers and a pad, that is, the rollers support relatively light loads during rail deployment and vehicle traveling while a pad supports heavy loads during blanket replacement. Regarding the rotation mechanism, the double helical gear is adopted, because it has higher contact ratio than the normal spur gear and consequently can transfer higher force. The smaller double helical gear, 996 mm in diameter, can achieve 26% higher output torque, 123.5 kN·m, than that of the original spur gear of 1,460 mm in diameter, 98 kN·m. As a result, the manipulator becomes about 30% lighter, 8 tons, than the original weight, 11.2 tons. 2. Based on the compact design of the vehicle/manipulator, the

  14. Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

  15. Status of the European R and D on beryllium as multiplier material for breeder blankets

    International Nuclear Information System (INIS)

    Moeslang, A.; Boccaccini, L.V.; Rabaglino, E.; Piazza, G.; Cardella, A.; Sannen, L.; Scibetta, M.; Laan, J. van der; Hegeman, J.B.J.W.

    2004-01-01

    Within the international fusion community a variety of breeding blanket concepts are being considered, ranging from more conservative concepts to higher-risk concepts for fusion power reactors. In Europe, the Helium Cooled Pebble Bed (HCPB) blanket is one of the two reference concepts which will also be tested as Test Blanket Module (TBM) in ITER. In addition to the R and D for structural parts of the HCPB blanket, a considerable effort is devoted to the production and qualification of ceramic breeder and neutron multiplier (beryllium or beryllide) pebble beds. Since in the HCPB blanket pebbles made of lithium ceramics are foreseen, a high volume fraction of beryllium as a neutron multiplier to Li-based ceramic of about 4: l is needed. The typical loading conditions for beryllium are, with a neutron wall load of ∼12.5 MWa/m 2 and in ∼5 years lifetime: T min ∼300degC, T max ∼600-900degC, displacement damage ∼80 dpa, peak 4 He production ∼26000 appm and peak 3 H production ∼700 appm at the End-Of-Life. The behaviour of beryllium under irradiation is considered to be a key issue of the HCPB blanket, because of swelling due to helium bubbles and tritium retention. A large R and D programme on beryllium has been implemented in Europe, aimed at characterising and predicting the material behaviour before and under irradiation. An overview on experimental and modelling activities performed during the past 2 years is given with typical results on non-irradiated and irradiated Beryllium materials and pebble beds and the relevance of major results on future beryllium R and D is addressed. (author)

  16. Fabrication of prototype mockups of ITER shielding blanket with separable first wall

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Akiba, Masato

    2002-07-01

    Design of shielding blanket for ITER-FEAT applies the first wall which has the separable structure from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. This report summarizes the demonstrative fabrication of the ITER shielding blanket with separable first wall performed for the shielding blanket fabrication technology development, under the task agreement of G 16 TT 108 FJ (T420-2) in ITER Engineering Design Activity Extension Period. The objectives of the demonstrative fabrication are: to demonstrate the comprehensive fabrication technique in a large scale component (e.g the joining techniques for the beryllium armor/copper alloy and copper alloy/SS, and the slotting method of the FW and shield block); to develop an improved fabrication method for the shielding blanket based on the ITER-FEAT updated design. In this work, the fabrication technique of full scale separable first wall shield blanket was confirmed by fabricating full width Be armored first wall panel, full scale of 1/2 shield block with poloidal cooling channels. As the R and D for updated cooling channel configuration, the fabrication technique of the radial channel shield block was also demonstrated. Concluding to the all R and D results, it was demonstrated successfully that the fabrication technique and optimized conditions in the results obtained under the task agreement of G 16 TT 95 FJ (T420-1) was applicable to the prototype of the separable first wall blanket module. Additionally, basic echo data of ultra-sonic test method (UT) was obtained to show the applicability of UT method for in tube access detection of defect on the Cu alloy/SS tube interface. (author)

  17. Study of dynamic amplification factor of DEMO blanket caused by a gap at the supporting key

    International Nuclear Information System (INIS)

    Frosi, Paolo; Mazzone, Giuseppe

    2015-01-01

    Highlights: • With the preliminary hypothesis established, the dynamic displacements are not so high and the state of stress (not reported) does not exhibit large region with plastic strain. • The dynamic displacements show a certain dependency from the mesh adopted, and the geometry chosen. • The energy (kinetic or strain) of the whole structure gives useful information about the key behavior during impact. • In order to better understand the overall phenomenon other details (non-linear material, better evaluation of damping, other disruption rise-times and so on. - Abstract: Among the design activities of the in vessel components for DEMO promoted by European Fusion Development Agreement (EFDA) organization, this work deals with the gap required at the supporting keys of the blanket. Due to its higher operating temperatures compared to the vacuum vessel (VV) ones, this gap will increase during operation. The electro magnetic (EM) loads due to fast disruptions occur on a short time and might accelerate the blanket significantly before it touches the supporting keys, causing an impact of the blanket itself onto the keys. Depending on their stiffness, the EM loads with their short time scale could excite the structure's natural frequencies, causing dynamic amplification. Both phenomena (impact and dynamic amplification) can cause stresses in the structure significantly higher than the static ones. This work develops a finite element model of DEMO blanket to study its non-linear transient dynamic behavior under impact loadings. A VV sector, the ribs between the inner and outer VV, the backward manifolds and the supporting keys of the blanket have been modeled. The analyses have been performed with Abaqus [1] and Ansys [2] FEM codes focused on the displacements of the keys in their housing on the blanket. The dynamic amplification factor has been evaluated as the ratio of dynamic to static displacements in meaningful points of the structure for a growing gap

  18. Study of dynamic amplification factor of DEMO blanket caused by a gap at the supporting key

    Energy Technology Data Exchange (ETDEWEB)

    Frosi, Paolo, E-mail: paolo.frosi@enea.it; Mazzone, Giuseppe

    2015-10-15

    Highlights: • With the preliminary hypothesis established, the dynamic displacements are not so high and the state of stress (not reported) does not exhibit large region with plastic strain. • The dynamic displacements show a certain dependency from the mesh adopted, and the geometry chosen. • The energy (kinetic or strain) of the whole structure gives useful information about the key behavior during impact. • In order to better understand the overall phenomenon other details (non-linear material, better evaluation of damping, other disruption rise-times and so on. - Abstract: Among the design activities of the in vessel components for DEMO promoted by European Fusion Development Agreement (EFDA) organization, this work deals with the gap required at the supporting keys of the blanket. Due to its higher operating temperatures compared to the vacuum vessel (VV) ones, this gap will increase during operation. The electro magnetic (EM) loads due to fast disruptions occur on a short time and might accelerate the blanket significantly before it touches the supporting keys, causing an impact of the blanket itself onto the keys. Depending on their stiffness, the EM loads with their short time scale could excite the structure's natural frequencies, causing dynamic amplification. Both phenomena (impact and dynamic amplification) can cause stresses in the structure significantly higher than the static ones. This work develops a finite element model of DEMO blanket to study its non-linear transient dynamic behavior under impact loadings. A VV sector, the ribs between the inner and outer VV, the backward manifolds and the supporting keys of the blanket have been modeled. The analyses have been performed with Abaqus [1] and Ansys [2] FEM codes focused on the displacements of the keys in their housing on the blanket. The dynamic amplification factor has been evaluated as the ratio of dynamic to static displacements in meaningful points of the structure for a growing

  19. Assessing Advanced Airway Management Performance in a National Cohort of Emergency Medical Services Agencies.

    Science.gov (United States)

    Wang, Henry E; Donnelly, John P; Barton, Dustin; Jarvis, Jeffrey L

    2018-05-01

    Although often the focus of quality improvement efforts, emergency medical services (EMS) advanced airway management performance has few national comparisons, nor are there many assessments with benchmarks accounting for differences in agency volume or patient mix. We seek to assess variations in advanced airway management and conventional intubation performance in a national cohort of EMS agencies. We used EMS data from ESO Solutions, a national EMS electronic health record system. We identified EMS emergency responses with attempted advanced airway management (conventional intubation, rapid sequence intubation, sedation-assisted intubation, supraglottic airway insertion, and cricothyroidotomy). We also separately examined cases with initial conventional intubation. We determined EMS agency risk-standardized advanced airway management and initial conventional intubation success rates by using mixed-effects regression models, fitting agency as a random intercept, adjusting for patient age, sex, race, cardiac arrest, or trauma status, and use of rapid sequence or sedation-assisted intubation, and accounting for reliability variations from EMS agency airway volume. We assessed changes in agency advanced airway management and initial conventional intubation performance rank after risk and reliability adjustment. We also identified high and low performers (reliability-adjusted and risk-standardized success confidence intervals falling outside the mean). During 2011 to 2015, 550 EMS agencies performed 57,209 advanced airway management procedures. Among 401 EMS agencies with greater than or equal to 10 advanced airway management procedures, there were a total of 56,636 procedures. Median reliability-adjusted and risk-standardized EMS agency advanced airway management success was 92.9% (interquartile range 90.1% to 94.8%; minimum 58.2%; maximum 99.0%). There were 56 advanced airway management low-performing and 38 high-performing EMS agencies. Among 342 agencies with

  20. Strategy for the development of EU Test Blanket Systems instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Calderoni, P., E-mail: Pattrick.Calderoni@f4e.europa.eu; Ricapito, I.; Poitevin, Y.

    2013-10-15

    Highlights: • We developed a strategy for the development of instrumentation for EU ITER TBSs. • TBSs instrumentation functions: safety, operation and scientific mission. • Described activities are in support of ITER design review process. -- Abstract: The instrumentation of the HCLL and HCPB Test Blanket System is fundamental in ensuring that ITER safety and operational requirements are satisfied as well as in enabling the scientific mission of the TBM program. It carries out three essential functions: (i) safety, intended as compliance with ITER requirements toward public and workers protection; (ii) system control, intended as compliance with ITER operational requirements and investment protection; and (iii) scientific mission, intended as validating technology and predictive tools for blanket concepts relevant to fusion energy systems. This paper describes the strategy for instrumentation development by providing details of the following five steps to be implemented in procured activities in the short to mid-term (3–4 years): (i) provide mapping of sensors requirements based on critical review of preliminary design data; (ii) develop functional specifications for TBS sensors based on the analysis of operative conditions in the various ITER buildings in which they are located; (iii) assess availability of commercial sensors against developed specifications; (iv) develop prototypes when no available solution is identified; and (v) perform single effect tests for the most critical solicitations and post-test examination of commercial products and prototypes. Examples of technology assessment in two technical areas are included to reinforce and complement the strategy description.

  1. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 1

    International Nuclear Information System (INIS)

    Malang, S.; Reimann, J.; Sebening, H.; Barleon, L.; Bogusch, E.; Bojarsky, E.; Borgstedt, H.U.; Buehler, L.; Casal, V.; Deckers, H.; Feuerstein, H.; Fischer, U.; Frees, G.; Graebner, H.; John, H.; Jordan, T.; Kramer, W.; Krieg, R.; Lenhart, L.; Malang, S.; Meyder, R.; Norajitra, P.; Reimann, J.; Schwenk-Ferrero, A.; Schnauder, H.; Stieglitz, R.; Oschinski, J.; Wiegner, E.

    1991-12-01

    A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary, Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated R and D-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required R and D-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

  2. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 2. Detailed version

    International Nuclear Information System (INIS)

    John, H.; Malang, S.; Sebening, H.

    1991-12-01

    A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary. Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated RandD-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required RandD-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

  3. Key achievements in elementary R and Ds on water-cooled solid breeder blanket for ITER Test Blanket Module in JAERI

    International Nuclear Information System (INIS)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Tanigawa, H.; Tobita, K.; Akiba, M.; Hayashi, K.; Ochiai, K.; Nishitani, T.

    2005-01-01

    This paper presents significant progress in research and development (R and D) of key elementary technologies on the water-cooled solid breeder blanket for the ITER test blanket modules (TBMs) in JAERI. Development of module fabrication technology, bonding technology of armors, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup, and tritium release behavior from Li 2 TiO 3 pebble bed under neutron pulsed operation condition are summarized. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 deg C followed by normalizing at 930 deg C after the Hot Isostatic Pressing (HIP) process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a solid state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it was found that the thermal fatigue lifetime of F82H can be predicted by using Manson-Coffin's law. As for R and Ds on a breeder material, Li 2 TiO 3 , effective thermal conductivity of Li 2 TiO 3 pebble was measured under compressive force simulating the ITER TBM environment. The increase in the effective thermal conductivity of the pebble bed was about 2.5 % at the compressive strain of 0.9 % at 400 deg C. Neutronic performance of the blanket module mockup has been carried out by the 14 MeV neutron irradiation. It was confirmed that the measured tritium production rate agreed with the calculated values within about 10% difference. Also, tritium release from a Li 2 TiO 3 pebble bed was measured under pulsed neutron irradiation conditions simulating the ITER operation. (author)

  4. A new blanket tritium recovery experiment with intense DT neutron source at JAEA/FNS

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Kentaro, E-mail: ochiai.kentaro@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Edao, Yuki [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Kawamura, Yoshinori [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Ohta, Masayuki; Kwon, Saerom; Konno, Chikara [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan)

    2016-11-01

    Highlights: • For detail investigation of the tritium recovery performance on the fusion reactor blanket, we have started a new blanket tritium recovery experiment with ionization chamber at JAEA/FNS. • A new improved container was provided for the appropriate tritium measurement by IC and also utilized for the enhancement of TPR in the new container. The TPR was calculated with a calculation code MCNP5 and some typical nuclear data libraries and then the radioactivity of the tritium recovery with LSC corresponded with that of calculation. • The tritium release curves by the IC outputs are similar to those by the LSC output. However, it was indicated that the quantitative measurement by IC needed further improvement for the tritium recovery. - Abstract: We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

  5. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

    Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

  6. Feasibility study of LiF-BeF2 and chloride salts as blanket coolants for fusion power reactors

    International Nuclear Information System (INIS)

    Imamura, Y.

    1977-09-01

    The feasibility of using molten salts, in particular, nonberyllium-bearing chloride salts, as blanket coolants for Tokamak fusion reactors has been examined for the nucleonic and thermal/hydraulic aspects. It is concluded that the chloride salts, i.e., LiCl--KCl, LiCl--PbCl 2 and LiCl--SnCl 2 , can be used as the blanket coolant for a static lithium metal blanket provided that large blanket thickness can be tolerated, along with the use of U-238 for neutron multiplication in the cases of LiCl--KCl or LiCl--SnCl 2 cooled blankets. However, to make the appraisal complete, the tritium recovery and corrosion problems must be examined extensively, based on data not yet at hand. As for LiF--BeF 2 , it is observed that although the salt mixture can be used for a single fluid blanket with satisfactory nuclear performance, careful attention should be paid to the cooling capability

  7. Remote handling of the blanket segments: Testing of 1/3 scale mock-ups on the ROBERTINO facility

    International Nuclear Information System (INIS)

    Maisonnier, D.; Amelotti, F.; Chiasera, A.

    1994-01-01

    The remotized replacement of the blanket segments inside the Vacuum Vessel of a fusion reactor is one of the critical tasks for reactor components design, operational procedures, and safety. This open-quotes hostile environmentclose quotes task must be accomplished by a specific Blanket Handling Device, with a grasping device acting as open-quotes end-effectorclose quotes, because of intervention complexity, of components dimensions and weights, and of consequences of possible accidents during the blanket segments handling operations. Therefore, specific support experimental studies in this field appear to be necessary in order to: select appropriate blanket handling devices and procedures; assess the design of all components involved in the handling operations; perform checks in all field related to the robotized handling control (kinematics and dynamics of the grasping device trajectory planning and motion control, sensing and intelligence of the blanket handling devices, etc.); improve reliability and safety for the replacement sequences; give a realistic estimation of the time duration of the replacement duration. During the test phase, handling operations were carried out on the blanket mock-ups by means of different gripping devices. The operations were driven in the control room by means of the Motion command computer and the real time sensing data display allowed operations' control. The results were analyzed by charting the sensors' data

  8. Cross-section uncertainty study of the NET shielding blanket

    International Nuclear Information System (INIS)

    Jaeger, J.F.

    1990-11-01

    The Next European Torus (NET) is foreseen as the next step in the European development towards the controlled use of thermonuclear fusion. Detail design of the shielding blanket protecting the peripherals, more especially the super-conducting coils, is well advanced. A cross-section uncertainty, i.e. a study of the expected inaccuracy due to the nuclear cross-section data, has been done for the neutron-gamma reactions in the insulation of the coils for such a design. As an extension of previous work on the NET shielding blanket (e.g. MCNP calculations), it was deemed necessary to estimate the accuracy attainable with transport codes in view of the uncertainties in microscopic cross-sections. The code used, SENSIBL, is based on perturbation theory and uses covariance files, COVFILS-2, for the cross-section data. This necessitates forward and adjoint flux calculations with a transport code (e.g. ONEDANT, TRISM) and folding the information contained in these coupled fluxes with the accuracy estimates of the evaluators of the ENDF/B-V files. Transport, P 5 S 12 , calculations were done with the ONEDANT code, for a shielding blanket design with 714 MW plasma fusion power. Several runs were done to obtain well converged forward and adjoint fluxes (ca. 1%). The forward and adjoint integral responses agree to 2%, which is consistent with the above accuracy. The n-γ response was chosen as it is typical of the general accuracy and is available for all materials considered. The present version of SENSIBL allows direct use of the geometric files of ONEDANT (or TRISM) which simplifies the input. Covariance data is not available at present in COVFILS-2 for all of the materials considered. Only H, C, N, O, Al, Si, Fe, Ni, and Pb could be considered, the big absentee being copper. The resulting uncertainty for the neutron-gamma reactions in the insulation of the coil was found to be 17%. Simulating copper by aluminium produces a negligible increase in the uncertainty, mainly

  9. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Rampal, G.; Boccaccini, L.V.; Meyder, R.; Neuberger, H.; Laesser, R.; Poitevin, Y.; Zmitko, M.; Rigal, E.

    2006-01-01

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R-and-D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  10. Preliminary investigation on welding and cutting methods for first wall support leg in ITER blanket module

    International Nuclear Information System (INIS)

    Mohri, Kensuke; Suzuki, Satoshi; Enoeda, Mikio; Kakudate, Satoshi; Shibanuma, Kiyoshi; Akiba, Masato

    2006-08-01

    Concept of a module type of blanket has been applied to ITER shield blanket, of which size is typically 1mW x 1mH x 0.4mB with the weight of 4 ton, in order to enhance its maintainability and fabricability. Each shield blanket module consists of a shield block and four first walls which are separable from the shield block for the purpose of reduction of an electro-magnetic force in disruption events, radio-active waste reduction in the maintenance work and cost reduction in fabrication process. A first wall support leg, a part of the first wall component located between the first wall and the shield block, is required not only to be connected metallurgically to the shield block in order to withstand the electro-magnetic force and coolant pressure, but also to be able to replace the first wall more than 2 times in the hot cell during the life time of the reactor. Therefore, the consistent structure where remote handling equipment can be access to the joint and carry out the welding/cutting works perfectly to replace the first wall in the hot cell is required in the shield blanket design. This study shows an investigation of the blanket module no.10 design with a new type of the first wall support leg structure based on Disc-Cutter technology, which had been developed for the main pipe cutting in the maintenance phase and was selected out of a number of candidate methods, taking its large advantages into account, such as 1) a post-treatment can be eliminated in the hot cell because of no making material chips and of no need of lubricant, 2) the cut surface can be rewelded without any machining. And also, a design for the small type of Disc-Cutter applied to the new blanket module no.10 has been investigated. In conclusion, not only the good performance of Disc-Cutter technology applied to the updated blanket module, but also consistent structure of the simplified shield blanket module including the first wall support leg in order to satisfy the requirements in the

  11. Processing and waste disposal representative for fusion breeder blanket systems

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1987-01-01

    This study is an evaluation of the waste handling concepts applicable to fusion breeder systems. Its goal is to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under US Nuclear Regulatory regulations. The radionuclides expected in the materials used in fusion reactor blankets are described, as are plans for reprocessing and disposal of the components of different breeder blankets. An estimate of the operating costs involved in waste disposal is made

  12. A review of fusion breeder blanket technology, part 1

    International Nuclear Information System (INIS)

    Jackson, D.P.; Selander, W.N.; Townes, B.M.

    1985-01-01

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  13. Electromagnetic analysis of ITER shield blanket under VDE

    International Nuclear Information System (INIS)

    Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

    2010-01-01

    Electromagnetic force and torque of ITER shield blanket system and their surrounding major component under vertical displacement event (VDE) were calculated with finite element method. ANSYS APDL was used to simulate the shape and magnitude of plasmas current dynamically in the VDE course, and external magnetic field was imposed, then the induced current distribution inside the all conductor including the blanket was obtained from the calculation. The force and torque for every blanket module was obtained to assess the safety of blanket system under VDE. (authors)

  14. The lithium blanket program at the LOTUS facility

    International Nuclear Information System (INIS)

    File, J.; Haldy, P.A.; Quanci, J.

    1987-01-01

    An experimental program of neutron transport studies of the lithium Blanket Module (LBM) carried out with the LOTUS point-neutron source at the Ecole Polytechnique Federale de Lausanne (EPTL), Switzerland has been concluded. The major objectives of this program are to perform a series of neutron transport and tritium breeding experiments to qualify the LBM for future experiments on toroidal fusion devices such as TFTR to perform neutron multiplier experiments on the LBM employing various materials in a removable slab geometry; and, to compare the experimental results of radiation dosimetry and tritium breeding with the calculations of two and three dimensional neutron transport codes. An overview of the results from the radiation dosimetry and tritium assay are presented and compared to the two and three dimensional neutron transport codes

  15. Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

    International Nuclear Information System (INIS)

    Zhou, Z.; Yang, Y.; Xu, H.

    2011-01-01

    This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

  16. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    Noda, Kenji

    1998-03-01

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li 2 TiO 3 , tritium release behavior of Li 2 TiO 3 and Li 2 ZrO 3 including tritium diffusion, modeling of tritium release from Li 2 ZrO 3 in ITER condition, helium release behavior from Li 2 O, results of tritium release irradiation tests of Li 4 SiO 4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  17. The impact of attaining the Welsh Baccalaureate Advanced Diploma on academic performance in bioscience higher education

    OpenAIRE

    Yhnell, E; Wood, H; Baker, M.D; Amici-Dargan, S; Taylor, C; Randerson, P; Shore, A

    2016-01-01

    Since the introduction of the Welsh Baccalaureate Advanced Diploma Qualification (WBQ) in 2003, an increasing number of students are applying to higher education institutions (HEIs) with this qualification. The advanced-level WBQ is regarded as equivalent to one General Certificate of Education A-Level (GCE A-Level). This study assesses the impact of attaining the WBQ in addition to three GCE A-Levels on overall university degree performance in comparison to attaining four GCE A-Levels, in th...

  18. Application of Advanced Technology to Improve Plant Performance. Safety and Performance in Current NPPs

    International Nuclear Information System (INIS)

    Hashemian, H.M.

    2011-01-01

    Advances in computer technologies, signal processing, analytical modeling, and the advent of wireless sensors have provided the nuclear industry with ample means to automate and optimize maintenance activities and improve safety, efficiency, and availability, while reducing costs and radiation exposure to maintenance personnel. This paper provides a review of these developments and presents examples of their use in the nuclear power industry and the financial and safety benefits that they have produced. As the current generation of nuclear power plants have passed their mid-life, increased monitoring of their health is critical to their safe operation. This is especially true now that license renewal of nuclear power plants has accelerated, allowing some plants to operate up to 60 years or more. Furthermore, many utilities are maximizing their power output through uprating projects and retrofits. This puts additional demand and more stress on the plant equipment such as the instrumentation and control (I and C) systems and the reactor internal components making them more vulnerable to the effects of aging, degradation, and failure. In the meantime, the nuclear power industry is working to reduce generation costs by adopting condition-based maintenance strategies and automation of testing activities. These developments have stimulated great interest in on-line monitoring (OLM) technologies and new diagnostic and prognostic methods to anticipate, identify, and resolve equipment and process problems and ensure plant safety, efficiency, and immunity to accidents. The foundation for much of the required technologies has already been established through 40 years of research and development (R and D) efforts performed by numerous organizations, scientists, and engineers around the world including the author. This paper provides examples of these technologies and demonstrates how the gap between some of the more important R and D efforts and end users have been filled

  19. Multiple Module Simulation of Water Cooled Breeding Blankets in K-DEMO Using Thermal-Hydraulic Analysis Code MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.

  20. The ARIES-AT advanced tokamak, Advanced technology fusion power plant

    International Nuclear Information System (INIS)

    Najmabadi, Farrokh; Abdou, A.; Bromberg, L.

    2006-01-01

    The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R and D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ x = 2.2) which is the result of a 'thinner' blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β N . ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb-17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 deg. C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb-17Li to about 1000 deg. C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES

  1. Systematic methodology for estimating direct capital costs for blanket tritium processing systems

    International Nuclear Information System (INIS)

    Finn, P.A.

    1985-01-01

    This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared

  2. About possible technologies of creation nanostructures blankets

    International Nuclear Information System (INIS)

    Blednova, Zh.M.; Chaevskij, M.I.; Rusinov, P.O.

    2008-01-01

    Possible technologies of formation nanostructures blankets are considered: a method of thermal carrying over of weights in the conditions of a high gradient of temperatures; the combined method including cathode-plasma nitriding in the conditions of low pressure and drawing of nitride of the titan in a uniform work cycle; the combined method including high-frequency ionic nitriding and drawing of carbide of chrome by pyrolysis chrome and organic of connections in plasma of the decaying category. Possibility of formation layered nanostructures layers is shown.

  3. Fabrication and characterization of advanced neutron multipliers for DEMO blanket

    Directory of Open Access Journals (Sweden)

    Masaru Nakamichi

    2016-12-01

    Full Text Available Prototypic pebbles with Be12V composition, which do not undergo a peritectic reaction during cooling, were fabricated and characterized, because this composition is not only unnecessary for the homogenization treatment, but also able to prevent increase of specific surface area. The results of granulation experiments indicated that the prototypic pebbles of single-phase Be12V were successfully fabricated without a homogenization treatment. The results of hydrogen generation reaction experiments showed that the prototypic pebbles with Be12V composition exhibited superior oxidation properties compared to pure Be pebbles and similar to those of as-granulated Be–Ti beryllide pebbles: as-granulated Be12V pebbles exhibited good resistance to water vapor. The results of deuterium retention experiments indicated that beryllides exhibit lower deuterium-trapping efficiency than other tested materials. Because of a small desorption from beryllides, the total retention of deuterium in Be12V was evaluated to be approximately 20% of that in pure Be.

  4. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  5. Development of a qualitative evaluation framework for performance shaping factors (PSFs) in advanced MCR HRA

    International Nuclear Information System (INIS)

    Lee, Seung Woo; Kim, Ar Ryum; Ha, Jun Su; Seong, Poong Hyun

    2011-01-01

    Highlights: → Context changes in advanced MCR have impact on PSFs. → PSFs in the 1st and 2nd generation HRA methods are reviewed. → We made a qualitative evaluation framework for PSF based on human factor issues. - Abstract: Human reliability analysis (HRA) is performed as part of the probabilistic risk assessment to identify and quantify human actions and the associated impacts on structures, systems, and components of complex facilities. In performing HRA, conditions that influence human performance have been analyzed in terms of several context factors. These context factors, which are called performance shaping factors (PSFs) are used to adjust the basic human error probability (BHEP), and PSFs have been derived in various ways depending on the HRA methods used. As the design of instrumentation and control (I and C) systems for nuclear power plants (NPPs) is rapidly moving toward fully digital I and C, and modern computer techniques have been gradually introduced into the design of advanced main control room (MCR), computer-based human-system interfaces (HSIs), such as CRT-based displays, large display panels (LDPs), advanced information systems, soft control, and computerized procedure system (CPS) will be applied in advanced MCR. Environmental changes in MCR have some implications for PSFs, and they have an influence on when PSFs should be applied in HRA because different situations might induce different internal or external factors which can lead to human errors. In this study, PSFs for advanced MCR HRA are derived, and a new qualitative evaluation framework for these PSFs is suggested. First, PSFs from various HRA methods are collected, and these PSFs are further grouped into PSFs categories to be used in advanced MCR HRA. Second, human factor (HF) issues in advanced MCR are analyzed and derived to be used as an evaluation framework for PSFs.

  6. APT Blanket System Loss-of-Flow Accident (LOFA) Analysis Based on Initial Conceptual Design - Case 1: with Beam Shutdown and Active RHR

    International Nuclear Information System (INIS)

    Hamm, L.L.

    1998-01-01

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report

  7. Methods to enhance blanket power density in low-power fusion devices

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow testing of advanced reactor blanket modules on INTOR at representative conditions. The tentative approach adopted for this task consists of three parts. First, the requirements for augmented heating of the test module are outlined for different applications of interest. Second, methods are identified which have potential for augmenting the heating power in a test module, and this list of methods is narrowed to those which appear to be most useful. Finally, these methods are examined in more detail to determine the practical benefits of employing each

  8. Comprehensive structural analysis of the HCPB demo blanket under thermal, mechanical, electromagnetic and radiation induced loads

    International Nuclear Information System (INIS)

    Boccaccini, L.V.; Norajitra, P.; Ruatto, P.; Scaffidi-Argentina, F.

    1998-01-01

    For the helium-cooled pebble bed (HCPB) blanket, which is one of the two reference concepts studied within the European Demo Development Program, a comprehensive finite element (FEM) structural analysis has been performed. The analysis refers to the steady-state operating conditions of an outboard blanket segment. On the basis of a three-dimensional model of radial-toroidal sections of the segment box, thermal stresses caused by the temperature gradients in the blanket structure have been calculated. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions as well as an accidental over-pressurization of the blanket box have been accounted for. The stresses caused by a central plasma major disruption from an initial current of 20 MA to zero in 20 ms have been also taken into account. Radiation-induced dimensional changes of breeder and multiplier material caused by both helium production and neutron damage, have also been evaluated and discussed. All the above loads have been combined as input for a FEM stress analysis and the resulting stress distribution has been evaluated according to the American Society of Mechanical Engineers (ASME) norms. (orig.)

  9. Experimental investigation of MHD pressure losses in a mock-up of a liquid metal blanket

    Science.gov (United States)

    Mistrangelo, C.; Bühler, L.; Brinkmann, H.-J.

    2018-03-01

    Experiments have been performed to investigate the influence of a magnetic field on liquid metal flows in a scaled mock-up of a helium cooled lead lithium (HCLL) blanket. During the experiments pressure differences between points on the mock-up have been recorded for various values of flow rate and magnitude of the imposed magnetic field. The main contributions to the total pressure drop in the test-section have been identified as a function of characteristic flow parameters. For sufficiently strong magnetic fields the non-dimensional pressure losses are practically independent on the flow rate, namely inertia forces become negligible. Previous experiments on MHD flows in a simplified test-section for a HCLL blanket showed that the main contributions to the total pressure drop in a blanket module originate from the flow in the distributing and collecting manifolds. The new experiments confirm that the largest pressure drops occur along manifolds and near the first wall of the blanket module, where the liquid metal passes through small openings in the stiffening plates separating two breeder units. Moreover, the experimental data shows that with the present manifold design the flow does not distribute homogeneously among the 8 stacked boxes that form the breeding zone.

  10. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  11. Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

    International Nuclear Information System (INIS)

    Barrera, G.; Branas, B.; Lucas, J.; Doncel, J.; Medrano, M.; Garcia, A.; Giancarli, L.; Ibarra, A.; Li Puma, A.; Maisonnier, D.; Sardain, P.

    2008-01-01

    The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 deg. C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 deg. C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions

  12. Transient electromagnetic and dynamic structural analyses of a blanket structure with coupling effects

    Energy Technology Data Exchange (ETDEWEB)

    Koganezawa, K. [Mitsubishi Atomic Power Industries, Inc., Yokohama (Japan); Kushiyama, M. [Mitsubishi Atomic Power Industries, Inc., Yokohama (Japan); Niikura, S. [Mitsubishi Atomic Power Industries, Inc., Yokohama (Japan); Kudough, F. [Mitsubishi Atomic Power Industries, Inc., Yokohama (Japan); Onozuka, M. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan); Koizumi, K. [Japan Atomic Energy Research Inst., Ibaraki (Japan)

    1995-12-31

    Transient electromagnetic and dynamic structural analyses of a blanket structure in the fusion experimental reactor (FER) under a plasma disruption event and a vertical displacement event (VDE) have been performed to investigate the dynamic structural characteristics and the feasibility of the structure. Coupling effects between eddy currents and dynamic deflections have also been taken into account in these analyses. In this study, the inboard blanket was employed because of our computer memory limitation. A 1/192 segment model of a full torus was analyzed using the analytical code, EDDYCUFF. In the plasma disruption event, the maximum magnetic pressure caused by eddy currents and poloidal fields was 1.2MPa. The maximum stress intensity by this magnetic pressure was 114MPa. In the VDE, the maximum magnetic pressure was 2.4MPa and the maximum stress intensity was 253MPa. This stress was somewhat beyond the allowable stress limit. Therefore, the blanket structure and support design should be reviewed to reduce the stress to a suitable value. In summary, the dynamic structural characteristics and design issues of the blanket structure have been identified. (orig.).

  13. Acoustic contributions of a sound absorbing blanket placed in a double panel structure: absorption versus transmission.

    Science.gov (United States)

    Doutres, Olivier; Atalla, Noureddine

    2010-08-01

    The objective of this paper is to propose a simple tool to estimate the absorption vs. transmission loss contributions of a multilayered blanket unbounded in a double panel structure and thus guide its optimization. The normal incidence airborne sound transmission loss of the double panel structure, without structure-borne connections, is written in terms of three main contributions; (i) sound transmission loss of the panels, (ii) sound transmission loss of the blanket and (iii) sound absorption due to multiple reflections inside the cavity. The method is applied to four different blankets frequently used in automotive and aeronautic applications: a non-symmetric multilayer made of a screen in sandwich between two porous layers and three symmetric porous layers having different pore geometries. It is shown that the absorption behavior of the blanket controls the acoustic behavior of the treatment at low and medium frequencies and its transmission loss at high frequencies. Acoustic treatment having poor sound absorption behavior can affect the performance of the double panel structure.

  14. Effect of advanced injection timing on the performance of rapeseed oil in diesel engines

    International Nuclear Information System (INIS)

    Nwafor, O.M.I.; Rice, G.; Ogbonna, A.I.

    2000-01-01

    Combustion studies on both diesel fuel and vegetable oil fuels, with the standard and advanced injection timing, were carried out using the same engine and test procedures so that comparative assessments may be made. The diesel engine principle demands self-ignition of the fuel as it is injected at some degrees before top dead centre (BTDC) into the hot compressed cylinder gas. Longer delays between injection and ignition lead to unacceptable rates of pressure rise with the result of diesel knock because too much fuel is ready to take part in premixed combustion. Alternative fuels have been noted to exhibit longer delay periods and slower burning rate especially at low load operating conditions hence resulting in late combustion in the expansion stroke. Advanced injection timing is expected to compensate these effects. The engine has standard injection timing of 30degC BTDC. The injection was first advanced by 5.5degC given injection timing of 35.5degC BTDC. The engine performance was very erratic on this timing. The injection was then advanced by 3.5degC and the effects are presented in this paper. The engine performance was smooth especially at low load levels. The ignition delay was reduced through advanced injection but tended to incur a slight increase in fuel consumption. Moderate advanced injection timing is recommended for low speed operations. (Author)

  15. Welding techniques development of CLAM steel for Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Li Chunjing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China)], E-mail: lcj@ipp.ac.cn; Huang Qunying; Wu Qingsheng; Liu Shaojun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Lei Yucheng [Jiangsu University, Zhenjiang, Jiangsu, 212013 (China); Muroga, Takeo; Nagasaka, Takuya [National Institute for Fusion Science, Toki, Jifu, 509-5292 (Japan); Zhang Jianxun [Xi' an Jiaotong University, Xi' an, Shanxi, 710049 (China); Li Jinglong [Northwestern Polytechnical University, Xi' an, Shanxi, 710072 (China)

    2009-06-15

    Fabrication techniques for Test Blanket Module (TBM) with CLAM are being under development. Effect of surface preparation on the HIP diffusion bonding joints was studied and good joints with Charpy impact absorbed energy close to that of base metal have been obtained. The mechanical properties test showed that effect of HIP process on the mechanical properties of base metal was little. Uniaxial diffusion bonding experiments were carried out to study the effect of temperature on microstructure and mechanical properties. And preliminary experiments on Electron Beam Welding (EBW), Tungsten Inert Gas (TIG) Welding and Laser Beam Welding (LBW) were performed to find proper welding techniques to assemble the TBM. In addition, the thermal processes assessed with a Gleeble thermal-mechanical machine were carried out as well to assist the fusion welding research.

  16. Blanket and vacuum vessel design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    Iida, H.; Minato, A.; Kitamura, K.

    1983-01-01

    The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li 2 O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

  17. Two-phase-flow cooling concept for fusion reactor blankets

    International Nuclear Information System (INIS)

    Bender, D.J.; Hoffman, M.A.

    1977-01-01

    The new two-phase heat transfer medium proposed is a mixture of potassium droplets and helium which permits blanket operation at hih temperature and low pressure, while maintaining acceptable pumping power requirements, coolant ducting size, and blanket structure fractions. A two-phase flow model is described. The helium pumping power and the primary heat transfer loop are discussed

  18. Overview of the TFTB lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

    The Lithium Blanket Module (LBM) is an ∼ 80-cm 3 module, representative of a helium-cooled lithium oxide fusion reactor blanket module. This paper summarizes the design, development, and construction of the LBM, and indicates the present status of the LBM program

  19. Accelerator driven heavy water blanket on circulating fue