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Sample records for super insulation blankets

  1. 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

  2. Super-insulation

    International Nuclear Information System (INIS)

    Gerold, J.

    1985-01-01

    The invention concerns super-insulation, which also acts as spacing between two pressurized surfaces, where the crossing bars in at least two layers are provided, with interposed foil. The super-insulation is designed so that it can take compression forces and limits thermal radiation and thermal conduction sufficiently, where the total density of heat flow is usually limited to a few watts per m 2 . The solution to the problem is characterized by the fact that the bars per layer are parallel and from layer to layer they are at an angle to each other and the crossover positions of the bars of different layers are at fixed places and so form contact columns. The basic idea is that bars crossing over each other to support compression forces are used so that contact columns are formed, which are compressed to a certain extent by the load. (orig./PW) [de

  3. Super insulating aerogel glazing

    DEFF Research Database (Denmark)

    Schultz, Jørgen Munthe; Jensen, Karsten Ingerslev; Kristiansen, Finn Harken

    2004-01-01

    form the weakest part of the thermal envelope with respect to heat loss coefficient, but on the other hand also play an important role for passive solar energy utilisation. For window orientations other than south, the net energy balance will be close to or below zero. However, the properties......Monolithic silica aerogel offers the possibility of combining super insulation and high solar energy transmittance, which has been the background for a previous and a current EU project on research and development of monolithic silica aerogel as transparent insulation in windows. Generally, windows...... of aerogel glazing will allow for a positive net energy gain even for north facing vertical windows in a Danish climate during the heating season. This means that high quality daylight can be obtained even with additional energy gain. On behalf of the partners of the two EU projects, results related...

  4. Development of insulating coatings for liquid metal blankets

    International Nuclear Information System (INIS)

    Malang, S.; Borgstedt, H.U.; Farnum, E.H.; Natesan, K.; Vitkovski, I.V.

    1994-07-01

    It is shown that self-cooled liquid metal blankets are feasible only with electrically insulating coatings at the duct walls. The requirements on the insulation properties are estimated by simple analytical models. Candidate insulator materials are selected based on insulating properties and thermodynamic consideration. Different fabrication technologies for insulating coatings are described. The status of the knowledge on the most crucial feasibility issue, the degradation of the resisivity under irradiation, is reviewed

  5. Thermal performance measurements of a 100 percent polyester MLI [multilayer insulation] system for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.

    1989-09-01

    The plastic materials used in the multilayer insulation (MLI) blankets of the superconducting magnets of the Superconducting Super Collider (SSC) are comprised entirely of polyesters. This paper reports on tests conducted in three separate experimental blanket arrangements. The tests explore the thermal performance of two candidate blanket joint configurations each employing a variation of a stepped-butted joint nested between sewn blanket seams. The results from the joint configurations are compared to measurements made describing the thermal performance of the basic blanket materials as tested in an ideal joint configuration. Twenty foil sensors were incorporated within each test blanket to measure interstitial layer and joint layer temperatures. Heat flux and thermal gradients are reported for high and degraded insulating vacuums, and during transient and steady state conditions. In complement with this paper is an associate paper bearing the same title head but with the title extension 'Part 1: Instrumentation and experimental preparation (300K-80K)'. 5 refs., 8 figs., 2 tabs

  6. 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.

  7. Vibration damage testing of thermal barrier fibrous blanket insulation

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.

    1984-01-01

    GA Technologies is engaged in a long-term, multiphase program to determine the vibration characteristics of thermal barrier components leading to qualification of assemblies for High Temperature Gas-Cooled Reactor (HTGR) service. The phase of primary emphasis described herein is the third in a series of acoustic tests and uses as background the more elemental tests preceding it. Two sizes of thermal barrier coverplates with one fibrous blanket insulation type were tested in an acoustic environment at sound pressure levels up to 160 dB. Three tests were conducted using sinusoidal and random noise for up to 200 h duration at room temperature. Frequent inspections were made to determine the progression of degradation using definition of stages from a prior test program. Initially the insulation surface adjacent to the metallic seal sheets (noise side) assumed a chafed or polished appearance. This was followed by flattening of the as-received pillowed surface. This stage was followed by a depression being formed in the vicinity of the free edge of the coverplate. Next, loose powder from within the blanket and from fiber erosion accumulated in the depression. Prior experience showed that the next stage of deterioration exhibited a consolidation of the powder to form a local crust. In this test series, this last stage generally failed to materialize. Instead, surface holes generated by solid ceramic particulates (commonly referred to as 'shot') constituted the stage following powdering. With the exception of some manufacturing-induced anomalies, the final stage, namely, gross fiber breakup, did not occur. It is this last stage that must be prevented for the thermal barrier to maintain its integrity. (orig./GL)

  8. Multilayer insulation (MLI) in the Superconducting Super Collider: A practical engineering approach to physical parameters governing MLI thermal performance

    International Nuclear Information System (INIS)

    Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.

    1989-03-01

    Multilayer insulation (MLI) is employed in cryogenic devices to control the heat load of those devices. The physics defining the thermal performance of an MLI system is extremely complex due to the thermal dynamics of numerous interdependent parameters which in themselves contribute differently depending on whether boundary conditions are transient or steady-state. The Multilayer Insulation 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 blankets will be required in the 10,000 cryogenic devices comprising the SSC accelerator. Each blanket will be nearly 56 feet long by 6 feet wide and will consist of as many as 32 reflective and 31 spacer layers of material. Discussed are MLI material choices, and the physical parameters which contribute to the operational performance of MLI systems. Disclosed is a method for fabricating MLI blankets by employing a large diameter winding mandrel having a circumference sufficient for the required blanket length. The blanket fabrication method assures consistency in mass produced MLI blankets by providing positive control of the dimensional parameters which contribute to the MLI blanket thermal performance. The fabrication method can be used to mass produce prefabricated MLI blankets that by virtue of the product 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. 9 refs., 4 figs., 2 tabs

  9. Examination of compression and resilience characteristics of fibrous insulation blankets

    International Nuclear Information System (INIS)

    Brislin, R.J.; Middleton, A.

    1979-08-01

    Load-deflection characteristics of alumina and alumino-silicate fibrous blankets were experimentally determined. Load retention and springback capability of combinations of these materials were measured in a 10,000-hour test at surface temperatures of 650 to 1000 0 C (1200 to 1832 0 F). Experimental results are presented and future testing plans are discussed

  10. 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.)

  11. Design of the multilayer insulation system for the Superconducting Super Collider 50mm dipole cryostat

    International Nuclear Information System (INIS)

    Boroski, W.N.; Nicol, T.H.; Schoo, C.J.

    1991-03-01

    The development of the multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) 50 mm collider dipole cryostat is an ongoing extension of work conducted during the 40 mm cryostat program. While the basic design of the MLI system for the 50 mm cryostat resembles that of the 40 mm cryostat, results from measurements of MLI thermal performance below 80K have prompted a re-design of the MLI system for the 20K thermal radiation shield. Presented is the design of the MLI system for the 50 mm collider dipole cryostat, with discussion focusing on system performance, blanket geometry, cost-effective fabrication techniques, and built-in quality control measures that assure consistent thermal performance throughout the SSC accelerator. 16 refs., 8 figs., 2 tabs

  12. Aerogels: transparent and super-insulating materials; Les aerogels: isolants transparent-super isolants

    Energy Technology Data Exchange (ETDEWEB)

    Melka, S.; Rigacci, A.; Achard, P.; Bezian, J.J. [Ecole des Mines de Paris, 06 - Sophia-Antipolis (France); Sallee, H.; Chevalier, B. [Centre des Sciences et Techniques du Batiment, 38 - Saint-Martin-d`Heres (France)

    1996-12-31

    Recent studies have demonstrated the super-insulating properties of silica aerogel in its monolithic or finely divided state. In its monolithic state, this material conciliates excellent thermal insulation performances, a good transmission of visible light and interesting acoustic properties. Also its amazing structural characteristics (lightness, high global porosity, small diameter of pores) are particularly interesting for its use in double glazing windows as transparent insulating spacer. The aim of the work carried out by the Energetic Centre of the Ecole des Mines of Paris is to understand the thermal transfer phenomena in all forms of silica aerogel. In this paper, the main steps of the synthesis process of monolithic silica aerogel is presented with the thermal conductivities obtained. Then, a model is built to describe the thermal transfer mechanisms in finely divided aerogel beds. Finally, the hot wire thermal characterization method is presented and the results obtained on silica aerogels are discussed. (J.S.) 16 refs.

  13. Dynamic test of the ITER blanket key and ceramic insulated pad

    International Nuclear Information System (INIS)

    Khomyakov, S.; Sysoev, G.; Strebkov, Yu.; Kucherov, A.; Ioki, K.

    2010-01-01

    The dynamic testing of the blanket module's key integrated into ITER vacuum vessel portion has been performed in 2008 to investigate its capability to react the electro-magnetic (EM) loads. The preliminary analysis showed the large dynamic amplification factor (DAF) of the reactions because of technological gaps between the blanket module and key. Shock load may yield the bronze pads, which protect the blanket electrical insulation from damage. However the dynamic analysis of such particularly non-linear system needs an experimental ground and confirmation. Toward this end, as well as demonstration of the key reliability, the special test facility has been made, and the full-scale mock-up of the inboard intermodular key was tested. So as not to scale non-linear dynamic parameters, 1-ton mass was built on the single flexible support. The key was welded in a 60-mm thick steel plate modeled with a fragment of the VV. The different gaps were set in between the bronze pad of the key and the mass shock worker. This system (supplemented with some additional constraints) has natural oscillations like as the 4-ton module built on four flexible supports. Thus the most critical radial torque might be modeled with a straight force. The objectives of the test were as follows: dynamic response, DAF and damping factor determination; measurement of the strain oscillations in the key's base and in the weld seam; comparison of the measured data with computation results. The paper will present the analytical grounds of the testing conditions, test facility description, analytical adaptation of the facility, experimental results, its comparison with analysis and discussion, and guidelines for the next experimental phase.

  14. Stratification of TAD boundaries reveals preferential insulation of super-enhancers by strong boundaries.

    Science.gov (United States)

    Gong, Yixiao; Lazaris, Charalampos; Sakellaropoulos, Theodore; Lozano, Aurelie; Kambadur, Prabhanjan; Ntziachristos, Panagiotis; Aifantis, Iannis; Tsirigos, Aristotelis

    2018-02-07

    The metazoan genome is compartmentalized in areas of highly interacting chromatin known as topologically associating domains (TADs). TADs are demarcated by boundaries mostly conserved across cell types and even across species. However, a genome-wide characterization of TAD boundary strength in mammals is still lacking. In this study, we first use fused two-dimensional lasso as a machine learning method to improve Hi-C contact matrix reproducibility, and, subsequently, we categorize TAD boundaries based on their insulation score. We demonstrate that higher TAD boundary insulation scores are associated with elevated CTCF levels and that they may differ across cell types. Intriguingly, we observe that super-enhancers are preferentially insulated by strong boundaries. Furthermore, we demonstrate that strong TAD boundaries and super-enhancer elements are frequently co-duplicated in cancer patients. Taken together, our findings suggest that super-enhancers insulated by strong TAD boundaries may be exploited, as a functional unit, by cancer cells to promote oncogenesis.

  15. Analyses of Hubble Space Telescope Aluminized-Teflon Multilayer Insulation Blankets Retrieved After 19 Years of Space Exposure

    Science.gov (United States)

    de Groh, Kim K.; Perry, Bruce A.; Mohammed, Jelila S.; Banks, Bruce

    2015-01-01

    Since its launch in April 1990, the Hubble Space Telescope (HST) has made many important observations from its vantage point in low Earth orbit (LEO). However, as seen during five servicing missions, the outer layer of multilayer insulation (MLI) has become increasingly embrittled and has cracked in many areas. In May 2009, during the 5th servicing mission (called SM4), two MLI blankets were replaced with new insulation and the space-exposed MLI blankets were retrieved for degradation analyses by teams at NASA Glenn Research Center (GRC) and NASA Goddard Space Flight Center (GSFC). The retrieved MLI blankets were from Equipment Bay 8, which received direct sunlight, and Equipment Bay 5, which received grazing sunlight. Each blanket was divided into several regions based on environmental exposure and/or physical appearance. The aluminized-Teflon (DuPont, Wilmington, DE) fluorinated ethylene propylene (Al-FEP) outer layers of the retrieved MLI blankets have been analyzed for changes in optical, physical, and mechanical properties, along with chemical and morphological changes. Pristine and as-retrieved samples (materials) were heat treated to help understand degradation mechanisms. When compared to pristine material, the analyses have shown how the Al-FEP was severely affected by the space environment. Most notably, the Al-FEP was highly embrittled, fracturing like glass at strains of 1 to 8 percent. Across all measured properties, more significant degradation was observed for Bay 8 material as compared to Bay 5 material. This paper reviews the tensile and bend-test properties, density, thickness, solar absorptance, thermal emittance, x-ray photoelectron spectroscopy (XPS) and energy dispersive spectroscopy (EDS) elemental composition measurements, surface and crack morphologies, and atomic oxygen erosion yields of the Al-FEP outer layer of the retrieved HST blankets after 19 years of space exposure.

  16. Protection against cold in prehospital care-thermal insulation properties of blankets and rescue bags in different wind conditions.

    Science.gov (United States)

    Henriksson, Otto; Lundgren, J Peter; Kuklane, Kalev; Holmér, Ingvar; Bjornstig, Ulf

    2009-01-01

    In a cold, wet, or windy environment, cold exposure can be considerable for an injured or ill person. The subsequent autonomous stress response initially will increase circulatory and respiratory demands, and as body core temperature declines, the patient's condition might deteriorate. Therefore, the application of adequate insulation to reduce cold exposure and prevent body core cooling is an important part of prehospital primary care, but recommendations for what should be used in the field mostly depend on tradition and experience, not on scientific evidence. The objective of this study was to evaluate the thermal insulation properties in different wind conditions of 12 different blankets and rescue bags commonly used by prehospital rescue and ambulance services. The thermal manikin and the selected insulation ensembles were setup inside a climatic chamber in accordance to the modified European Standard for assessing requirements of sleeping bags. Fans were adjusted to provide low (value, Itr (m2 C/Wclo; where C = degrees Celcius, and W = watts), was calculated from ambient air temperature (C), manikin surface temperature (C), and heat flux (W/m2). In the low wind condition, thermal insulation of the evaluated ensembles correlated to thickness of the ensembles, ranging from 2.0 to 6.0 clo (1 clo = 0.155 m2 C/W), except for the reflective metallic foil blankets that had higher values than expected. In moderate and high wind conditions, thermal insulation was best preserved for ensembles that were windproof and resistant to the compressive effect of the wind, with insulation reductions down to about 60-80% of the original insulation capacity, whereas wind permeable and/or lighter materials were reduced down to about 30-50% of original insulation capacity. The evaluated insulation ensembles might all be used for prehospital protection against cold, either as single blankets or in multiple layer combinations, depending on ambient temperatures. However, with extended

  17. 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

  18. Thermal performance measurements of a 100 percent polyester MLI [multilayer insulation] system for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Boroski, W.N.; Gonczy, J.D.; Niemann, R.C.

    1989-09-01

    Thermal performance measurements of a 100 percent polyester multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) were conducted in a Heat Leak Test Facility (HLTF) under three experimental test arrangements. Each experiment measured the thermal performance of a 32-layer MLI blanket instrumented with twenty foil sensors to measure interstitial layer temperatures. Heat leak values and sensor temperatures were monitored during transient and steady state conditions under both design and degraded insulating vacuums. Heat leak values were measured using a heatmeter. MLI interstitial layer temperatures were measured using Cryogenic Linear Temperature Sensors (CLTS). Platinum resistors monitored system temperatures. High vacuum was measured using ion gauges; degraded vacuum employed thermocouple gauges. A four-wire system monitored instrumentation sensors and calibration heaters. An on-line computerized data acquisition system recorded and processes data. This paper reports on the instrumentation and experimental preparation used in carrying out these measurements. In complement with this paper is an associate paper bearing the same title head, but with the title extension 'Part 2: Laboratory results (300K--80K). 13 refs., 7 figs

  19. Code development for analysis of MHD pressure drop reduction in a liquid metal blanket using insulation technique based on a fully developed flow model

    International Nuclear Information System (INIS)

    Smolentsev, Sergey; Morley, Neil; Abdou, Mohamed

    2005-01-01

    The paper presents details of a new numerical code for analysis of a fully developed MHD flow in a channel of a liquid metal blanket using various insulation techniques. The code has specially been designed for channels with a 'sandwich' structure of several materials with different physical properties. The code includes a finite-volume formulation, automatically generated Hartmann number sensitive meshes, and effective convergence acceleration technique. Tests performed at Ha ∼ 10 4 have showed very good accuracy. As an illustration, two blanket flows have been considered: Pb-17Li flow in a channel with a silicon carbide flow channel insert, and Li flow in a channel with insulating coating

  20. The Leger house: an affordable super insulated home

    Energy Technology Data Exchange (ETDEWEB)

    Anon

    2000-03-01

    Building and energy efficient house in a cost-effective manner is not an easy task. The house described in this article is a 1200 sq.ft house, built on the principle of the Arkansas house, a sophisticated example of advanced wood frame construction, meets both these requirements, i.e. it is both energy efficient and good value for the money. It is built on a cast-in place concrete foundation with advanced framing construction techniques, staggered studs spaced on 24 inch centres, on separate plates forming an eight inch thick wall. The exterior sheathing is Dow Styrofoam, over which vinyl siding is installed. Thermal bridging, which is the weak point in conventional framing, received special attention. The outer walls are placed in such a way that they do not touch the floor joist and the plywood subfloor. Windows are wood framed, double glazed, a combination of casement and double hung, with storm windows, and sliding units. The clear span roof truss design gives the freedom of an open interior allowing a continuous unbroken ceiling vapour diffusion retarder. Drywall and oak flooring was installed without interference from partitions which were installed after the ceiling and the floor were in place. All lighting fixtures are on the partitions, so there is no need for penetration of the top plates in the partitions. Similarly, there is no wiring or plumbing in the exterior walls. Insulated steel doors, with magnetic weather stripping open to two airlocks. Domestic hot water and space heating is provided by a tank-less 43,800 Btu/hr Paloma natural gas heater. The water is kept at 160 degrees F and supplied to 46 feet of baseboard. A small conventional air-conditioning unit is installed but was only used for or a total of 12 hours since 1979, since during most of the hot summer days the rooms remain satisfactorily cool at about 80 degrees F, provided doors and windows are kept shut. A heat recovery ventilator has been installed subsequently, but is not considered

  1. Similarities between normal- and super-currents in topological insulator magnetic tunnel junctions

    International Nuclear Information System (INIS)

    Soodchomshom, Bumned; Chantngarm, Peerasak

    2010-01-01

    This work compares the normal-current in a NM/Fi/NM junction with the super-current in a SC/Fi/SC junction, where both are topological insulator systems. NM and Fi are normal region and ferromagnetic region of thickness d with exchange energy m playing a role of the mass of the Dirac electrons and with the gate voltage V G , respectively. SC is superconducting region induced by a s-wave superconductor. We show that, interestingly, the critical super-current passing through a SC/Fi/SC junction behaves quite similar to the normal-current passing through a NM/Fi/NM junction. The normal-current and super-current exhibit N-peak oscillation, found when currents are plotted as a function of the magnetic barrier strength χ ∼ md/hv F . With the barrier strength Z ∼ V G d/hv F , the number of peaks N is determined through the relation Z ∼ Nπ + σπ (with 0 < σ≤1 for χ < Z). The normal- and the super-currents also exhibit oscillating with the same height for all of peaks, corresponding to the Dirac fermion tunneling behavior. These anomalous oscillating currents due to the interplay between gate voltage and magnetic field in the barrier were not found in graphene-based NM/Fi/NM and SC/Fi/SC junctions. This is due to the different magnetic effect between the Dirac fermions in topological insulator and graphene.

  2. Beta cloth durability assessment for Space Station Freedom (SSF) Multi-Layer Insulation (MLI) blanket covers

    International Nuclear Information System (INIS)

    Koontz, S.L.; Jacobs, S.; Le, J.

    1993-03-01

    MLI blankets for the Space Station Freedom (SSF) must comply with general program requirements and recommendations for long life and durability in the low-Earth orbit (LEO) environment. Atomic oxygen and solar ultraviolet/vacuum ultraviolet are the most important factors in the SSF natural environment which affect materials life. Two types of Beta cloth (Teflon coated woven glass fabric), which had been proposed as MLI blanket covers, were tested for long-term durability in the LEO environment. General resistance to atomic oxygen attack and permeation were evaluated in the high velocity atomic oxygen beam system at Los Alamos National Laboratories. Long-term exposure to the LEO environment was simulated in the laboratory using a radio frequency oxygen plasma asher. The plasma asher treated Beta cloth specimens were tested for thermo-optical properties and mechanical durability. Space exposure data from the Long Duration Exposure Facility and the Intelsat Solar Array Coupon were also used in the durability assessment. Beta cloth fabricated to Rockwell specification MBO 135-027 (Chemglas 250) was shown to have acceptable durability for general use as an MLI blanket cover material in the LEO environment while Sheldahl G414500 should be used only in locations which are protected from direct Ram atomic oxygen

  3. Proposal for the award of a blanket purchase contract for the design, supply, installation and maintenance of automatic fire-detection, fire-protection and voice-alarm systems for the Super Proton Synchrotron

    CERN Document Server

    2017-01-01

    Proposal for the award of a blanket purchase contract for the design, supply, installation and maintenance of automatic fire-detection, fire-protection and voice-alarm systems for the Super Proton Synchrotron

  4. Highly Insulating and Light Transmitting Aerogel Glazing for Super Insulating Windows (HILIT+)

    DEFF Research Database (Denmark)

    Jensen, Karsten Ingerslev

    2005-01-01

    batch. Furthermore the production time has been reduced to 1/3 of the initial production time through detailed theoretical and experimental analyses of especially the supercritical washing step included in the drying phase. At the same time the production plant have been modified to recycle most...... insulation purposes. The edge seal solution shows only a very limited thermal bridge effect. The final glazing has a total solar energy transmittance above 85% and a U-value of 0.7 W/m2 K for about 14 mm aerogel thickness, which for a 20 mm thickness corresponds to a U-value of approximately 0.5 W/m2K...

  5. Electronic phase separation in insulating (Ga, Mn) As with low compensation: super-paramagnetism and hopping conduction

    Science.gov (United States)

    Yuan, Ye; Wang, Mao; Xu, Chi; Hübner, René; Böttger, Roman; Jakiela, Rafal; Helm, Manfred; Sawicki, Maciej; Zhou, Shengqiang

    2018-03-01

    In the present work, low compensated insulating (Ga,Mn)As with 0.7% Mn is obtained by ion implantation combined with pulsed laser melting. The sample shows variable-range hopping transport behavior with a Coulomb gap in the vicinity of the Fermi energy, and the activation energy is reduced by an external magnetic field. A blocking super-paramagnetism is observed rather than ferromagnetism. Below the blocking temperature, the sample exhibits a colossal negative magnetoresistance. Our studies confirm that the disorder-induced electronic phase separation occurs in (Ga,Mn)As samples with a Mn concentration in the insulator-metal transition regime, and it can account for the observed superparamagnetism and the colossal magnetoresistance.

  6. Development of ecological and economical super-insulations for various applications. Subproject 1: scientific development of ecological super-insulations for industrial application. Subproject 2: experimental synthesis and development of a pilot plant for continuously production and realisation of multilayer-insulation materials. Final report; Entwicklung oekologischer und wirtschaftlicher Super-Isolationen fuer vielfaeltige Anwendungen. Teilvorhaben 1: Wissenschaftliche Entwicklung oekologischer Super-Isolationen fuer industrielle Anwendungen. Teilvorhaben 2: Experimentelle Struktursynthese und Entwicklung einer Technikumsanlage zur kontinuierlichen Herstellung von Mehrschicht-Daemmstoffen. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Offermann, P.; Freudenberg, C.; Schenk, A.; Doerfel, A.; Hoffmann, G.; Roedel, H.; Schierz, C.; Hopf, W.

    2002-07-01

    Heat insulation materials are used in many applications with special tasks. Insulating materials like mineral wool, hard foams are used in civil engineering and for industrial insulation. Insulating materials from natural fibres are used in civil engineering on a small scale, too. In the clothing area are applied knitted fleece particularly for out-door-clothing in addition to non-woven made of synthetic polymers or wool. The aim of the project consists in the development of an insulating material with a very low heat conductivity and density as well as a multitude of degree of freedom to the structure and material parameters. A mathematical model has been developed for the determination of an optimised structure regarding to heat conductivity and density. The development was done by using the electrostatic flocking technology. After the material selection practical investigations have been done about the mode of function of the selected materials regarding their thermal insulation behaviour. A pilot plant for continuous production of the flocked material has been installed and tested. The result of this project is a very variable structure of insulating materials with excellent properties. The developed material is called Super-Insulation-Flock-Material (SIFM). Using defined structural parameters and skillfully selected materials it would be possible to get a heat conductivity between 0,027 W/mK and 0,30 W/mK. The density of these structures is between 10 kg/m{sup 3} and 20 kg/m{sup 3}. Structures with a density of only 7 kg/m{sup 3} are able to attend for applications without high mechanical demands. The Super-Insulation-Fock-Material is used in the clothing area and the technical sector. Sample products, e.g. a cold protective jacket, a jacket for fire fighters, insulation of airplanes as well as heat protective plates for the automotive industry, are found out. New fields for further applications of the Super-Insulation-Flock-Material result from the

  7. An overview of the multilayer insulation system for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Boroski, W.; Nicol, T.; Schoo, C.J.

    1991-08-01

    The MLI system for the SSC is designed to meet strict performance requirements over the 25 year life of the accelerator. Thermal measurements at 80K and 20K have been used to create an MLI system that limits heat flow to design values while incorporating features that permit the use of large-scale fabrication techniques. The result is a cost-effective means of mass-producing MLI blankets of consistent geometry and thermal performance

  8. Free-carrier-compensated charged domain walls produced with super-bandgap illumination in insulating ferroelectrics

    Czech Academy of Sciences Publication Activity Database

    Bednyakov, Petr; Sluka, T.; Tagantsev, A.; Damjanovic, D.; Setter, N.

    2016-01-01

    Roč. 28, č. 43 (2016), s. 9498-9503 ISSN 0935-9648 R&D Projects: GA ČR GA15-04121S Institutional support: RVO:68378271 Keywords : super-bandgap illumination * charged domain walls * ferroelectric BaTiO 3 * free-carrier generation Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 19.791, year: 2016

  9. 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

  10. Multi-scale cellulose based new bio-aerogel composites with thermal super-insulating and tunable mechanical properties.

    Science.gov (United States)

    Seantier, Bastien; Bendahou, Dounia; Bendahou, Abdelkader; Grohens, Yves; Kaddami, Hamid

    2016-03-15

    Bio-composite aerogels based on bleached cellulose fibers (BCF) and cellulose nanoparticles having various morphological and physico-chemical characteristics are prepared by a freeze-drying technique and characterized. The various composite aerogels obtained were compared to a BCF aerogel used as the reference. Severe changes in the material morphology were observed by SEM and AFM due to a variation of the cellulose nanoparticle properties such as the aspect ratio, the crystalline index and the surface charge density. BCF fibers form a 3D network and they are surrounded by the cellulose nanoparticle thin films inducing a significant reduction of the size of the pores in comparison with a neat BCF based aerogel. BET analyses confirm the appearance of a new organization structure with pores of nanometric sizes. As a consequence, a decrease of the thermal conductivities is observed from 28mWm(-1)K(-1) (BCF aerogel) to 23mWm(-1)K(-1) (bio-composite aerogel), which is below the air conductivity (25mWm(-1)K(-1)). This improvement of the insulation properties for composite materials is more pronounced for aerogels based on cellulose nanoparticles having a low crystalline index and high surface charge (NFC-2h). The significant improvement of their insulation properties allows the bio-composite aerogels to enter the super-insulating materials family. The characteristics of cellulose nanoparticles also influence the mechanical properties of the bio-composite aerogels. A significant improvement of the mechanical properties under compression is obtained by self-organization, yielding a multi-scale architecture of the cellulose nanoparticles in the bio-composite aerogels. In this case, the mechanical property is more dependent on the morphology of the composite aerogel rather than the intrinsic characteristics of the cellulose nanoparticles. Copyright © 2015 Elsevier Ltd. All rights reserved.

  11. 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

  12. 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.)

  13. Report on the program of 4 K irradiation of insulating materials for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Spindel, A.

    1993-07-01

    This report is intended to serve as an aid to material selection. The results reported herein are the product of a careful investigation and can be used with confidence in their validity. The selection of materials based on this data, however, is not the responsibility of the author. This report will not approve or disapprove any specific material for use in the Super Collider. The author of this report does not assume any design responsibility or responsibility for material selection for any application. It is, therefore, very important that those with design responsibility use this report wisely. For this reason, the following informational guide to the material selection process has been provided. There are several issues to take into account when evaluating a material for radiation resistance. It is very important that the design criteria and operating loads for the application be known. For many applications the actual loading, and therefore required properties, are unknown. Certain materials have empirically been used successfully in a similar application and those materials have often been selected on that basis. Both percent degradation and the magnitude of the actual properties after irradiation need to be considered. Consider the scenario where two materials are being compared that both have acceptable properties after exposure to 10 9 rads. It is preferable to choose the material with less degradation because degradation tends to be a threshold phenomena with properties declining rapidly with dose after a certain threshold dose. The properties of the initially strong material, therefore, will be extremely sensitive to dose in that dose range and slight magnet-to-magnet differences in dose may, depending on the application, lead to performance variations

  14. Super insulating aerogel glazing

    DEFF Research Database (Denmark)

    Schultz, Jørgen Munthe; Jensen, Karsten Ingerslev; Kristiansen, Finn Harken

    2005-01-01

    of the glass and a heat-treatment of the aerogel increases the visible quality and the solar energy transmittance. A low-conductive rim seal solution with the required vacuum barrier properties has been developed along with a reliable assembly and evacuation process. The prototypes have a centre heat loss...

  15. Proposal to negotiate an amendment to an exisiting contract for the supply of additional Multi-layer Insulation (MLI) blankets for the LHC

    CERN Document Server

    2006-01-01

    This document concerns the proposal to negotiate an amendment to an existing contract for the supply of additional sets of MLI blankets. For the reasons explained in this document, the Finance Committee is invited to approve an amendment to an existing contract with JEHIER (FR) for the supply of additional MLI blanket sets for an amount of 1 020 000 euros (1 625 000 Swiss francs), bringing the total to a maximum amount of 8 954 359 euros (14 266 084 Swiss francs), subject to revision for inflation. The amounts in Swiss francs have been calculated using the present rate of exchange.

  16. Proposal to negotiate an amendment to an existing contract for the supply of additional Multi-Layer Insulation (MLI) blankets for the LHC

    CERN Document Server

    2005-01-01

    This document concerns the proposal to negotiate an amendment to an existing contract for the supply of additional sets of MLI blankets. For the reasons explained in this document, the Finance Committee is invited to approve an amendment to an existing contract with JEHIER (FR) for the supply of an additional 2 420 MLI blankets for an amount of 1 863 400 euros (2 909 513 Swiss francs), subject to revision for inflation, bringing the total to a maximum amount of 7 934 359 euros (12 388 708 Swiss francs), subject to revision for inflation. The amounts in Swiss francs have been calculated using the present rate of exchange.

  17. Electrical insulators for the theta-pinch fusion reactor

    International Nuclear Information System (INIS)

    Clinard, F.W. Jr.

    1976-01-01

    The five major applications for electrical insulators in the Reference Theta Pinch Reactor are as follows: (1) first-wall insulator, (2) blanket intersegment insulator, (3) graphite encapsulating insulator, (4) implosion coil insulator, and (5) compression coil insulator. Insulator design proposals and some preliminary test results are given for each application

  18. Assessment of Cumulative Trauma Disorder (CTD) Risk for 3 Different Tasks Constructing and Repairing Multi-Layer Insulation (MLI) Blankets, Preparing the Dough for a Pizza, and Operating the Becton-Dickinson FACSAria Flow Cytometer

    Science.gov (United States)

    Gentzler, Marc; Kline, Martin; Palmer, Andrew; Terrone, Mark

    2007-01-01

    The Cumulative Trauma Disorder (CTD) risks for three different tasks using McCauley-Bell and Badiru's (1993) formula based on task, personal, and organizational factors were examined. For the Multi-Layer Insulation (MLI) blanket task, the results showed that the task, personal, and organizational risks were at about the same level. The personal risk factors for this task were evaluated using a hypothetical female employee age 52. For the pizza dough task, it was shown that the organizational risk was particularly high, with task related factors also at quite dangerous levels. On the other hand, there was a very low level of personal risk factors, based on a female age 17. The flow cytometer task was assessed with three different participants, a11 of whom had quite disparate levels of personal risk, which slightly affected the overall CTD risk. This reveals how individual difference variables certainly need to be considered. The task and organizational risks for this task were rated at about the same moderate level. The overall CTD risk averaged across the three participants was .335, indicating some risk. Compruing across the tasks revealed that the pizza dough task created the greatest overall CTD risk by far (.568), with the MLI (.325) and flow cytometer task (.335) having some risk associated with them. Future research should look into different tasks for more of a comparison

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.)

  5. 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.)

  6. A new method for characterizing super-insulators. Application to the identification of conduction modes; Nouvelle methode de caracterisation thermique des super-isolants. Application a l'identification des differents modes de conduction

    Energy Technology Data Exchange (ETDEWEB)

    Demange, D.; Bejet, M.; Laizet, J.C.

    2002-07-01

    In order to obtain thermal data necessary to the design of space systems, the ONERA developed a method to measure the thermal flux crossing an insulating structure under high thermal gradients. This method gives the thermal conductivity of material for an homogeneous composition of the structure. It allows the characterization of insulators under controlled atmosphere and at very high temperature, 2500 C. (A.L.B.)

  7. 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

  8. 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

  9. 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

  10. 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.)

  11. Polyester Apparel Cutting Waste as Insulation Material

    OpenAIRE

    Trajković, Dušan; Jordeva, Sonja; Tomovska, Elena; Zafirova, Koleta

    2017-01-01

    Polyester waste is the dominant component of the clothing industry waste stream, yet its recycling in this industry is rarely addressed. This paper proposes using polyester cutting waste as an insulation blanket for roofing and buildings’ internal walls in order to reduce environmental pollution. The designed textile structures used waste cuttings from different polyester fabrics without opening the fabric to fibre. Thermal insulation, acoustic insulation, fire resistance and biodegradation o...

  12. 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.).

  13. 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

  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. 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.

  16. 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

  17. Reusable Surface Insulation

    Science.gov (United States)

    1997-01-01

    Advanced Flexible Reusable Surface Insulation, developed by Ames Research Center, protects the Space Shuttle from the searing heat that engulfs it on reentry into the Earth's atmosphere. Initially integrated into the Space Shuttle by Rockwell International, production was transferred to Hi-Temp Insulation Inc. in 1974. Over the years, Hi-Temp has created many new technologies to meet the requirements of the Space Shuttle program. This expertise is also used commercially, including insulation blankets to cover aircrafts parts, fire barrier material to protect aircraft engine cowlings and aircraft rescue fire fighter suits. A Fire Protection Division has also been established, offering the first suit designed exclusively by and for aircraft rescue fire fighters. Hi-Temp is a supplier to the Los Angeles City Fire Department as well as other major U.S. civil and military fire departments.

  18. 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.

  19. Integrated Multilayer Insulation

    Science.gov (United States)

    Dye, Scott

    2009-01-01

    Integrated multilayer insulation (IMLI) is being developed as an improved alternative to conventional multilayer insulation (MLI), which is more than 50 years old. A typical conventional MLI blanket comprises between 10 and 120 metallized polymer films separated by polyester nets. MLI is the best thermal- insulation material for use in a vacuum, and is the insulation material of choice for spacecraft and cryogenic systems. However, conventional MLI has several disadvantages: It is difficult or impossible to maintain the desired value of gap distance between the film layers (and consequently, it is difficult or impossible to ensure consistent performance), and fabrication and installation are labor-intensive and difficult. The development of IMLI is intended to overcome these disadvantages to some extent and to offer some additional advantages over conventional MLI. The main difference between IMLI and conventional MLI lies in the method of maintaining the gaps between the film layers. In IMLI, the film layers are separated by what its developers call a micro-molded discrete matrix, which can be loosely characterized as consisting of arrays of highly engineered, small, lightweight, polymer (typically, thermoplastic) frames attached to, and placed between, the film layers. The term "micro-molded" refers to both the smallness of the frames and the fact that they are fabricated in a process that forms precise small features, described below, that are essential to attainment of the desired properties. The term "discrete" refers to the nature of the matrix as consisting of separate frames, in contradistinction to a unitary frame spanning entire volume of an insulation blanket.

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. High temperature blankets for the production of synthetic fuels

    International Nuclear Information System (INIS)

    Powell, J.R.; Steinberg, M.; Fillo, J.; Makowitz, H.

    1977-01-01

    The application of very high temperature blankets to improved efficiency of electric power generation and production of H 2 and H 2 based synthetic fuels is described. The blanket modules have a low temperature (300 to 400 0 C) structure (SS, V, Al, etc.) which serves as the vacuum/coolant pressure boundary, and a hot (>1000 0 C) thermally insulated interior. Approximately 50 to 70% of the fusion energy is deposited in the hot interior because of deep penetration by high energy neutrons. Separate coolant circuits are used for the two temperature zones: water for the low temperature structure, and steam or He for the hot interior. Electric generation efficiencies of approximately 60% and H 2 production efficiencies of approximately 50 to 70%, depending on design, are projected for fusion reactors using these high temperature blankets

  5. Insulating process for HT-7U central solenoid model coils

    International Nuclear Information System (INIS)

    Cui Yimin; Pan Wanjiang; Wu Songtao; Wan Yuanxi

    2003-01-01

    The HT-7U superconducting Tokamak is a whole superconducting magnetically confined fusion device. The insulating system of its central solenoid coils is critical to its properties. In this paper the forming of the insulating system and the vacuum-pressure-impregnating (VPI) are introduced, and the whole insulating process is verified under the super-conducting experiment condition

  6. 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.)

  7. High-insulated glass house, Egebjerggaard, Ballerup; Det hoejisolerede glashus. Egebjerggaard, Ballerup

    Energy Technology Data Exchange (ETDEWEB)

    Wittchen, K.B.; Aggerholm, S.

    1999-11-01

    New, super-insulating transparent and translucent glazing offers new perspectives for use of glass in architecture to achieve new facade idioms, spatial and light effects and low energy consumption. The new types of glazing are being tested in practice through the construction of a super-insulated glass house for Ballerup Ejendomsselskab in the district of Egebjerggaard west of Copenhagen. The project is based on SBI Report 220, Super-insulated glass houses (1993), in which use of new, super-insulating transparent and translucent glazing is analysed in relation to architecture, light conditions, indoor climate and energy consumption - for a detached house and a terraced house. (EHS)

  8. Dielectric and Insulating Technology 2004 : Review & Forecast

    Science.gov (United States)

    Okamoto, Tatsuki

    This article reports the state-of-art of DEIS activites. DEIS activiteis are basically based on the activites of 8-10 investigation committees’ under DEIS committee. Recent DEIS activites are categlized into three functions in this article and remarkable activity or trend of each category is mentioned. Those are activities on insulation diagnosis (AI application and asset management), activities on new insulation technology for power tansmission (high Tc super conducting cable insulation and all solid sinulated substation), and activities on new insulating materials (Nanocomposite).

  9. 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

  10. 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.

  11. Super jackstraws and super waterwheels

    International Nuclear Information System (INIS)

    Cho, Jin-Ho

    2007-01-01

    We construct various new BPS states of D-branes preserving 8 supersymmetries. These include super Jackstraws (a bunch of scattered D- or (p, q)-strings preserving supersymmetries), and super waterwheels (a number of D2-branes intersecting at generic angles on parallel lines while preserving supersymmetries). Super D-Jackstraws are scattered in various dimensions but are dynamical with all their intersections following a common null direction. Meanwhile, super (p, q)-Jackstraws form a planar static configuration. We show that the SO(2) subgroup of SL(2, R), the group of classical S-duality transformations in IIB theory, can be used to generate this latter configuration of variously charged (p, q)-strings intersecting at various angles. The waterwheel configuration of D2-branes preserves 8 supersymmetries as long as the 'critical' Born-Infeld electric fields are along the common direction

  12. Super differential forms on super Riemann surfaces

    International Nuclear Information System (INIS)

    Konisi, Gaku; Takahasi, Wataru; Saito, Takesi.

    1994-01-01

    Line integral on the super Riemann surface is discussed. A 'super differential operator' which possesses both properties of differential and of differential operator is proposed. With this 'super differential operator' a new theory of differential form on the super Riemann surface is constructed. We call 'the new differentials on the super Riemann surface' 'the super differentials'. As the applications of our theory, the existency theorems of singular 'super differentials' such as 'super abelian differentials of the 3rd kind' and of a super projective connection are examined. (author)

  13. 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

  14. Thermal insulation

    International Nuclear Information System (INIS)

    Aspden, G.J.; Howard, R.S.

    1988-01-01

    The patent concerns high temperature thermal insulation of large vessels, such as the primary vessel of a liquid metal cooled nuclear reactor. The thermal insulation consists of multilayered thermal insulation modules, and each module comprises a number of metal sheet layers sandwiched between a back and front plate. The layers are linked together by straps and clips to control the thickness of the module. (U.K.)

  15. Cellulose Insulation

    Science.gov (United States)

    1980-01-01

    Fire retardant cellulose insulation is produced by shredding old newspapers and treating them with a combination of chemicals. Insulating material is blown into walls and attics to form a fiber layer which blocks the flow of air. All-Weather Insulation's founders asked NASA/UK-TAP to help. They wanted to know what chemicals added to newspaper would produce an insulating material capable of meeting federal specifications. TAP researched the query and furnished extensive information. The information contributed to successful development of the product and helped launch a small business enterprise which is now growing rapidly.

  16. Synthesis and Characterization of Fibre Reinforced Silica Aerogel Blankets for Thermal Protection

    Directory of Open Access Journals (Sweden)

    S. Chakraborty

    2016-01-01

    Full Text Available Using tetraethoxysilane (TEOS as the source of silica, fibre reinforced silica aerogels were synthesized via fast ambient pressure drying using methanol (MeOH, trimethylchlorosilane (TMCS, ammonium fluoride (NH4F, and hexane. The molar ratio of TEOS/MeOH/(COOH2/NH4F was kept constant at 1 : 38 : 3.73 × 10−5 : 0.023 and the gel was allowed to form inside the highly porous meta-aramid fibrous batting. The wet gel surface was chemically modified (silylation process using various concentrations of TMCS in hexane in the range of 1 to 20% by volume. The fibre reinforced silica aerogel blanket was obtained subsequently through atmospheric pressure drying. The aerogel blanket samples were characterized by density, thermal conductivity, hydrophobicity (contact angle, and Scanning Electron Microscopy. The radiant heat resistance of the aerogel blankets was examined and compared with nonaerogel blankets. It has been observed that, compared to the ordinary nonaerogel blankets, the aerogel blankets showed a 58% increase in the estimated burn injury time and thus ensure a much better protection from heat and fire hazards. The effect of varying the concentration of TMCS on the estimated protection time has been examined. The improved thermal stability and the superior thermal insulation of the flexible aerogel blankets lead to applications being used for occupations that involve exposure to hazards of thermal radiation.

  17. 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

  18. Super families

    International Nuclear Information System (INIS)

    Amato, N.; Maldonado, R.H.C.

    1989-01-01

    The study on phenomena in the super high energy region, Σ E j > 1000 TeV revealed events that present a big dark spot in central region with high concentration of energy and particles, called halo. Six super families with halo were analysed by Brazil-Japan Cooperation of Cosmic Rays. For each family the lateral distribution of energy density was constructed and R c Σ E (R c ) was estimated. For studying primary composition, the energy correlation with particles released separately in hadrons and gamma rays was analysed. (M.C.K.)

  19. Thermal insulation

    International Nuclear Information System (INIS)

    Pinsky, G.P.

    1977-01-01

    Thermal insulation for vessels and piping within the reactor containment area of nuclear power plants is disclosed. The thermal insulation of this invention can be readily removed and replaced from the vessels and piping for inservice inspection, can withstand repeated wettings and dryings, and can resist high temperatures for long periods of time. 4 claims, 3 figures

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. Repeatability of Cryogenic Multilayer Insulation

    Science.gov (United States)

    Johnson, W. L.; Vanderlaan, M.; Wood, J. J.; Rhys, N. O.; Guo, W.; Van Sciver, S.; Chato, D. J.

    2017-12-01

    Due to the variety of requirements across aerospace platforms, and one off projects, the repeatability of cryogenic multilayer insulation (MLI) has never been fully established. The objective of this test program is to provide a more basic understanding of the thermal performance repeatability of MLI systems that are applicable to large scale tanks. There are several different types of repeatability that can be accounted for: these include repeatability between identical blankets, repeatability of installation of the same blanket, and repeatability of a test apparatus. The focus of the work in this report is on the first two types of repeatability. Statistically, repeatability can mean many different things. In simplest form, it refers to the range of performance that a population exhibits and the average of the population. However, as more and more identical components are made (i.e. the population of concern grows), the simple range morphs into a standard deviation from an average performance. Initial repeatability testing on MLI blankets has been completed at Florida State University. Repeatability of five Glenn Research Center (GRC) provided coupons with 25 layers was shown to be +/- 8.4% whereas repeatability of repeatedly installing a single coupon was shown to be +/- 8.0%. A second group of 10 coupons has been fabricated by Yetispace and tested by Florida State University, the repeatability between coupons has been shown to be +/- 15-25%. Based on detailed statistical analysis, the data has been shown to be statistically significant.

  7. 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)

  8. 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)

  9. 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

  10. 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)

  11. Moving ring field-reversed mirror blanket design considerations

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, L.; Kessel, C.; Norman, J.; Schultz, K.R.

    1981-01-01

    A blanket design for the Moving Ring Field-Reversed Mirror Reactor (MRFRM) is presented in this paper. The design emphasis is placed on minimizing the induced radioactivities in the first-wall, blanket and shield. To this end, aluminum-alloy was selected as the reference structural material, giving dose rates two weeks after shutdown that are 3 to 4 orders of magnitude lower than comparable steel structures. The aluminum first-wall is water-cooled and thermally insulated from the high temperature SiC-clad Li 2 O tritium breeding zone. A local tritium breeding ratio of 1.05 was obtained for the design. The tritium is extracted from the Li 2 O by the use of a small dry helium purge stream through the SiC tubes. About 1 ppM hydrogen is added to the helium purge stream to enhance the tritium recovery rate. Helium at 28 atmospheres pressure is circulated through the blanket and shield, with an outlet temperature of 850 0 C, which is coupled with an existing small size closed-cycle gas turbine (CCGT) power conversion system. The spatial and temporal variations of the first-wall temperature caused by the translational movement of the plasma rings along the axis of the cylindrical reactor were evaluated. The after-heat cooling problems of the first-wall were also considered

  12. 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

  13. The Super Patalan Numbers

    OpenAIRE

    Richardson, Thomas M.

    2014-01-01

    We introduce the super Patalan numbers, a generalization of the super Catalan numbers in the sense of Gessel, and prove a number of properties analagous to those of the super Catalan numbers. The super Patalan numbers generalize the super Catalan numbers similarly to how the Patalan numbers generalize the Catalan numbers.

  14. Blanket coverage : small technology companies hope to mine some Athabasca riches of their own

    Energy Technology Data Exchange (ETDEWEB)

    Marsters, S.

    2006-09-15

    This article presented details of the Pyrogel 6350, an insulating fabric comprised of a nanotechnology-enabled flexible fabric with aerogel integrated into its matrix. Aerogel is the lightest solid known to science and is created by replacing the liquid phase in a gel with gas. Originally developed for National American Space Agency (NASA) spacesuits, the United States military uses aerogel-based blankets to provide infrared suppression around engine compartments and hot components. Designed specifically for the oilsands market, the Pyrogel 6350 combines extreme thermal performance in a flexible blanket form which is ideal for the insulation of process equipment, pipelines and vessels used in oil sands production. The Pyrogel 6350 is currently being installed on 3 kilometres of high-pressure steam lines at Devon Canada's steam-assisted gravity drainage (SAGD) Jackfish project. The fabric's R-value per inch is 4 to 6 times greater than conventional types of insulation. It was concluded that use of the aerogel blankets will result in a 3 inch reduction in pipe insulation thickness, and provide significant savings in installation costs. 2 figs.

  15. 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

  16. 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

  17. Thermal insulation

    International Nuclear Information System (INIS)

    Durston, J.G.; Birch, W.; Facer, R.I.; Stuart, R.A.

    1977-01-01

    Reference is made to liquid metal cooled nuclear reactors. In the arrangement described the reactor vessel is clad with thermal insulation comprising a layer of insulating blocks spaced from the wall and from each other; each block is rigidly secured to the wall, and the interspaces are substantially closed against convectional flow of liquid by resilient closure members. A membrane covering is provided for the layer of blocks, with venting means to allow liquid from the reactor vessel to penetrate between the covering and the layer of blocks. The membrane covering may comprise a stainless steel sheet ribbed in orthogonal pattern to give flexibility for the accommodation of thermal strain. The insulating blocks may be comprised of stainless steel or cellular or porous material and may be hollow shells containing ceramic material or gas fillings. (U.K.)

  18. Lighter touch keeps in the heat. [Advantages of low-thermal-mass insulation

    Energy Technology Data Exchange (ETDEWEB)

    Pipes, A.

    1979-04-01

    Low-thermal-mass insulation of ceramic fibers and light refractory materials is more suitable to applications with intermittent processes and lower-temperature melting and retreating, where the heat-retention requirements do not require traditional furnace design. Old furnaces can be retrofitted by replacing bricks with insulation or by veneering. Insulating materials include ceramic, alumina, and quartz fibers, and microtherm in the form of blocks, blankets and other shapes. 4 figures. (DCK)

  19. Topological insulators

    CERN Document Server

    Franz, Marcel

    2013-01-01

    Topological Insulators, volume six in the Contemporary Concepts of Condensed Matter Series, describes the recent revolution in condensed matter physics that occurred in our understanding of crystalline solids. The book chronicles the work done worldwide that led to these discoveries and provides the reader with a comprehensive overview of the field. Starting in 2004, theorists began to explore the effect of topology on the physics of band insulators, a field previously considered well understood. However, the inclusion of topology brings key new elements into this old field. Whereas it was

  20. 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)

  1. A super soliton connection

    International Nuclear Information System (INIS)

    Gurses, M.; Oguz, O.

    1985-07-01

    Integrable super non-linear classical partial differential equations are considered. A super s1(2,R) algebra valued connection 1-form is constructed. It is shown that curvature 2-form of this super connection vanishes by virtue of the integrable super equations of motion. A super extension of the AKNS scheme is presented and a class of super extension of the Lax hierarchy and super non-linear Schroedinger equation are found. O(N) extension and the Baecklund transformations of the above super equations are also considered. (author)

  2. Optimisation of Multilayer Insulation an Engineering Approach

    CERN Document Server

    Chorowski, M; Parente, C; Riddone, G

    2001-01-01

    A mathematical model has been developed to describe the heat flux through multilayer insulation (MLI). The total heat flux between the layers is the result of three distinct heat transfer modes: radiation, residual gas conduction and solid spacer conduction. The model describes the MLI behaviour considering a layer-to-layer approach and is based on an electrical analogy, in which the three heat transfer modes are treated as parallel thermal impedances. The values of each of the transfer mode vary from layer to layer, although the total heat flux remains constant across the whole MLI blanket. The model enables the optimisation of the insulation with regard to different MLI parameters, such as residual gas pressure, number of layers and boundary temperatures. The model has been tested with experimental measurements carried out at CERN and the results revealed to be in a good agreement, especially for insulation vacuum between 10-5 Pa and 10-3 Pa.

  3. Acoustic Design of Super-light Structures

    DEFF Research Database (Denmark)

    Christensen, Jacob Ellehauge; Hertz, Kristian Dahl; Brunskog, Jonas

    in a controlled laboratory environment have been conducted with the element in order to evaluate its performance in airborne and impact sound insulation. These results have been employed in simulations of the flanking transmission to estimate the in-situ performance of the super-light slab element. The flanking...... aggregate (leca) along with a newly developed technology called pearl-chain reinforcement, which is a system for post-tensioning. Here, it is shown how to combine these technologies within a precast super-light slab element, while honoring the requirements of a holistic design. Acoustic experiments...

  4. Super Dielectric Materials

    Directory of Open Access Journals (Sweden)

    Samuel Fromille

    2014-12-01

    Full Text Available Evidence is provided here that a class of materials with dielectric constants greater than 105 at low frequency (<10−2 Hz, herein called super dielectric materials (SDM, can be generated readily from common, inexpensive materials. Specifically it is demonstrated that high surface area alumina powders, loaded to the incipient wetness point with a solution of boric acid dissolved in water, have dielectric constants, near 0 Hz, greater than 4 × 108 in all cases, a remarkable increase over the best dielectric constants previously measured for energy storage capabilities, ca. 1 × 104. It is postulated that any porous, electrically insulating material (e.g., high surface area powders of silica, titania, etc., filled with a liquid containing a high concentration of ionic species will potentially be an SDM. Capacitors created with the first generated SDM dielectrics (alumina with boric acid solution, herein called New Paradigm Super (NPS capacitors display typical electrostatic capacitive behavior, such as increasing capacitance with decreasing thickness, and can be cycled, but are limited to a maximum effective operating voltage of about 0.8 V. A simple theory is presented: Water containing relatively high concentrations of dissolved ions saturates all, or virtually all, the pores (average diameter 500 Å of the alumina. In an applied field the positive ionic species migrate to the cathode end, and the negative ions to the anode end of each drop. This creates giant dipoles with high charge, hence leading to high dielectric constant behavior. At about 0.8 V, water begins to break down, creating enough ionic species to “short” the individual water droplets. Potentially NPS capacitor stacks can surpass “supercapacitors” in volumetric energy density.

  5. 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

  6. 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

  7. Load responsive multilayer insulation performance testing

    International Nuclear Information System (INIS)

    Dye, S.; Kopelove, A.; Mills, G. L.

    2014-01-01

    Cryogenic insulation designed to operate at various pressures from one atmosphere to vacuum, with high thermal performance and light weight, is needed for cryogenically fueled space launch vehicles and aircraft. Multilayer insulation (MLI) performs well in a high vacuum, but the required vacuum shell for use in the atmosphere is heavy. Spray-on foam insulation (SOFI) is often used in these systems because of its light weight, but can have a higher heat flux than desired. We report on the continued development of Load Responsive Multilayer Insulation (LRMLI), an advanced thermal insulation system that uses dynamic beam discrete spacers that provide high thermal performance both in atmosphere and vacuum. LRMLI consists of layers of thermal radiation barriers separated and supported by micromolded polymer spacers. The spacers have low thermal conductance, and self-support a thin, lightweight vacuum shell that provides internal high vacuum in the insulation. The dynamic load responsive spacers compress to support the external load of a vacuum shell in one atmosphere, and decompress under reduced atmospheric pressure for lower heat leak. Structural load testing was performed on the spacers with various configurations. LRMLI was installed on a 400 liter tank and boil off testing with liquid nitrogen performed at various chamber pressures from one atmosphere to high vacuum. Testing was also performed with an MLI blanket on the outside of the LRMLI

  8. Load responsive multilayer insulation performance testing

    Energy Technology Data Exchange (ETDEWEB)

    Dye, S.; Kopelove, A. [Quest Thermal Group, 6452 Fig Street Suite A, Arvada, CO 80004 (United States); Mills, G. L. [Ball Aerospace and Technologies Corp, 1600 Commerce Street, Boulder, CO 80301 (United States)

    2014-01-29

    Cryogenic insulation designed to operate at various pressures from one atmosphere to vacuum, with high thermal performance and light weight, is needed for cryogenically fueled space launch vehicles and aircraft. Multilayer insulation (MLI) performs well in a high vacuum, but the required vacuum shell for use in the atmosphere is heavy. Spray-on foam insulation (SOFI) is often used in these systems because of its light weight, but can have a higher heat flux than desired. We report on the continued development of Load Responsive Multilayer Insulation (LRMLI), an advanced thermal insulation system that uses dynamic beam discrete spacers that provide high thermal performance both in atmosphere and vacuum. LRMLI consists of layers of thermal radiation barriers separated and supported by micromolded polymer spacers. The spacers have low thermal conductance, and self-support a thin, lightweight vacuum shell that provides internal high vacuum in the insulation. The dynamic load responsive spacers compress to support the external load of a vacuum shell in one atmosphere, and decompress under reduced atmospheric pressure for lower heat leak. Structural load testing was performed on the spacers with various configurations. LRMLI was installed on a 400 liter tank and boil off testing with liquid nitrogen performed at various chamber pressures from one atmosphere to high vacuum. Testing was also performed with an MLI blanket on the outside of the LRMLI.

  9. 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)

  10. 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

  11. Low activity blanket designs and heat transfer for experimental power reactors

    International Nuclear Information System (INIS)

    Fillo, J.; Tichler, P.; Lazareth, O.; Powell, J.

    1976-01-01

    Two minimum activity blanket designs are described, based on the ANL TEPR circular design parameters. A first wall loading (plasma on) of 1.0 MW(th)/m 2 has been assumed. The first option is composed of SAP (sintered aluminum product) modules. The oval shaped SAP shell, in which approximately 45 percent of the fusion energy is removed, is maintained at a temperature of approximately 400 0 C by a He coolant stream. The remaining 55 percent of the fusion energy is deposited in a thermally insulated hot interior (SiC and B 4 C) and removed by a separate He coolant, with exit temperature of 800 0 C. In the second option, the blanket is a thick graphite block structure (approximately 50 cm thickness) with SAP coolant tubes carrying He (50 atm) embedded deep within the graphite to minimize radiation damage. The neutron and gamma energy deposited in the graphite is radiated along internal slots and conducted through the graphite to the coolant tubes. To reduce surface evaporation above 2000 0 C, the blanket surface is radiatively cooled to a low temperature radiation sink, a bank of He cooled SAP tubes. Approximately 20 percent of the fusion energy is removed in this region, the remaining 80 percent in the primary graphite-aluminum blanket. Both blanket options are mounted on heavy Al backing plates, cooled by He, which are in turn supported from the fixed shield

  12. 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

  13. White Dwarf Model Atmospheres: Synthetic Spectra for Super Soft Sources

    OpenAIRE

    Rauch, Thomas

    2011-01-01

    The T\\"ubingen NLTE Model-Atmosphere Package (TMAP) calculates fully metal-line blanketed white dwarf model atmospheres and spectral energy distributions (SEDs) at a high level of sophistication. Such SEDs are easily accessible via the German Astrophysical Virtual Observatory (GAVO) service TheoSSA. We discuss applications of TMAP models to (pre) white dwarfs during the hottest stages of their stellar evolution, e.g. in the parameter range of novae and super soft sources.

  14. Super power generators

    International Nuclear Information System (INIS)

    Martin, T.H.; Johnson, D.L.; McDaniel, D.H.

    1977-01-01

    PROTO II, a super power generator, is presently undergoing testing at Sandia Laboratories. It has operated with an 80 ns, 50 ns, 35 ns, and 20 ns positive output pulse high voltage mode and achieved total current rates of rise of 4 x 10 14 A/s. The two sided disk accelerator concept using two diodes has achieved voltages of 1.5 MV and currents of 4.5 MA providing a power exceeding 6 TW in the electron beam and 8 TW in the transmission lines. A new test bed named MITE (Magnetically Insulated Transmission Experiment) was designed and is now being tested. The pulse forming lines are back to back short pulse Blumleins which use untriggered water switching. Output data showing a ten ns half width power pulse peaking above one terrawatt were obtained. MITE is a module being investigated for use in the Electron Beam Fusion Accelerator and will be used to test the effects of short pulses propagating down vacuum transmission lines

  15. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-02-01

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17LI is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Ph-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure.

  16. Basic principles of lead and lead-bismuth eutectic application in blanket of fusion reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Pinaev, S.S.; Muraviev, E.V.; Romanov, P.V.

    2005-01-01

    High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease magnetohydrodynamic resistance authors propose to form insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the insulating coatings characteristics ρδ is ∼ 10 -5 Ohm·m 2 for steels and 5,0x10 -6 - 5,0x10 -5 Ohm·m 2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steamgenerators and equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem of technology of lead and lead-bismuth coolants for power high temperature radioactive facilities has been solved. Accidents, emergency situations such as leakage of steamgenerators or depressurization of gas system in facilities with lead and lead-bismuth coolants have been explored and suppressed. (author)

  17. (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

  18. 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.)

  19. 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

  20. 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)

  1. Tank Insulation

    Science.gov (United States)

    1979-01-01

    For NASA's Apollo program, McDonnell Douglas Astronautics Company, Huntington Beach, California, developed and built the S-IVB, uppermost stage of the three-stage Saturn V moonbooster. An important part of the development task was fabrication of a tank to contain liquid hydrogen fuel for the stage's rocket engine. The liquid hydrogen had to be contained at the supercold temperature of 423 degrees below zero Fahrenheit. The tank had to be perfectly insulated to keep engine or solar heat from reaching the fuel; if the hydrogen were permitted to warm up, it would have boiled off, or converted to gaseous form, reducing the amount of fuel available to the engine. McDonnell Douglas' answer was a supereffective insulation called 3D, which consisted of a one-inch thickness of polyurethane foam reinforced in three dimensions with fiberglass threads. Over a 13-year development and construction period, the company built 30 tanks and never experienced a failure. Now, after years of additional development, an advanced version of 3D is finding application as part of a containment system for transporting Liquefied Natural Gas (LNG) by ship.

  2. 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.)

  3. 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....

  4. 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

  5. Status of ITER blanket attachment design and related R and D

    International Nuclear Information System (INIS)

    Sadakov, S.; Khomiakov, S.; Calcagno, B.; Chappuis, Ph.; Dellopoulos, G.; Kolganov, V.; Merola, M.; Poddubnyi, I.; Raffray, R.; Raharijaona, J.J.; Ulrickson, M.; Zhmakin, A.

    2013-01-01

    Highlights: • ITER blanket attachment system went through a significant design upgrade and become basically compliant with specified design loads and required cyclic lifetime. • Upgrade of flexible supports allowed the doubling of cross sections of central bolts. Ceramic coatings were relocated to much larger areas on conical pairs screwed into shield blocks. • Key pads were relocated from keys of vacuum vessel into keyways of shield blocks and reshaped to enlarge areas of lateral interfaces with ceramic electro-insulating coatings. • Ceramic coatings are hidden between pads and enclosures in keyways with a purpose to minimize their wear rate, which depends on peak friction stress and cyclic sliding path. • Ceramic coatings to be verified by experiment, with several R and D aimed to collect statistically sufficient data on their reliability and durability in ITER relevant cyclic loading conditions. -- Abstract: Main function of the ITER blanket system [1–3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R and D is also briefly described

  6. Design of Multilayer Insulation for the Multipurpose Hydrogen Test Bed

    Science.gov (United States)

    Marlow, Weston A.

    2011-01-01

    Multilayer insulation (MLI) is a critical component for future, long term space missions. These missions will require the storage of cryogenic fuels for extended periods of time with little to no boil-off and MLI is vital due to its exceptional radiation shielding properties. Several MLI test articles were designed and fabricated which explored methods of assembling and connecting blankets, yielding results for evaluation. Insight gained, along with previous design experience, will be used in the design of the replacement blanket for the Multipurpose Hydrogen Test Bed (MHTB), which is slated for upcoming tests. Future design considerations are discussed which include mechanical testing to determine robustness of such a system, as well as cryostat testing of samples to give insight to the loss of thermal performance of sewn panels in comparison to the highly efficient, albeit laborious application of the original MHTB blanket.

  7. 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

    Lithium is often the preferred choice as breeder and coolant in fusion blankets as it offers excellent heat transfer and corrosion properties, and most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and exacerbates plant safety concerns. For this reason, over the years numerous blanket concepts have been proposed with the scope of reducing concerns associated with lithium. The European helium cooled pebble bed breeding blanket (HCPB) physically confines lithium within ceramic pebbles. The pebbles reside within a low activation martensitic ferritic steel structure and are cooled by helium. The blanket is composed of the tritium breeding lithium ceramic pebbles and neutron multiplying beryllium pebbles. Other blanket designs utilize lead to lower chemical reactivity; LiPb alone can serve as a breeder, coolant, neutron multiplier, and tritium carrier. Blankets employing LiPb coolants alongside silicon carbide structural components can achieve high plant efficiency, low afterheat, and low operation pressures. This alloy can also be used alongside of helium such as in the dual-coolant lead-lithium concept (DCLL); helium is utilized to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket. The helium-cooled steel and lead-lithium alloy are separated by flow channel inserts (usually made out of silicon carbide) which thermally insulate the self-cooled breeder region from the helium cooled steel walls. This creates a LiPb breeder with a much higher exit temperature than the steel which increases the power cycle efficiency and also lowers the magnetohydrodynamic (MHD) pressure drop [6]. Molten salt blankets with a mixture of lithium, beryllium, and fluorides (FLiBe) offer good tritium breeding

  8. Simulation of flanking transmission in super-light structures for airborne and impact sound

    DEFF Research Database (Denmark)

    Christensen, Jacob Ellehauge; Hertz, Kristian Dahl; Brunskog, Jonas

    2012-01-01

    . Previously the airborne and impact sound insulation has been measured for a super-light deck element in a laboratory. This paper presents a flanking transmission analysis based on the measured results and are carried out for the Super-light deck elements by means of the acoustical software Bastian...... to design buildings with super-light deck elements while achieving a good acoustical environment in the building, fulfilling various acoustical requirements from the building regulations....

  9. 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)

  10. 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

  11. 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.)

  12. 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.

  13. 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

  14. 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)

  15. Cryogenic Testing of Different Seam Concepts for Multilayer Insulation Systems

    Science.gov (United States)

    Johnson, Wesley L.; Fesmire, J. E.

    2009-01-01

    Recent testing in a cylindrical, comparative cryostat at the Cryogenics Test Laboratory has focused on various seam concepts for multilayer insulation systems. Three main types of seams were investigated: straight overlap, fold-over, and roll wrapped. Each blanket was comprised of 40 layer pairs of reflector and spacer materials. The total thickness was approximately 12.5-mm, giving an average layer density of 32 layers per centimeter. The blankets were tested at high vacuum, soft vacuum, and no vacuum using liquid nitrogen to maintain the cold boundary temperature at 77 K. Test results show that all three seam concepts are all close in thermal performance; however the fold-over method provides the lowest heat flux. For the first series of tests, seams were located 120 degrees around the circumference of the cryostat from the previous seam. This technique appears to have lessened the degradation of the blanket due to the seams. In a follow-on test, a 20 layer blanket was tested in a roll wrapped configuration and then cut down the side of the cylinder, taped together, and re-tested. This test result shows the thermal performance impact of having the seams all in one location versus having the seams clocked around the vessel. This experimental investigation indicates that the method of joining the seams in multilayer insulation systems is not as critical as the quality of the installation process.

  16. 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.

  17. 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.)

  18. 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.

  19. 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

  20. 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

  1. 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)

  2. 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)

  3. Aerogel-Based Insulation for Industrial Steam Distribution Systems

    Energy Technology Data Exchange (ETDEWEB)

    John Williams

    2011-03-30

    Thermal losses in industrial steam distribution systems account for 977 trillion Btu/year in the US, more than 1% of total domestic energy consumption. Aspen Aerogels worked with Department of Energy’s Industrial Technologies Program to specify, develop, scale-up, demonstrate, and deliver Pyrogel XT®, an aerogel-based pipe insulation, to market to reduce energy losses in industrial steam systems. The product developed has become Aspen’s best selling flexible aerogel blanket insulation and has led to over 60 new jobs. Additionally, this product has delivered more than ~0.7 TBTU of domestic energy savings to date, and could produce annual energy savings of 149 TBTU by 2030. Pyrogel XT’s commercial success has been driven by it’s 2-4X better thermal performance, improved durability, greater resistance to corrosion under insulation (CUI), and faster installation times than incumbent insulation materials.

  4. Super Dielectric Materials.

    Science.gov (United States)

    Fromille, Samuel; Phillips, Jonathan

    2014-12-22

    Evidence is provided here that a class of materials with dielectric constants greater than 10⁵ at low frequency (dielectric materials (SDM), can be generated readily from common, inexpensive materials. Specifically it is demonstrated that high surface area alumina powders, loaded to the incipient wetness point with a solution of boric acid dissolved in water, have dielectric constants, near 0 Hz, greater than 4 × 10⁸ in all cases, a remarkable increase over the best dielectric constants previously measured for energy storage capabilities, ca. 1 × 10⁴. It is postulated that any porous, electrically insulating material (e.g., high surface area powders of silica, titania, etc. ), filled with a liquid containing a high concentration of ionic species will potentially be an SDM. Capacitors created with the first generated SDM dielectrics (alumina with boric acid solution), herein called New Paradigm Super (NPS) capacitors display typical electrostatic capacitive behavior, such as increasing capacitance with decreasing thickness, and can be cycled, but are limited to a maximum effective operating voltage of about 0.8 V. A simple theory is presented: Water containing relatively high concentrations of dissolved ions saturates all, or virtually all, the pores (average diameter 500 Å) of the alumina. In an applied field the positive ionic species migrate to the cathode end, and the negative ions to the anode end of each drop. This creates giant dipoles with high charge, hence leading to high dielectric constant behavior. At about 0.8 V, water begins to break down, creating enough ionic species to "short" the individual water droplets. Potentially NPS capacitor stacks can surpass "supercapacitors" in volumetric energy density.

  5. Magnetohydrodynamic flow in ducts with discontinuous electrical insulation

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Bühler, L.

    2015-01-01

    Highlights: • Liquid metal MHD flows in ducts with flow channel inserts. • Study of the influence of local interruption of electrical insulation. • 3D numerical simulations. - Abstract: In liquid metal blankets the interaction of the moving breeder with the intense magnetic field that confines the fusion plasma results in significant modifications of the velocity distribution and increased pressure drop compared to hydrodynamic flows. Those changes are due to the occurrence of electromagnetic forces that slow down the core flow and which are balanced by large driving pressure heads. The resulting magnetohydrodynamic (MHD) pressure losses are proportional to the electric current density induced in the fluid and they can be reduced by electrically decoupling the wall from the liquid metal. For applications to dual coolant blankets it is foreseen to loosely insert electrically insulating liners into the ducts. In long channels the insulation could consist of a number of shorter inserts, which implies a possible local interruption of the insulation. Three dimensional numerical simulations have been performed to investigate MHD flows in electrically well-conducting channels with internal discontinuous insulating inserts. The local jump in the electric conductivity of the duct wall results in induced 3D electric currents and related electromagnetic forces yielding additional pressure losses and increased velocity in boundary layers parallel to the magnetic field.

  6. A Wideroee pre-accelerator for the SuperHILAC

    International Nuclear Information System (INIS)

    Staples, J.; Alonso, J.; Behrsing, G.; Clark, D.; Grunder, H.; Olivier, M.; Spence, D.; Yourd, R.

    1976-01-01

    Plans to upgrade both the Bevatron vacuum system and the SuperHILAC ion sources and injectors have been formulated. A proposed new pre-accelerator based on an air-insulated Cockcroft-Walton and a Wideroee linac is presented

  7. 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.)

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

  14. Comparison of lithium and the eutectic lead lithium alloy, two candidate liquid metal breeder materials for self-cooled blankets

    International Nuclear Information System (INIS)

    Malang, S.; Mattas, R.

    1994-06-01

    Liquid metals are attractive candidates for both near-term and long-term fusion applications. The subjects of this comparison are the differences between the two candidate liquid metal breeder materials Li and LiPb for use in breeding blankets in the areas of neutronics, magnetohydrodynamics, tritium control, compatibility with structural materials, heat extraction system, safety, and required R ampersand D program. Both candidates appear to be promising for use in self-cooled breeding blankets which have inherent simplicity with the liquid metal serving as both breeders and coolant. The remaining feasibility question for both breeder materials is the electrical insulation between liquid metal and duct walls. Different ceramic coatings are required for the two breeders, and their crucial issues, namely self-healing of insulator cracks and radiation induced electrical degradation are not yet demonstrated. Each liquid metal breeder has advantages and concerns associated with it, and further development is needed to resolve these concerns

  15. Magnetohydrodynamic (MHD) considerations for liquid metal blanket and a SiC/SiC composite structure

    International Nuclear Information System (INIS)

    Scholz, R.; Greeff, J. de; Vinche, C.

    1998-01-01

    The electrical conductivity was measured on SiC/SiC composite specimens, in the as-received conditions and after neutron irradiation, for temperatures between 20 deg. C and 1000 deg. C. The tests were aimed at estimating the magnitude of MHD effects in liquid metal blankets and a SiC/SiC composites structure. The electrical conductivity of the unirradiated samples increased continuously with temperature and ranged from 330 (Ω m) -1 at 20 deg. C to 550 (Ω m) -1 at 1000 deg.C. The irradiation reduced only slightly the magnitude of σ indicating the materials tested cannot be treated as an electrical insulator in a MHD analysis for liquid metal blankets. (authors)

  16. Magnetohydrodynamic (MHD) considerations for liquid metal blanket and a SiC/SiC composite structure

    Energy Technology Data Exchange (ETDEWEB)

    Scholz, R.; Greeff, J. de; Vinche, C. [Commission Europeenne Community, JRC, Vatican City State, Holy See (Italy)

    1998-07-01

    The electrical conductivity was measured on SiC/SiC composite specimens, in the as-received conditions and after neutron irradiation, for temperatures between 20 deg. C and 1000 deg. C. The tests were aimed at estimating the magnitude of MHD effects in liquid metal blankets and a SiC/SiC composites structure. The electrical conductivity of the unirradiated samples increased continuously with temperature and ranged from 330 ({omega} m){sup -1} at 20 deg. C to 550 ({omega} m){sup -1} at 1000 deg.C. The irradiation reduced only slightly the magnitude of {sigma} indicating the materials tested cannot be treated as an electrical insulator in a MHD analysis for liquid metal blankets. (authors)

  17. Qualification of MHD effects in dual-coolant DEMO blanket and approaches to their modelling

    International Nuclear Information System (INIS)

    Mas de les Valls, E.; Batet, L.; Medina, V. de; Fradera, J.; Sedano, L.A.

    2011-01-01

    Design refinements of vertical insulated banana-shaped liquid metal channels are being considered as a progress of conceptual design of dual-coolant liquid metal blankets (DEMO specifications). Among them: (a) optimised channel geometry and (b) improvements on flow channel inserts. Progress of channel conceptual design is conducted in parallel with underlying physics of MHD models in diverse aspects: (1) MHD models, (2) MHD turbulence, (3) LM buoyancy effects, (4) three-dimensional flows, and (5) LM/FCI/wall electrical and thermal coupling; in order to progress on common liquid metal flow characterisation, pressure drop and three-dimensional flows. The analyses are assumed as extension of those previous carried out for the DCLL blankets for new design refinements. At the present stage of the conceptual design progress, a preliminary thermofluid MHD study is of crucial interest for further design improvements and future detailed modelling. The paper overviews the ongoing modelling studies, making model refinements explicit, and anticipates some modelling results.

  18. 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

  19. Summary report for ITER task - T68: MHD facility preparation for Li/V blanket option

    International Nuclear Information System (INIS)

    Reed, C.B.; Haglund, R.C.; Miller, M.E.

    1995-08-01

    A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the question of insulator coatings. Design calculations show that an electrically insulating layer is necessary to maintain an acceptably low MHD pressure drop. To enable experimental investigations of the MHD performance of candidate insulator materials and the technology for putting them in place, the room-temperature ALEX (Argonne's Liquid Metal EXperiment) NaK facility was upgraded to a 300 degrees C lithium system. The objective of this upgrade was to modify the existing facility to the minimum extent necessary, consistent with providing a safe, flexible, and easy to operate MHD test facility which uses lithium at ITER-relevant temperatures, Hartmann numbers, and interaction parameters. The facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups. The system design description for this lithium upgrade of the ALEX facility is given in this document

  20. Super Riemann surfaces

    International Nuclear Information System (INIS)

    Rogers, Alice

    1990-01-01

    A super Riemann surface is a particular kind of (1,1)-dimensional complex analytic supermanifold. From the point of view of super-manifold theory, super Riemann surfaces are interesting because they furnish the simplest examples of what have become known as non-split supermanifolds, that is, supermanifolds where the odd and even parts are genuinely intertwined, as opposed to split supermanifolds which are essentially the exterior bundles of a vector bundle over a conventional manifold. However undoubtedly the main motivation for the study of super Riemann surfaces has been their relevance to the Polyakov quantisation of the spinning string. Some of the papers on super Riemann surfaces are reviewed. Although recent work has shown all super Riemann surfaces are algebraic, some areas of difficulty remain. (author)

  1. Thermally tunable broadband omnidirectional and polarization-independent super absorber using phase change material VO2

    Directory of Open Access Journals (Sweden)

    Zhejun Liu

    Full Text Available In this letter, we numerically demonstrate a thermally tunable super absorber by using phase change material VO2 as absorbing layer in metal-insulator-metal structure. An omnidirectional super absorption at λ=2.56μm can be realized by heating the patterned grating VO2 film due to magnetic resonance mechanism. Furthermore, a broadband super absorption higher than 0.8 in the entire 1.6μm–4μm region is achieved when VO2 film is patterned chessboard structure and transformed to metal phase beyond transition temperature. This broadband super absorption can be fulfilled in a wide range of incident angle (0°–70° and under all polarization conditions. Keywords: Phase change material, Metal-insulator-metal, Super absorption, Magnetic resonance

  2. Supermanifolds and super Riemann surfaces

    International Nuclear Information System (INIS)

    Rabin, J.M.

    1986-09-01

    The theory of super Riemann surfaces is rigorously developed using Rogers' theory of supermanifolds. The global structures of super Teichmueller space and super moduli space are determined. The super modular group is shown to be precisely the ordinary modular group. Super moduli space is shown to be the gauge-fixing slice for the fermionic string path integral

  3. Calculus super review

    CERN Document Server

    2012-01-01

    Get all you need to know with Super Reviews! Each Super Review is packed with in-depth, student-friendly topic reviews that fully explain everything about the subject. The Calculus I Super Review includes a review of functions, limits, basic derivatives, the definite integral, combinations, and permutations. Take the Super Review quizzes to see how much you've learned - and where you need more study. Makes an excellent study aid and textbook companion. Great for self-study!DETAILS- From cover to cover, each in-depth topic review is easy-to-follow and easy-to-grasp - Perfect when preparing for

  4. Algebra & trigonometry super review

    CERN Document Server

    2012-01-01

    Get all you need to know with Super Reviews! Each Super Review is packed with in-depth, student-friendly topic reviews that fully explain everything about the subject. The Algebra and Trigonometry Super Review includes sets and set operations, number systems and fundamental algebraic laws and operations, exponents and radicals, polynomials and rational expressions, equations, linear equations and systems of linear equations, inequalities, relations and functions, quadratic equations, equations of higher order, ratios, proportions, and variations. Take the Super Review quizzes to see how much y

  5. Nuclear design of the blanket/shield system for a Tokamak Experimental Power Reactor

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1976-01-01

    The various options and trade-offs in the nuclear design of the blanket/shield for a Tokamak Experimental Power Reactor (TEPR) are investigated. The TEPR size and cost are particularly sensitive to the blanket/shield thickness, Δ/sub BS/, on the inner side of the torus. Radition damage to the components of the superconducting magnet and refrigeration power requirements set lower limits on Δ/sub BS/. These limits are developed in terms of TEPR design parameters such as the wall loading, duty cycle, and frequency of magnet anneals. The study of the nuclear performance of various material compositions shows that mixtures of tungsten, or tantalum, or stainless-steel alloys and boron carbide require the smallest Δ/sub BS/ for a given attenuation. This Δ/sub BS/ has to be doubled if the low induced activation materials graphite and aluminum are used. The space problems are greatly eased in the Argonne National Laboratory ANL-TEPR reference design by using two separate segments of the blanket/shield. The inner segment occupies the region of the high magnetic field, uses very efficient attenuators (tungsten- or tantalum- or stainless-steel-boron carbide mixtures), and is only 1 m thick. The outer blanket/shield is 131 cm and consists of an optimized composition of stainless steel and boron carbide. For the design parameters of 0.2 MW/m 2 neutron wall loading and 50 percent duty cycle, the reactor components can operate satisfactorily up to (a) 10 yr for the stainless-steel first wall, (b) 10 yr for the superconductor composite after which magnet warmup becomes necessary, and (c) 30 yr for the Mylar insulation. Nuclear heat generation rates in the blanket/shield and magnet are well within the practical limits for heat removal

  6. Thermal performance of various multilayer insulation systems below 80K

    International Nuclear Information System (INIS)

    Boroski, W.N.; Nicol, T.H.; Schoo, C.J.

    1992-04-01

    The SSC collider dipole cryostat consists of a vacuum shell operating at room temperature, two thermal shields operating near 80K and 20K respectively, and the superconducting magnet assembly operating near 4K. The cryostat design incorporates multilayer insulation (MLI) blankets to limit radiant heat transfer into the 80K and 20K thermal shields. Also, an MLI blanket is used to impede heat transfer through residual gas conduction into the 4K superconducting magnet assembly. A measurement facility at Fermilab has been used to experimentally optimize the thermal insulation system for the dipole cryostat. Previous thermal measurements have been used to define the 80K MLI system configuration and verify system performance. With the 80K MLI system defined, the current effort has focused on experimentally defining the optimum insulation scheme for the 20K thermal shield. The SSC design specification requires that radiant heat transfer be limited to 0.093 W/m 2 at an insulating vacuum of 10 -6 torr

  7. 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

  8. Helium Loop for the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

    2006-01-01

    In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

  9. Advantages of super-light structures

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl

    2009-01-01

    Super-light structures with pearl-chain reinforcement is a new revolutionary technology that opens possibilities of building load-bearing structures much cheaper and with several other advantages compared to traditional constructions of concrete and steel. Some benefits are: 1 Half price or less. 2...... Architectural expressions never seen before. 3 Material savings of more than 50 %. 4 Energy and CO2 savings of more than 50 %. 5 No Scaffolding and cheaper moulds. 6 Improved durability. 7 Extreme thermal insulation. 8 Better indoor-climate. 9 User friendly operation and maintenance. 10 Increased safety...

  10. 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

  11. 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

  12. 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

  13. The conversion of a room temperature NaK loop to a high temperature MHD facility for Li/V blanket testing

    International Nuclear Information System (INIS)

    Reed, C.B.; Haglund, R.C.; Miller, M.E.; Nasiatka, J.R.; Kirillov, I.R.; Ogorodnikov, A.P.; Preslitski, G.V.; Goloubovitch, G.P.; Xu, Zeng Yu

    1996-01-01

    The Vanadium/Lithium system has been the recent focus of ANL's Blanket Technology Pro-ram, and for the last several years, ANL's Liquid Metal Blanket activities have been carried out in direct support of the ITER (International Thermonuclear Experimental Reactor) breeding blanket task area. A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the Near the development of insulator coatings. Design calculations, Hua and Gohar, show that an electrically insulating layer is necessary to maintain an acceptably low magneto-hydrodynamic (MHD) pressure drop in the current ITER design. Consequently, the decision was made to convert Argonne's Liquid Metal EXperiment (ALEX) from a 200 degrees C NaK facility to a 350 degrees C lithium facility. The upgraded facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups at Hartmann numbers (M) and interaction parameters (N) in the range of 10 3 to 10 5 in lithium at 350 degrees C. Following completion of the upgrade work, a short performance test was conducted, followed by two longer multiple-hour, MHD tests, all at 230 degrees C. The modified ALEX facility performed up to expectations in the testing. MHD pressure drop and test section voltage distributions were collected at Hartmann numbers of 1000

  14. 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

  15. Economically optimal thermal insulation

    Energy Technology Data Exchange (ETDEWEB)

    Berber, J.

    1978-10-01

    Exemplary calculations to show that exact adherence to the demands of the thermal insulation ordinance does not lead to an optimal solution with regard to economics. This is independent of the mode of financing. Optimal thermal insulation exceeds the values given in the thermal insulation ordinance.

  16. 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.)

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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.)

  2. 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

  3. The Structural Heat Intercept-Insulation-Vibration Evaluation Rig (SHIVER)

    Science.gov (United States)

    Johnson, W. L.; Zoeckler, J. G.; Best-Ameen, L. M.

    2015-01-01

    NASA is currently investigating methods to reduce the boil-off rate on large cryogenic upper stages. Two such methods to reduce the total heat load on existing upper stages are vapor cooling of the cryogenic tank support structure and integration of thick multilayer insulation systems to the upper stage of a launch vehicle. Previous efforts have flown a 2-layer MLI blanket and shown an improved thermal performance, and other efforts have ground-tested blankets up to 70 layers thick on tanks with diameters between 2 3 meters. However, thick multilayer insulation installation and testing in both thermal and structural modes has not been completed on a large scale tank. Similarly, multiple vapor cooled shields are common place on science payload helium dewars; however, minimal effort has gone into intercepting heat on large structural surfaces associated with rocket stages. A majority of the vapor cooling effort focuses on metallic cylinders called skirts, which are the most common structural components for launch vehicles. In order to provide test data for comparison with analytical models, a representative test tank is currently being designed to include skirt structural systems with integral vapor cooling. The tank is 4 m in diameter and 6.8 m tall to contain 5000 kg of liquid hydrogen. A multilayer insulation system will be designed to insulate the tank and structure while being installed in a representative manner that can be extended to tanks up to 10 meters in diameter. In order to prove that the insulation system and vapor cooling attachment methods are structurally sound, acoustic testing will also be performed on the system. The test tank with insulation and vapor cooled shield installed will be tested thermally in the B2 test facility at NASAs Plumbrook Station both before and after being vibration tested at Plumbrooks Space Power Facility.

  4. Thermal insulating panel

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, J.T.

    1985-09-11

    A panel of thermal insulation material has at least one main portion which comprises a dry particulate insulation material compressed within a porous envelope so that it is rigid or substantially rigid and at least one auxiliary portion which is secured to and extends along at least one of the edges of the main portions. The auxiliary portions comprise a substantially uncompressed dry particulate insulation material contained within an envelope. The insulation material of the auxiliary portion may be the same as or may be different from the insulation material of the main portion. The envelope of the auxiliary portion may be made of a porous or a non-porous material. (author).

  5. Nonlinear Super Integrable Couplings of Super Classical-Boussinesq Hierarchy

    Directory of Open Access Journals (Sweden)

    Xiuzhi Xing

    2014-01-01

    Full Text Available Nonlinear integrable couplings of super classical-Boussinesq hierarchy based upon an enlarged matrix Lie super algebra were constructed. Then, its super Hamiltonian structures were established by using super trace identity. As its reduction, nonlinear integrable couplings of the classical integrable hierarchy were obtained.

  6. 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

  7. 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

  8. 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

  9. 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.)

  10. 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

  11. Airborne and impact sound transmission in super-light structures

    DEFF Research Database (Denmark)

    Christensen, Jacob Ellehauge; Hertz, Kristian Dahl; Brunskog, Jonas

    2011-01-01

    -aggregate concrete. A super-light deck element is developed. It is intended to be lighter than traditional deck structures without compromising the acoustic performance. It is primarily the airborne sound insulation, which is of interest as the requirements for the impact sound insulation to a higher degree can...... be fulfilled by external means such as floorings. The acoustical performance of the slab element is enhanced by several factors. Load carrying internal arches stiffens the element. This causes a decrease in the modal density, which is further improved by the element being lighter. These parameters also...

  12. Standard Practice for Evaluating Thermal Insulation Materials for Use in Solar Collectors

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1994-01-01

    1.1 This practice sets forth a testing methodology for evaluating the properties of thermal insulation materials to be used in solar collectors with concentration ratios of less than 10. Tests are given herein to evaluate the pH, surface burning characteristics, moisture adsorption, water absorption, thermal resistance, linear shrinkage (or expansion), hot surface performance, and accelerated aging. This practice provides a test for surface burning characteristics but does not provide a methodology for determining combustibility performance of thermal insulation materials. 1.2 The tests shall apply to blanket, rigid board, loose-fill, and foam thermal insulation materials used in solar collectors. Other thermal insulation materials shall be tested in accordance with the provisions set forth herein and should not be excluded from consideration. 1.3 The assumption is made that elevated temperature, moisture, and applied stresses are the primary factors contributing to the degradation of thermal insulation mat...

  13. Super periodic potential

    Science.gov (United States)

    Hasan, Mohammd; Mandal, Bhabani Prasad

    2018-04-01

    In this paper we introduce the concept of super periodic potential (SPP) of arbitrary order n, n ∈I+, in one dimension. General theory of wave propagation through SPP of order n is presented and the reflection and transmission coefficients are derived in their closed analytical form by transfer matrix formulation. We present scattering features of super periodic rectangular potential and super periodic delta potential as special cases of SPP. It is found that the symmetric self-similarity is the special case of super periodicity. Thus by identifying a symmetric fractal potential as special cases of SPP, one can obtain the tunnelling amplitude for a particle from such fractal potential. By using the formalism of SPP we obtain the close form expression of tunnelling amplitude of a particle for general Cantor and Smith-Volterra-Cantor potentials.

  14. NETL Super Computer

    Data.gov (United States)

    Federal Laboratory Consortium — The NETL Super Computer was designed for performing engineering calculations that apply to fossil energy research. It is one of the world’s larger supercomputers,...

  15. Insulating Coating Development for Vanadium Alloys. Phase I Technical Report

    International Nuclear Information System (INIS)

    Gunda, N.; Sastri, S.; Jayaraman, M.; Karandikar, P.

    2000-01-01

    Self-cooled liquid-lithium/vanadium blanket offers many advantages for fusion power systems. Liquid metals moving through a magnetic field are subjected to magnetohydrodynamic (MHD) effects that can increase the pressure drop and affect the flow profiles and heat transfer. Insulating coatings are required to eliminate this effect. Based on the thermodynamic stability data five different coatings were selected PVD and CVD processes were developed to deposit these coatings. All coatings have resistivities much higher than the minimum required. Liquid lithium testing at Argonne National Laboratory indicates that one of the coatings showed only partial spalling. Thus, further refinement of this coating has significant potential to satisfy the requirements for Li/V blanket technology

  16. 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

  17. ITER blanket module connectors. Design, analysis and testing for procurement arrangement

    International Nuclear Information System (INIS)

    Khomiakov, S.; Poddubnyi, I.; Kolganov, V.; Zhmakin, A.; Parshutin, E.; Danilov, I.; Strebkov, Yu.; Skladnov, K.; Vlasov, D.; Cheburova, A.; Romannikov, A.; Raffray, R.; Egorov, K.; Chappuis, Ph.; Sadakov, S.; Calcagno, B.; Roccella, R.

    2016-01-01

    Highlights: • Procurement Arrangement on Blanket Module Connections (BMC) was signed by ITER Organization and Russian Federation Domestic Agency in late 2014. • “N.A. Dollezhal Research and Development Institute of Power Engineering” (NIKIET) was selected as a general supplier of BMC. • NIKIET plays a key role in design development, analytical and experimental justification and manufacturing of BMC. • NIKIET shall fabricate, test and deliver to ITER 2109 flexible supports, 2561 pads, 1053 electrical straps and 1053 pedestals. - Abstract: A standard ITER Blanket module (BM) is attached to the Vacuum Vessel (VV) with a special system of Blanket Module Connections (BMCs) comprising flexible supports, insulating key pads and electrical straps. BMCs fix the modules relative to the VV and manage the current flow. They accommodate transient, cyclic, thermal and electro-magnetic (EM) loads in a vacuum environment and under neutron radiation. Dynamic, thermal-structural and strength analyses have been performed in support of the BMC design and the results have been experimentally confirmed. The components with uncertain behavior including partially and non-preloaded threads, insulation coating, and electrical contacts were designed by experiments. The effort to develop a reliable and robust design of the BMCs in time for the signature of the Procurement Arrangement on BMCs between ITER Organization and Russian Federation in late 2014 spanned several years. It includes design and analysis as well as experimental activities by the ITER Organization and by JSC “NIKIET” (Russia), which, as an affirmed subcontractor will manufacture and supply BMCs to the ITER site. This paper summarizes the overall effort focusing in particular on the more recent PA supporting activities.

  18. ITER blanket module connectors. Design, analysis and testing for procurement arrangement

    Energy Technology Data Exchange (ETDEWEB)

    Khomiakov, S., E-mail: khomias58@mail.ru [Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Str. 2/8, Moscow (Russian Federation); Poddubnyi, I.; Kolganov, V.; Zhmakin, A.; Parshutin, E.; Danilov, I.; Strebkov, Yu.; Skladnov, K.; Vlasov, D.; Cheburova, A. [Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Str. 2/8, Moscow (Russian Federation); Romannikov, A. [Institution “Project Center ITER”, 123098, Academic Kurchatov' s Sq.,1, Moscow (Russian Federation); Raffray, R.; Egorov, K.; Chappuis, Ph.; Sadakov, S.; Calcagno, B.; Roccella, R. [ITER Organization, Route de Vinon sur Verdon, 13067 St. Paul-Lez-Durance (France)

    2016-11-01

    Highlights: • Procurement Arrangement on Blanket Module Connections (BMC) was signed by ITER Organization and Russian Federation Domestic Agency in late 2014. • “N.A. Dollezhal Research and Development Institute of Power Engineering” (NIKIET) was selected as a general supplier of BMC. • NIKIET plays a key role in design development, analytical and experimental justification and manufacturing of BMC. • NIKIET shall fabricate, test and deliver to ITER 2109 flexible supports, 2561 pads, 1053 electrical straps and 1053 pedestals. - Abstract: A standard ITER Blanket module (BM) is attached to the Vacuum Vessel (VV) with a special system of Blanket Module Connections (BMCs) comprising flexible supports, insulating key pads and electrical straps. BMCs fix the modules relative to the VV and manage the current flow. They accommodate transient, cyclic, thermal and electro-magnetic (EM) loads in a vacuum environment and under neutron radiation. Dynamic, thermal-structural and strength analyses have been performed in support of the BMC design and the results have been experimentally confirmed. The components with uncertain behavior including partially and non-preloaded threads, insulation coating, and electrical contacts were designed by experiments. The effort to develop a reliable and robust design of the BMCs in time for the signature of the Procurement Arrangement on BMCs between ITER Organization and Russian Federation in late 2014 spanned several years. It includes design and analysis as well as experimental activities by the ITER Organization and by JSC “NIKIET” (Russia), which, as an affirmed subcontractor will manufacture and supply BMCs to the ITER site. This paper summarizes the overall effort focusing in particular on the more recent PA supporting activities.

  19. 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

  20. 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)

  1. Panels of microporous insulation

    Energy Technology Data Exchange (ETDEWEB)

    McWilliams, J.A.; Morgan, D.E.; Jackson, J.D.J.

    1990-08-07

    Microporous thermal insulation materials have a lattice structure in which the average interstitial dimension is less than the mean free path of the molecules of air or other gas in which the material is arranged. This results in a heat flow which is less than that attributable to the molecular heat diffusion of the gas. According to this invention, a method is provided for manufacturing panels of microporous thermal insulation, in particular such panels in which the insulation material is bonded to a substrate. The method comprises the steps of applying a film of polyvinyl acetate emulsion to a non-porous substrate, and compacting powdery microporous thermal insulation material against the film so as to cause the consolidated insulation material to bond to the substrate and form a panel. The polyvinyl acetate may be applied by brushing or spraying, and is preferably allowed to dry prior to compacting the insulation material. 1 fig.

  2. Wall insulation system

    Energy Technology Data Exchange (ETDEWEB)

    Kostek, P.T.

    1987-08-11

    In a channel specially designed to fasten semi-rigid mineral fibre insulation to masonry walls, it is known to be constructed from 20 gauge galvanized steel or other suitable material. The channel is designed to have pre-punched holes along its length for fastening of the channel to the drywall screw. The unique feature of the channel is the teeth running along its length which are pressed into the surface of the butted together sections of the insulation providing a strong grip between the two adjacent pieces of insulation. Of prime importance to the success of this system is the recent technological advancements of the mineral fibre itself which allow the teeth of the channel to engage the insulation fully and hold without mechanical support, rather than be repelled or pushed back by the inherent nature of the insulation material. After the insulation is secured to the masonry wall by concrete nail fastening systems, the drywall is screwed to the channel.

  3. Polyimide-Foam/Aerogel Composites for Thermal Insulation

    Science.gov (United States)

    Williams, Martha; Fesmire, James; Sass, Jared; Smith, Trent; Weoser. Erol

    2009-01-01

    Composites of specific types of polymer foams and aerogel particles or blankets have been proposed to obtain thermal insulation performance superior to those of the neat polyimide foams. These composites have potential to also provide enhanced properties for vibration dampening or acoustic attenuation. The specific type of polymer foam is denoted "TEEK-H", signifying a series, denoted H, within a family of polyimide foams that were developed at NASA s Langley Research Center and are collectively denoted TEEK (an acronym of the inventors names). The specific types of aerogels include Nanogel aerogel particles from Cabot Corporation in Billerica, MA. and of Spaceloft aerogel blanket from Aspen Aerogels in Northborough, MA. The composites are inherently flame-retardant and exceptionally thermally stable. There are numerous potential uses for these composites, at temperatures from cryogenic to high temperatures, in diverse applications that include aerospace vehicles, aircraft, ocean vessels, buildings, and industrial process equipment. Some low-temperature applications, for example, include cryogenic storage and transfer or the transport of foods, medicines, and chemicals. Because of thermal cycling, aging, and weathering most polymer foams do not perform well at cryogenic temperatures and will undergo further cracking over time. The TEEK polyimides are among the few exceptions to this pattern, and the proposed composites are intended to have all the desirable properties of TEEK-H foams, plus improved thermal performance along with enhanced vibration or acoustic-attenuation performance. A composite panel as proposed would be fabricated by adding an appropriate amount of TEEK friable balloons into a mold to form a bottom layer. A piece of flexible aerogel blanket material, cut to the desired size and shape, would then be placed on the bottom TEEK layer and sandwiched between another top layer of polyimide friable balloons so that the aerogel blanket would become

  4. Advances in Thermal Insulation. Vacuum Insulation Panels and Thermal Efficiency to Reduce Energy Usage in Buildings

    Energy Technology Data Exchange (ETDEWEB)

    Thorsell, Thomas

    2012-07-01

    We are coming to realize that there is an urgent need to reduce energy usage in buildings and it has to be done in a sustainable way. This thesis focuses on the performance of the building envelope; more precisely thermal performance of walls and super insulation material in the form of vacuum insulation. However, the building envelope is just one part of the whole building system, and super insulators have one major flaw: they are easily adversely affected by other problems in the built environment. Vacuum Insulation Panels are one fresh addition to the arsenal of insulation materials available to the building industry. They are composite material with a core and an enclosure which, as a composite, can reach thermal conductivities as low as 0.004 W/(mK). However, the exceptional performance relies on the barrier material preventing gas permeation, maintaining a near vacuum into the core and a minimized thermal bridge effect from the wrapping of barrier material round the edge of a panel. A serpentine edge is proposed to decrease the heat loss at the edge. Modeling and testing shows a reduction of 60 % if a reasonable serpentine edge is used. A diffusion model of permeation through multilayered barrier films with metallization coatings was developed to predict ultimate service life. The model combines numerical calculations with analytical field theory allowing for more precise determination than current models. The results using the proposed model indicate that it is possible to manufacture panels with lifetimes exceeding 50 years with existing manufacturing. Switching from the component scale to the building scale; an approach of integrated testing and modeling is proposed. Four wall types have been tested in a large range of environments with the aim to assess the hydrothermal nature and significance of thermal bridges and air leakages. The test procedure was also examined as a means for a more representative performance indicator than R-value (in USA). The

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. 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.

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. Translucent insulating building envelope

    DEFF Research Database (Denmark)

    Rahbek, Jens Eg

    1997-01-01

    A new type of translucent insulating material has been tested. This material is made of Celulose-Acetat and have a honey-comb structure. The material has a high solar transmittance and is highly insulating. The material is relatively cheap to produce. Danish Title: Translucent isolerende klimaskærm....

  17. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)]. E-mail: wongc@fusion.gat.com; Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Sawan, M. [University of Wisconsin, Madison, WI (United States); Dagher, M. [University of California, Los Angeles, CA (United States); Smolentsev, S. [University of California, Los Angeles, CA (United States); Merrill, B. [INEEL, Idaho Falls, ID (United States); Youssef, M. [University of California, Los Angeles, CA (United States); Reyes, S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Sze, D.K. [University of California, San Diego, CA (United States); Morley, N.B. [University of California, Los Angeles, CA (United States); Sharafat, S. [University of California, Los Angeles, CA (United States); Calderoni, P. [University of California, Los Angeles, CA (United States); Sviatoslavsky, G. [University of Wisconsin, Madison, WI (United States); Kurtz, R. [Pacific Northwest Laboratory, Richland, WA (United States); Fogarty, P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Zinkle, S. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Abdou, M. [University of California, Los Angeles, CA (United States)

    2006-02-15

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC{sub f}/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 deg. C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R and D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

  18. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-07-05

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperture of 700C. We have identified critical issues for the concept, some of which inlude the first wall design, the assessment of MHD effectrs with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time, we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

  19. Sound Insulation between Dwellings

    DEFF Research Database (Denmark)

    Rasmussen, Birgit

    2011-01-01

    Regulatory sound insulation requirements for dwellings exist in more than 30 countries in Europe. In some countries, requirements have existed since the 1950s. Findings from comparative studies show that sound insulation descriptors and requirements represent a high degree of diversity...... and initiate – where needed – improvement of sound insulation of new and existing dwellings in Europe to the benefit of the inhabitants and the society. A European COST Action TU0901 "Integrating and Harmonizing Sound Insulation Aspects in Sustainable Urban Housing Constructions", has been established and runs...... 2009-2013. The main objectives of TU0901 are to prepare proposals for harmonized sound insulation descriptors and for a European sound classification scheme with a number of quality classes for dwellings. Findings from the studies provide input for the discussions in COST TU0901. Data collected from 24...

  20. Performing the Super Instrument

    DEFF Research Database (Denmark)

    Kallionpaa, Maria

    2016-01-01

    can empower performers by producing super instrument works that allow the concert instrument to become an ensemble controlled by a single player. The existing instrumental skills of the performer can be multiplied and the qualities of regular acoustic instruments extended or modified. Such a situation......The genre of contemporary classical music has seen significant innovation and research related to new super, hyper, and hybrid instruments, which opens up a vast palette of expressive potential. An increasing number of composers, performers, instrument designers, engineers, and computer programmers...... have become interested in different ways of “supersizing” acoustic instruments in order to open up previously-unheard instrumental sounds. Super instruments vary a great deal but each has a transformative effect on the identity and performance practice of the performing musician. Furthermore, composers...

  1. 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.)

  2. 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

  3. Raspberry Pi super cluster

    CERN Document Server

    Dennis, Andrew K

    2013-01-01

    This book follows a step-by-step, tutorial-based approach which will teach you how to develop your own super cluster using Raspberry Pi computers quickly and efficiently.Raspberry Pi Super Cluster is an introductory guide for those interested in experimenting with parallel computing at home. Aimed at Raspberry Pi enthusiasts, this book is a primer for getting your first cluster up and running.Basic knowledge of C or Java would be helpful but no prior knowledge of parallel computing is necessary.

  4. Highly Insulating and Light Transmitting Aerogel Glazing for Super Insulating Windows (HILIT+)

    DEFF Research Database (Denmark)

    Jensen, Karsten Ingerslev; Kristiansen, Finn Harken; Schultz, Jørgen Munthe

    2005-01-01

    to 1000 m²/g), the material is proposed to serve as substrate for catalytic materials. • The special pore structure of aerogel could be used for gas filters in the 20 to 100 nm region. • The sound velocity within aerogel is in the range of 100 to 300 m/s, which should be one of the lowest for an inorganic......-free nano-structured aerogel materials through a reasonably fast and reproducible process. The applicative part of this project aimed at elaborating, studying and optimising “state-of-the-art” (0.5 W/m2 K) aerogel glazings for windows. An important issue was the risk of outside condensation and rime and its....... No other known glazing exhibits such an excellent combination of solar transmittance and heat loss coefficient. The annual energy savings compared to triple low energy glazing is in the range of 10 – 20% depending on type of building. Beside the application in glazing production the HILIT+ aerogel material...

  5. The super-resolution debate

    Science.gov (United States)

    Won, Rachel

    2018-05-01

    In the quest for nanoscopy with super-resolution, consensus from the imaging community is that super-resolution is not always needed and that scientists should choose an imaging technique based on their specific application.

  6. 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

  7. 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

  8. Frames in super Hilbert modules

    Directory of Open Access Journals (Sweden)

    Mehdi Rashidi-Kouchi

    2018-01-01

    Full Text Available In this paper, we define super Hilbert module and investigate frames in this space. Super Hilbert modules are  generalization of super Hilbert spaces in Hilbert C*-module setting. Also, we define frames in a super Hilbert module and characterize them by using of the concept of g-frames in a Hilbert C*-module. Finally, disjoint frames in Hilbert C*-modules are introduced and investigated.

  9. 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)

  10. 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.)

  11. Heat insulation support device

    International Nuclear Information System (INIS)

    Takahashi, Hiroyuki; Koda, Tomokazu; Motojima, Osamu; Yamamoto, Junya.

    1994-01-01

    The device of the present invention comprises a plurality of heat insulation legs disposed in a circumferential direction. Each of the heat insulative support legs has a hollow shape, and comprises an outer column and an inner column as support structures having a heat insulative property (heat insulative structure), and a thermal anchor which absorbs compulsory displacement by a thin flat plate (displacement absorber). The outer column, the thermal anchor and the inner column are connected by a support so as to offset the positional change of objects to be supported due to shrinkage when they are shrunk. In addition, the portion between the superconductive coils as the objects to be supported and the inner column is connected by the support. The superconductive thermonuclear device is entirely contained in a heat insulative vacuum vessel, and the heat insulative support legs are disposed on a lower lid of the heat insulative vacuum vessel. With such a constitution, they are strengthened against lateral load and buckling, thereby enabling to reduce the amount of heat intrusion while keeping the compulsory displacement easy to be absorbed. (I.N.)

  12. Handbook of Super 8 Production.

    Science.gov (United States)

    Telzer, Ronnie, Ed.

    This handbook is designed for anyone interested in producing super 8 films at any level of complexity and cost. Separate chapters present detailed discussions of the following topics: super 8 production systems and super 8 shooting and editing systems; budgeting; cinematography and sound recording; preparing to edit; editing; mixing sound tracks;…

  13. Super-resolution

    DEFF Research Database (Denmark)

    Nasrollahi, Kamal; Moeslund, Thomas B.

    2014-01-01

    Super-resolution, the process of obtaining one or more high-resolution images from one or more low-resolution observations, has been a very attractive research topic over the last two decades. It has found practical applications in many real world problems in different fields, from satellite...

  14. Superconducting Super Collider project

    International Nuclear Information System (INIS)

    Perl, M.L.

    1986-04-01

    The scientific need for the Superconducting Super Collider (SSC) is outlined, along with the history of the development of the SSC concept. A brief technical description is given of each of the main points of the SSC conceptual design. The construction cost and construction schedule are discussed, followed by issues associated with the realization of the SSC. 8 refs., 3 figs., 3 tabs

  15. Super Refractory Status Epilepticus

    African Journals Online (AJOL)

    user

    et al did retrospective cohort study from 1 January st. 1994 to 31 March 1998 at Presbyterian Medical. Centre in Columbia, to determine the frequency, risk factors and impact on the outcome of RSE. They found out that 69% of seizures recurred after. Key Words: Super refractory status epilepticus, Zambia. Medical Journal of ...

  16. Optimal Super Dielectric Material

    Science.gov (United States)

    2015-09-01

    plate capacitor will reduce the net field to an unprecedented extent. This family of materials can form materials with dielectric values orders of... Capacitor -Increase Area (A)............8 b. Multi-layer Ceramic Capacitor -Decrease Thickness (d) .......10 c. Super Dielectric Material-Increase...circuit modeling, from [44], and B) SDM capacitor charge and discharge ...................................................22 Figure 15. Dielectric

  17. SuperHILAC

    International Nuclear Information System (INIS)

    Nemetz, R.; Selph, F.; Barnes, A.C.

    1976-01-01

    A brief discussion is given of improvements, operations, and research programs at the SuperHILAC. Improvements were made in beam injection, ion sources, and computer control systems. The research efficiency ranged between 70 and 90 percent during most of the year

  18. Vacuum foil insulation system

    International Nuclear Information System (INIS)

    Hanson, J.P.; Sabolcik, R.E.; Svedberg, R.C.

    1976-01-01

    In a multifoil thermal insulation package having a plurality of concentric cylindrical cups, means are provided for reducing heat loss from the penetration region which extends through the cups. At least one cup includes an integral skirt extending from one end of the cup to intersection with the penetration means. Assembly of the insulation package with the skirted cup is facilitated by splitting the cup to allow it to be opened up and fitted around the other cups during assembly. The insulation is for an implantable nuclear powered artificial heart

  19. The effects of imperfect insulator coatings on MHD and heat transfer in rectangular duct

    International Nuclear Information System (INIS)

    Ying, A.Y.; Gaizer, A.A.

    1994-01-01

    In self cooled liquid metal blankets, the use of an insulator coating to reduce the flow of the eddy current to the structure leads to a significant reduction in MHD pressure drop. Furthermore, this insulating layer alters the velocity structure by reducing the potential difference between the side wall and boundary layer. The questions which arise are: (1) How the imperfections in the insulator coating affect the velocity profiles and their consequent impacts on heat transfer performance?; and, (2) How much crack can lead to an unacceptable MHD pressure drop? The dynamics of the crack healing in an insulator coating duct is one of the important subjects requiring study. The purpose of this work is to present numerical simulations of fully developed MHD flow and developing heat transfer characteristics in imperfectly insulated ducts, and to quantify the influences of crack locations, sizes and resistivities on 2-D MHD pressure drops. Comparisons of finite element solutions of pressure drops in partially insulated ducts with analytical solutions obtained from a circuit analogy show excellent agreement. In addition, the remarkable side layer velocity profile observed in a laminar MHD flow of a conducting duct gradually diminishes as the resistance of the insulating layer increases. The average side wall Nusselt number drops by a factor of 2 as the duct becomes fully insulated

  20. 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

  1. 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.)

  2. 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

  3. 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.

  4. 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

  5. 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

  6. Super-energy wrap-up model for renovation of standard wooden houses in Greenland?

    DEFF Research Database (Denmark)

    Bjarløv, Søren Peter; Vladyková, Petra

    2012-01-01

    This paper is based on the developed theoretical practice of super-insulation solution for renovation of exiting standard wooden houses in Greenland built from 1950s until 2006, more than half the building stock in Arctic Greenland. From various perspectives, the wrap-up system is evaluated as a ...

  7. Cooper Pairs in Insulators?

    International Nuclear Information System (INIS)

    Valles, James

    2008-01-01

    Nearly 50 years elapsed between the discovery of superconductivity and the emergence of the microscopic theory describing this zero resistance state. The explanation required a novel phase of matter in which conduction electrons joined in weakly bound pairs and condensed with other pairs into a single quantum state. Surprisingly, this Cooper pair formation has also been invoked to account for recently uncovered high-resistance or insulating phases of matter. To address this possibility, we have used nanotechnology to create an insulating system that we can probe directly for Cooper pairs. I will present the evidence that Cooper pairs exist and dominate the electrical transport in these insulators and I will discuss how these findings provide new insight into superconductor to insulator quantum phase transitions.

  8. Gas insulated substations

    CERN Document Server

    2014-01-01

    This book provides an overview on the particular development steps of gas insulated high-voltage switchgear, and is based on the information given with the editor's tutorial. The theory is kept low only as much as it is needed to understand gas insulated technology, with the main focus of the book being on delivering practical application knowledge. It discusses some introductory and advanced aspects in the meaning of applications. The start of the book presents the theory of Gas Insulated Technology, and outlines reliability, design, safety, grounding and bonding, and factors for choosing GIS. The third chapter presents the technology, covering the following in detail: manufacturing, specification, instrument transformers, Gas Insulated Bus, and the assembly process. Next, the book goes into control and monitoring, which covers local control cabinet, bay controller, control schemes, and digital communication. Testing is explained in the middle of the book before installation and energization. Importantly, ...

  9. Deformations of super Riemann surfaces

    International Nuclear Information System (INIS)

    Ninnemann, H.

    1992-01-01

    Two different approaches to (Konstant-Leites-) super Riemann surfaces are investigated. In the local approach, i.e. glueing open superdomains by superconformal transition functions, deformations of the superconformal structure are discussed. On the other hand, the representation of compact super Riemann surfaces of genus greater than one as a fundamental domain in the Poincare upper half-plane provides a simple description of super Laplace operators acting on automorphic p-forms. Considering purely odd deformations of super Riemann surfaces, the number of linear independent holomorphic sections of arbitrary holomorphic line bundles will be shown to be independent of the odd moduli, leading to a simple proof of the Riemann-Roch theorem for compact super Riemann surfaces. As a further consequence, the explicit connections between determinants of super Laplacians and Selberg's super zeta functions can be determined, allowing to calculate at least the 2-loop contribution to the fermionic string partition function. (orig.)

  10. Deformations of super Riemann surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Ninnemann, H [Hamburg Univ. (Germany). 2. Inst. fuer Theoretische Physik

    1992-11-01

    Two different approaches to (Konstant-Leites-) super Riemann surfaces are investigated. In the local approach, i.e. glueing open superdomains by superconformal transition functions, deformations of the superconformal structure are discussed. On the other hand, the representation of compact super Riemann surfaces of genus greater than one as a fundamental domain in the Poincare upper half-plane provides a simple description of super Laplace operators acting on automorphic p-forms. Considering purely odd deformations of super Riemann surfaces, the number of linear independent holomorphic sections of arbitrary holomorphic line bundles will be shown to be independent of the odd moduli, leading to a simple proof of the Riemann-Roch theorem for compact super Riemann surfaces. As a further consequence, the explicit connections between determinants of super Laplacians and Selberg's super zeta functions can be determined, allowing to calculate at least the 2-loop contribution to the fermionic string partition function. (orig.).

  11. Wrapped Multilayer Insulation

    Science.gov (United States)

    Dye, Scott A.

    2015-01-01

    New NASA vehicles, such as Earth Departure Stage (EDS), Orion, landers, and orbiting fuel depots, need improved cryogenic propellant transfer and storage for long-duration missions. Current cryogen feed line multilayer insulation (MLI) performance is 10 times worse per area than tank MLI insulation. During each launch, cryogenic piping loses approximately 150,000 gallons (equivalent to $300,000) in boil-off during transfer, chill down, and ground hold. Quest Product Development Corp., teaming with Ball Aerospace, developed an innovative advanced insulation system, Wrapped MLI (wMLI), to provide improved thermal insulation for cryogenic feed lines. wMLI is high-performance multilayer insulation designed for cryogenic piping. It uses Quest's innovative discrete-spacer technology to control layer spacing/ density and reduce heat leak. The Phase I project successfully designed, built, and tested a wMLI prototype with a measured heat leak 3.6X lower than spiral-wrapped conventional MLI widely used for piping insulation. A wMLI prototype had a heat leak of 7.3 W/m2, or 27 percent of the heat leak of conventional MLI (26.7 W/m2). The Phase II project is further developing wMLI technology with custom, molded polymer spacers and advancing the product toward commercialization via a rigorous testing program, including developing advanced vacuuminsulated pipe for ground support equipment.

  12. 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

  13. 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

  14. 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)

  15. Accelerator driven heavy water blanket on circulating fuel

    International Nuclear Information System (INIS)

    Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

    1997-01-01

    A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

  16. 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)

  17. An assessment of the base blanket for ITER

    International Nuclear Information System (INIS)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-01-01

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored

  18. Objectives and status of EUROfusion DEMO blanket studies

    Energy Technology Data Exchange (ETDEWEB)

    Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

    2016-11-01

    Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

  19. 18 CFR 284.303 - OCS blanket certificates.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false OCS blanket certificates. 284.303 Section 284.303 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... Pipelines on Behalf of Others § 284.303 OCS blanket certificates. Every OCS pipeline [as that term is...

  20. 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)

  1. Survey of thermal insulation systems

    International Nuclear Information System (INIS)

    Kinoshita, Izumi

    1983-01-01

    Better thermal insulations have been developed to meet the growing demands of industry, and studies on thermal insulation at both high temperature and low temperature have been widely performed. The purpose of this survey is to summarize data on the performances and characteristics of thermal insulation materials and thermal insulation structures (for instance, gas cooled reactors, space vehicles and LNG storage tanks), and to discuss ravious problems regarding the design of thermal insulation structures of pool-type LMFBRs. (author)

  2. Super Talbot effect in indefinite metamaterial.

    Science.gov (United States)

    Zhao, Wangshi; Huang, Xiaoyue; Lu, Zhaolin

    2011-08-01

    The Talbot effect (or the self-imaging effect) can be observed for a periodic object with a pitch larger than the diffraction limit of an imaging system, where the paraxial approximation is applied. In this paper, we show that the super Talbot effect can be achieved in an indefinite metamaterial even when the period is much smaller than the diffraction limit in both two-dimensional and three-dimensional numerical simulations, where the paraxial approximation is not applied. This is attributed to the evanescent waves, which carry the information about subwavelength features of the object, can be converted into propagating waves and then conveyed to far field by the metamaterial, where the permittivity in the propagation direction is negative while the transverse ones are positive. The indefinite metamaterial can be approximated by a system of thin, alternating multilayer metal and insulator (MMI) stack. As long as the loss of the metamaterial is small enough, deep subwavelength image size can be obtained in the super Talbot effect.

  3. Coherence and correlations in a Mott insulator

    International Nuclear Information System (INIS)

    Gerbier, F.; Widera, A.; Foelling, S.; Mandel, O.; Gericke, T.; Bloch, I.

    2005-01-01

    The observation of the super fluid to Mott insulator transition has triggered an intense interest in studying ultracold quantum gases in optical lattices. Such a transition is commonly associated with the disappearance of the interference pattern observed when releasing a coherent (i.e. Bose condensed) ensemble from the lattice. In this talk, I will show that even in the insulating phase, the visibility of this interference pattern remains finite. Our results show that although long-range order is absent, short-range coherence still persists in a rather broad range, and that this can be identified as a characteristic feature of the system for large, but finite lattice depths. For even deeper lattices, the visibility is close to zero, and the interference pattern unobservable. I will explain that information is still present in such featureless images, and can be extracted by studying the density-density correlation function of the expanded cloud, as proposed theoretically. A sharp diffraction-like pattern observed in the correlations reveals the underlying lattice structure, and can be understood by generalizing the well-known Hanbury-Brown and Twiss effect to many bosonic sources '' emitting '' from each lattice site. This new detection method allows in principle the detection of spin ordering in a multi-component Mott insulator. As a first step in this direction, we have recently observed spin dynamics in a Mott insulator, where a spin-dependent collisional coupling induces strongly under damped Rabi oscillations between two-particle states with the same total magnetization. I will briefly report on these results. (author)

  4. EUROv Super Beam Studies

    International Nuclear Information System (INIS)

    Dracos, Marcos

    2011-01-01

    Neutrino Super Beams use conventional techniques to significantly increase the neutrino beam intensity compared to the present neutrino facilities. An essential part of these facilities is an intense proton driver producing a beam power higher than a MW. The protons hit a target able to accept the high proton beam intensity. The produced charged particles are focused by a system of magnetic horns towards the experiment detectors. The main challenge of these projects is to deal with the high beam intensity for many years. New high power neutrino facilities could be build at CERN profiting from an eventual construction of a high power proton driver. The European FP7 Design Study EUROv, among other neutrino beams, studies this Super Beam possibility. This paper will give the latest developments in this direction.

  5. SuperSegger

    DEFF Research Database (Denmark)

    Stylianidou, Stella; Brennan, Connor; Nissen, Silas B

    2016-01-01

    -colonies with many cells, facilitating the analysis of cell-cycle dynamics in bacteria as well as cell-contact mediated phenomena. This package has a range of built-in capabilities for characterizing bacterial cells, including the identification of cell division events, mother, daughter, and neighboring cells......Many quantitative cell biology questions require fast yet reliable automated image segmentation to identify and link cells from frame-to-frame, and characterize the cell morphology and fluorescence. We present SuperSegger, an automated MATLAB-based image processing package well......-suited to quantitative analysis of high-throughput live-cell fluorescence microscopy of bacterial cells. SuperSegger incorporates machine-learning algorithms to optimize cellular boundaries and automated error resolution to reliably link cells from frame-to-frame. Unlike existing packages, it can reliably segment micro...

  6. Super-Lagrangians

    International Nuclear Information System (INIS)

    Beyl, L.M.

    1979-01-01

    It is shown that the Einstein, Weyl, supergravity and superconformal theories are special cases of gauge transformations in SU(4vertical-barN). This group is shown to contain SU(2,2) x SU(N) x U(1) for its commuting or Bose part, and to contain 8N supersymmetry generators for its anticommuting or Fermi part. Using the electromagnetic Lagrangian as a model, a super-Lagrangian is constructed for vector potentials. Invariance is automatic in free space, but, in the presence of matter, restrictions on the supersymmetry transformations are necessary. The Weyl action and the Einstein cosmological field equations are obtained in the appropriate limits. Finally, a super-Lagrangian is constructed from nongeometric principles which includes the Dirac Lagrangian and except for a sum over symmetry indices resembles the electron-electromagnetic Lagrangian

  7. 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

  8. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

  9. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

    In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

  10. 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

  11. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

    1971-01-01

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  12. Self-cooled liquid-metal blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Arheidt, K.; Barleon, L.

    1988-01-01

    A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

  13. Minimal Super Technicolor

    DEFF Research Database (Denmark)

    Antola, M.; Di Chiara, S.; Sannino, F.

    2011-01-01

    We introduce novel extensions of the Standard Model featuring a supersymmetric technicolor sector (supertechnicolor). As the first minimal conformal supertechnicolor model we consider N=4 Super Yang-Mills which breaks to N=1 via the electroweak interactions. This is a well defined, economical......, between unparticle physics and Minimal Walking Technicolor. We consider also other N =1 extensions of the Minimal Walking Technicolor model. The new models allow all the standard model matter fields to acquire a mass....

  14. Development of radiation hard components for ITER blanket remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Makiko, E-mail: saito.makiko@jaea.go.jp; Anzai, Katsunori; Maruyama, Takahito; Noguchi, Yuto; Ueno, Kenichi; Takeda, Nobukazu; Kakudate, Satoshi

    2016-11-01

    Highlights: • Clarify the components that will degrade by gamma ray irradiation. • Perform the irradiation tests to BRHS components. • Optimize the materials to increase the radiation hardness. - Abstract: The ITER blanket remote handling system (BRHS) will be operated in a high radiation environment (250 Gy/h max.) and must stably handle the blanket modules, which weigh 4.5 t and are more than 1.5 m in length, with a high degree of position and posture accuracy. The reliability of the system can be improved by reviewing the failure events of the system caused by high radiation. A failure mode and effects analysis (FMEA) identified failure modes and determined that lubricants, O-rings, and electric insulation cables were the dominant components affecting radiation hardness. Accordingly, we tried to optimize the lubricants and cables of the AC servo motors by using polyphenyl ether (PPE)-based grease and polyether ether ketone (PEEK), respectively. Materials containing radiation protective agents were also selected for the cable sheaths and O-rings to improve radiation hardness. Gamma ray irradiation tests were performed on these components and as a result, a radiation hardness of 8 MGy was achieved for the AC servo motors. On the other hand, to develop the radiation hardness and BRHS compatibility furthermore, the improvement of materials of cable and O ring were performed.

  15. Application of the integrated blanket-coil concept (IBC) to fusion reactors

    International Nuclear Information System (INIS)

    Embrechts, M.J.; Steiner, D.; Mohanti, R.; Duggan, W.

    1987-01-01

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component and several unique applications to fusion reactor embodiments are identified. The proposed concept takes advantage of the fact that lithium is a good electrical conductor in addition to being a unique tritium-breeding material capable of energy recovery and transport at high temperatures. This concept, designated the ''integrated-blanket-coil (IBC) concept'' has the potential for: allowing fusion reactor embodiments which are easier to maintain; making fusion reactors more compact with an intrinsic ultra-high mass power density (net kW/sub E//metric tonne); and enhancing the tritium breeding potential for special coil applications such as ohmic heating and bean identation. By assuming a sandwich construction for the IBC walls (i.e., a layered combination of a thin wall of structural material, insulator and structural materials) the magnetohydrodynamic (MHD)-induced pressure drops and associated pressure stresses are modest and well below design limits. Possible unique applications of the IBC concept have been investigated and include the IBC concept applied to the poloidal field (PF) coils, toroidal field (TF) coils, divertor coils, ohmic heating (OH) coils, and identation coils for bean shaping

  16. Characterising Super-Earths

    Directory of Open Access Journals (Sweden)

    Valencia D.

    2011-02-01

    Full Text Available The era of Super-Earths has formally begun with the detection of transiting low-mass exoplanets CoRoT-7b and GJ 1214b. In the path of characterising super-Earths, the first step is to infer their composition. While the discovery data for CoRoT-7b, in combination with the high atmospheric mass loss rate inferred from the high insolation, suggested that it was a rocky planet, the new proposed mass values have widened the possibilities. The combined mass range 1−10 M⊕ allows for a volatile-rich (and requires it if the mass is less than 4 M⊕ , an Earth-like or a super-Mercury-like composition. In contrast, the radius of GJ 1214b is too large to admit a solid composition, thus it necessarily to have a substantial gas layer. Some evidence suggests that within this gas layer H/He is a small but non-negligible component. These two planets are the first of many transiting low-mass exoplanets expected to be detected and they exemplify the limitations faced when inferring composition, which come from the degenerate character of the problem and the large error bars in the data.

  17. Super-quantum curves from super-eigenvalue models

    Energy Technology Data Exchange (ETDEWEB)

    Ciosmak, Paweł [Faculty of Mathematics, Informatics and Mechanics, University of Warsaw,ul. Banacha 2, 02-097 Warsaw (Poland); Hadasz, Leszek [M. Smoluchowski Institute of Physics, Jagiellonian University,ul. Łojasiewicza 11, 30-348 Kraków (Poland); Manabe, Masahide [Faculty of Physics, University of Warsaw,ul. Pasteura 5, 02-093 Warsaw (Poland); Sułkowski, Piotr [Faculty of Physics, University of Warsaw,ul. Pasteura 5, 02-093 Warsaw (Poland); Walter Burke Institute for Theoretical Physics, California Institute of Technology,1200 E. California Blvd, Pasadena, CA 91125 (United States)

    2016-10-10

    In modern mathematical and theoretical physics various generalizations, in particular supersymmetric or quantum, of Riemann surfaces and complex algebraic curves play a prominent role. We show that such supersymmetric and quantum generalizations can be combined together, and construct supersymmetric quantum curves, or super-quantum curves for short. Our analysis is conducted in the formalism of super-eigenvalue models: we introduce β-deformed version of those models, and derive differential equations for associated α/β-deformed super-matrix integrals. We show that for a given model there exists an infinite number of such differential equations, which we identify as super-quantum curves, and which are in one-to-one correspondence with, and have the structure of, super-Virasoro singular vectors. We discuss potential applications of super-quantum curves and prospects of other generalizations.

  18. Super-quantum curves from super-eigenvalue models

    International Nuclear Information System (INIS)

    Ciosmak, Paweł; Hadasz, Leszek; Manabe, Masahide; Sułkowski, Piotr

    2016-01-01

    In modern mathematical and theoretical physics various generalizations, in particular supersymmetric or quantum, of Riemann surfaces and complex algebraic curves play a prominent role. We show that such supersymmetric and quantum generalizations can be combined together, and construct supersymmetric quantum curves, or super-quantum curves for short. Our analysis is conducted in the formalism of super-eigenvalue models: we introduce β-deformed version of those models, and derive differential equations for associated α/β-deformed super-matrix integrals. We show that for a given model there exists an infinite number of such differential equations, which we identify as super-quantum curves, and which are in one-to-one correspondence with, and have the structure of, super-Virasoro singular vectors. We discuss potential applications of super-quantum curves and prospects of other generalizations.

  19. Super-quantum curves from super-eigenvalue models

    Science.gov (United States)

    Ciosmak, Paweł; Hadasz, Leszek; Manabe, Masahide; Sułkowski, Piotr

    2016-10-01

    In modern mathematical and theoretical physics various generalizations, in particular supersymmetric or quantum, of Riemann surfaces and complex algebraic curves play a prominent role. We show that such supersymmetric and quantum generalizations can be combined together, and construct supersymmetric quantum curves, or super-quantum curves for short. Our analysis is conducted in the formalism of super-eigenvalue models: we introduce β-deformed version of those models, and derive differential equations for associated α/ β-deformed super-matrix integrals. We show that for a given model there exists an infinite number of such differential equations, which we identify as super-quantum curves, and which are in one-to-one correspondence with, and have the structure of, super-Virasoro singular vectors. We discuss potential applications of super-quantum curves and prospects of other generalizations.

  20. 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.

  1. 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

  2. Thermal Performance of Cryogenic Multilayer Insulation at Various Layer Spacings

    Science.gov (United States)

    Johnson, Wesley Louis

    2010-01-01

    Multilayer insulation (MLI) has been shown to be the best performing cryogenic insulation system at high vacuum (less that 10 (exp 3) torr), and is widely used on spaceflight vehicles. Over the past 50 years, many investigations into MLI have yielded a general understanding of the many variables that are associated with MLI. MLI has been shown to be a function of variables such as warm boundary temperature, the number of reflector layers, and the spacer material in between reflectors, the interstitial gas pressure and the interstitial gas. Since the conduction between reflectors increases with the thickness of the spacer material, yet the radiation heat transfer is inversely proportional to the number of layers, it stands to reason that the thermal performance of MLI is a function of the number of layers per thickness, or layer density. Empirical equations that were derived based on some of the early tests showed that the conduction term was proportional to the layer density to a power. This power depended on the material combination and was determined by empirical test data. Many authors have graphically shown such optimal layer density, but none have provided any data at such low densities, or any method of determining this density. Keller, Cunnington, and Glassford showed MLI thermal performance as a function of layer density of high layer densities, but they didn't show a minimal layer density or any data below the supposed optimal layer density. However, it was recently discovered that by manipulating the derived empirical equations and taking a derivative with respect to layer density yields a solution for on optimal layer density. Various manufacturers have begun manufacturing MLI at densities below the optimal density. They began this based on the theory that increasing the distance between layers lowered the conductive heat transfer and they had no limitations on volume. By modifying the circumference of these blankets, the layer density can easily be

  3. 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

  4. 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

  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. Insulators for fusion applications

    International Nuclear Information System (INIS)

    1987-04-01

    Design studies for fusion devices and reactors have become more detailed in recent years and with this has come a better understanding of requirements and operating conditions for insulators in these machines. Ceramic and organic insulators are widely used for many components of fusion devices and reactors namely: radio frequency (RF) energy injection systems (BeO, Al 2 O 3 , Mg Al 2 O 4 , Si 3 N 4 ); electrical insulation for the torus structure (SiC, Al 2 O 3 , MgO, Mg Al 2 O 4 , Si 4 Al 2 O 2 N 6 , Si 3 N 4 , Y 2 O 3 ); lightly-shielded magnetic coils (MgO, MgAl 2 O 4 ); the toroidal field coil (epoxies, polyimides), neutron shield (B 4 C, TiH 2 ); high efficiency electrical generation; as well as the generation of very high temperatures for high efficiency hydrogen production processes (ZrO 2 and Al 2 O 3 - mat, graphite and carbon - felt). Timely development of insulators for fusion applications is clearly necessary. Those materials to be used in fusion machines should show high resistance to radiation damage and maintain their structural integrity. Now the need is urgent for a variety of radiation resistant materials, but much effort in these areas is required for insulators to be considered seriously by the design community. This document contains 14 papers from an IAEA meeting. It was the objective of this meeting to identify existing problems in analysing various situations of applications and requirements of electrical insulators and ceramics in fusion and to recommend strategies and different stages of implementation. This meeting was endorsed by the International Fusion Research Council

  7. 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

  8. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

    In all fusion reactors using the deuterium-tritium fuel cycle, a large fraction approximately 80 percent of the fusion energy will be released as approximately 14 MeV neutrons which must be slowed down in a relatively thick blanket surrounding the plasma, thereby, converting their kinetic energy to high temperature heat which can be continuously removed by a coolant stream and converted in part to electricity in a conventional power turbine. Because of the primary goal of achieving minimum radioactivity, to date Brookhaven blanket concepts have been restricted to the use of some form of solid lithium, with inert gas-cooling and in some design cases, water-cooling of the shell structure. Aluminum and graphite have been identified as very promising structural materials for fusion blankets, and conceptual designs based on these materials have been made. Depending on the thermal loading on the ''first'' wall which surrounds the plasma as well as blanket design, heat transfer problems may be noticeably different in gas-cooled solid blankets. Approaches to solution of heat removal problems as well as explanation of: (a) the after-heat problems in blankets; (b) tritium breeding in solids; and (c) materials selection for radiation shields relative to the minimum activity blanket efforts at Brookhaven are discussed

  9. 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

  10. 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)

  11. Theory of super LIE groups

    International Nuclear Information System (INIS)

    Prakash, M.

    1985-01-01

    The theory of supergravity has attracted increasing attention in the recent years as a unified theory of elementary particle interactions. The superspace formulation of the theory is highly suggestive of an underlying geometrical structure of superspace. It also incorporates the beautifully geometrical general theory of relativity. It leads us to believe that a better understanding of its geometry would result in a better understanding of the theory itself, and furthermore, that the geometry of superspace would also have physical consequences. As a first step towards that goal, we develop here a theory of super Lie groups. These are groups that have the same relation to a super Lie algebra as Lie groups have to a Lie algebra. More precisely, a super Lie group is a super-manifold and a group such that the group operations are super-analytic. The super Lie algebra of a super Lie group is related to the local properties of the group near the identity. This work develops the algebraic and super-analytical tools necessary for our theory, including proofs of a set of existence and uniqueness theorems for a class of super-differential equations

  12. Fabricated super-hydrophobic film with potentiostatic electrolysis method on copper for corrosion protection

    International Nuclear Information System (INIS)

    Wang Peng; Qiu Ri; Zhang Dun; Lin Zhifeng; Hou Baorong

    2010-01-01

    A novel one-step potentiostatic electrolysis method was proposed to fabricate super-hydrophobic film on copper surface. The resulted film was characterized by contact angle tests, Fourier transform infrared spectra (FT-IR), X-ray photoelectron spectroscopy (XPS), Field emission scanning electron microscopy (FE-SEM) and electrochemical measurements. It could be inferred that the super-hydrophobic property resulted from the flower-like structure of copper tetradecanoate film. In the presence of super-hydrophobic film, the anodic and cathodic polarization current densities are reduced for more than five and four orders of magnitude, respectively. The air trapped in the film is the essential contributor of the anticorrosion property of film for its insulation, the copper tetradecanoate film itself acts as a 'frame' to trap air as well as a coating with inhibition effect. The super-hydrophobic film presents excellent inhibition effect to the copper corrosion and stability in water containing Cl - .

  13. JAPAN: Super-Kamiokande

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    Excavation for the Japanese Super- KAMIOKANDE 50,000-ton water Cherenkov imaging detector was completed at the end of June. The goals include a search for nucleon decay up to a lifetime of 10 33-34 years, high-statistics studies of solar and atmospheric neutrinos, and detection of any nearby supernova explosions. The project was approved in 1991, with the official 'groundbreaking' in December of that year about 1,000 m underground in the Kamioka mine in Gifu Prefecture, about 250 km west of Tokyo

  14. The super collider revisited

    International Nuclear Information System (INIS)

    Hussein, M.S.; Pato, M.P.

    1992-01-01

    In this paper, the authors suggest a revised version of the Superconducting Super Collider (SSC) that employs the planned SSC first stage machine as an injector of 0.5 TeV protons into a power laser accelerator. The recently developed Non-linear Amplification of Inverse Bremsstrahlung Acceleration (NAIBA) concept dictates the scenario of the next stage of acceleration. Post Star Wars lasers, available at several laboratories, can be used for the purpose. The 40 TeV CM energy, a target of the SSC, can be obtained with a new machine which can be 20 times smaller than the planned SSC

  15. The super-LHC

    CERN Document Server

    Mangano, Michelangelo L

    2010-01-01

    We review here the prospects of a long-term upgrade programme for the Large Hadron Collider (LHC), CERN laboratory's new proton-proton collider. The super-LHC, which is currently under evaluation and design, is expected to deliver of the order of ten times the statistics of the LHC. In addition to a non-technical summary of the principal physics arguments for the upgrade, I present a pedagogical introduction to the technological challenges on the accelerator and experimental fronts, and a review of the current status of the planning.

  16. Super-heptazethrene

    KAUST Repository

    Zeng, Wangdong

    2016-05-30

    The challenging synthesis of a laterally extended heptazethrene molecule, the super-heptazethrene derivative SHZ-CF3, is reported. This molecule was prepared using a strategy involving a multiple selective intramolecular Friedel–Crafts alkylation followed by oxidative dehydrogenation. Compound SHZ-CF3 exhibits an open-shell singlet diradical ground state with a much larger diradical character compared with the heptazethrene derivatives. An intermediate dibenzo-terrylene SHZ-2H was also obtained during the synthesis. This study provides a new synthetic method to access large-size quinoidal polycyclic hydrocarbons with unique physical properties.

  17. Overview of EU activities on DEMO liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Malang, S.; Reimann, J.; Perujo, A.

    1994-01-01

    The present paper gives an overview of both design and experimental activities within the European Union (EU) concerning the development of liquid metal breeder blankets for DEMO. After several years of studies on breeding blankets, two blanket concepts are presently considered, both using the eutectic Pb-17Li: the dual-coolant concept and the water-cooled concept. The analysis of such concepts has permitted to identify the experimental areas where further data are required. Tritium control and MHD-issues are, at present, the activities on which is devoted the greatest effort within the EU. (authors). 4 figs., 4 tabs., 39 refs

  18. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-01-01

    The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

  19. Molten salt cooling/17Li-83Pb breeding blanket concept

    International Nuclear Information System (INIS)

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

    1985-02-01

    A description of a fusion breeding blanket concept using draw salt coolant and static 17 Li- 83 Pb is presented. 17 Li- 83 Pb has high breeding capability and low tritium solubility. Draw salt operates at low pressure and is inert to water. Corrosion, MHD, and tritium containment problems associated with the MARS design are alleviated because of the use of a static LiPb blanket. Blanket tritium recovery is by permeation toward the plasma. A direct contact steam generator is proposed to eliminate some generic problems associated with a tube shell steam generator

  20. Conceptual design of blanket structures for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

  1. Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

    International Nuclear Information System (INIS)

    1978-01-01

    This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified

  2. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

    1989-01-01

    The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

  3. Cassette blanket and vacuum building: key elements in fusion reactor maintenance

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

    The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

  4. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

    2016-01-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  5. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

    2016-11-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  6. Evaluation of transmutation performance of long-lived fission products with a super fast reactor

    International Nuclear Information System (INIS)

    Lu, Haoliang; Han, Chiyoung; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    The performance of the Super Fast Reactor for transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super Fast Reactor. First is in the blanket assembly due to the ZrH 1.7 layer which can slow down the fast neutrons. Second is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected of transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe·y and 2.79%/GWe.y can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the outputs from 11.8 and 6.2 1000MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained by the Super Fast Reactor. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super Fast Reactor. It turns out that the 135 Cs transmutation is a challenge not only for the Super Fast Reactor but also for other commercial fast reactors. (author)

  7. Insulation Reformulation Development

    Science.gov (United States)

    Chapman, Cynthia; Bray, Mark

    2015-01-01

    The current Space Launch System (SLS) internal solid rocket motor insulation, polybenzimidazole acrylonitrile butadiene rubber (PBI-NBR), is a new insulation that replaced asbestos-based insulations found in Space Shuttle heritage solid rocket boosters. PBI-NBR has some outstanding characteristics such as an excellent thermal erosion resistance, low thermal conductivity, and low density. PBI-NBR also has some significant challenges associated with its use: Air entrainment/entrapment during manufacture and lay-up/cure and low mechanical properties such as tensile strength, modulus, and fracture toughness. This technology development attempted to overcome these challenges by testing various reformulated versions of booster insulation. The results suggest the SLS program should continue to investigate material alternatives for potential block upgrades or use an entirely new, more advanced booster. The experimental design was composed of a logic path that performs iterative formulation and testing in order to maximize the effort. A lab mixing baseline was developed and documented for the Rubber Laboratory in Bldg. 4602/Room 1178.

  8. Beyond insulation and isolation

    DEFF Research Database (Denmark)

    Højlund, Marie Koldkjær

    2016-01-01

    are insulation and isolation strategies to reduce measurable and perceptual noise levels. However, these strategies do not actively support the need to feel like an integral part of the shared hospital environment, which is a key element in creating healing environments, according to the paradigm of Evidence-Based...

  9. Self-Healing Wire Insulation

    Science.gov (United States)

    Parrish, Clyde F. (Inventor)

    2012-01-01

    A self-healing system for an insulation material initiates a self-repair process by rupturing a plurality of microcapsules disposed on the insulation material. When the plurality of microcapsules are ruptured, reactants within the plurality of microcapsules react to form a replacement polymer in a break of the insulation material. This self-healing system has the ability to repair multiple breaks in a length of insulation material without exhausting the repair properties of the material.

  10. Impressive Super Phenix

    International Nuclear Information System (INIS)

    Olds, F.C.

    1979-01-01

    The 1200-MWe fast breeder reactor, Super Phenix at Creys-Malville, is scheduled for commercial operation in 1983. This is the world's first near-commercial-sized fast breeder. As a near-commercial-sized unit, it represents essentially the technology and hardware of the first fully commercial follow-on units. In its size, its components, its design, the technology it represents, and its project schedule, it is impressive. As of May 1979, the Super Phenix nuclear steam boiler in the Creys-Malville plant bore an estimated cost of $700 million, without fuel. The total cost of the Creys-Malville plant now is estimated at about $1.4 billion. This is about twice the cost of a comparable standardized PWR being built in France today. However, it should be borne in mind that Creys-Malville carries the high cost of a first-of-the-line prototype, and that France's PWRs are standardized, second-generation units. Electricity from Creys-Malville is estimated to cost a little more than electricity would cost from a coal-fired plant complete with flue gas scrubbing

  11. Investigation of thermal transfers in super-fluid helium in porous media

    International Nuclear Information System (INIS)

    Allain, H.

    2009-10-01

    Particle accelerators are requiring increased magnetic fields for which niobium tin superconducting magnets are considered. This entails electric insulation and cooling problems. Porous ceramic insulations are potential candidates for cable insulation. As they are permeable to helium, they could allow a direct cooling by super-fluid helium. Therefore, this research thesis deals with the investigation of thermal transfers in superfluid helium in porous media. After a description of an accelerator's superconducting magnet, of its thermodynamics and its various cooling modes, the author describes the physical properties of super-fluid helium, its peculiarities with respect to conventional fluids as well as its different phases (fluid and super-fluid), its dynamics under different regimes (the Landau regime which is similar to the laminar regime for a conventional fluid, and the Gorter-Mellink regime which is the super-fluid turbulent regime). He determines the macroscopic equations governing the He II dynamics in porous media by applying the volume averaging method developed by Whitaker. Theoretical results are validated by comparison with a numerical analysis performed with a numerical code. Then, the author presents the various experimental setups which have been developed for the measurement of the intrinsic permeability, one at room temperature and another at high temperature. Experimental results are discussed, notably with respect to pore size and porosity

  12. Trial fabrication and preliminary characterization of electrical insulator for liquid metal system

    International Nuclear Information System (INIS)

    Nakamichi, Masaru; Kawamura, Hiroshi; Oyamada, Rokuro

    1995-03-01

    In the design of the liquid metal blanket, MHD pressure drop is one of critical issues. Ceramic coating on the surface of structural material is considered as an electrical insulator to reduce the MHD pressure drop. Ceramic coating such as Y 2 O 3 is a promising electrical insulator due to its high electrical resistivity and good compatibility with liquid lithium. This report describes the trial fabrication and preliminary characterization of electrical insulator for a design study of the liquid metal system. From the results of trial fabrication and preliminary characterization, it is concluded that densified atmospheric plasma spray Y 2 O 3 coating with 410SS undercoating between 316SS substrate and Y 2 O 3 coating is suitable for Y 2 O 3 coating fabrication. (author)

  13. Development of electrical insulator coatings for fusion power applications

    International Nuclear Information System (INIS)

    Park, J.H.; Domenico, T.; Dragel, G.; Clark, R.

    1995-01-01

    In the design of liquid-metal cooling systems for fusion blanket applications, the corrosion resistance of structural materials and the magnetohydrodynamic (MHD) force and its subsequent influence on thermal hydraulics and corrosion are major concerns. The objective of this study was to develop stable corrosion-resistant electrical insulator coatings at the liquid-metal-structural-material interface, with emphasis on electrically insulating coatings that prevent adverse MHD-generated currents from passing through the structural walls. Vanadium and V-base alloys (V-Ti or V-Ti-Cr) are leading candidate materials for structural applications in fusion reactors. When the system is cooled by liquid metals, insulator coatings are required on piping surfaces in contact with the coolant. Various intermetallic films were produced on V, V-5Ti, and V-20Ti, V-5Cr-5Ti, and V-15Cr-5Ti, and Ti, and on types 304 and 316 stainless steel. The intermetallic layers were developed by exposure of the materials to liquid Li containing 3-5at.% dissolved metallic solute (e.g. Al, Be, Mg, Si, Ca, Pt, and Cr) at temperatures of 416-880 C. Subsequently, electrical insulator coatings were produced by reaction of the reactive layers with dissolved N in liquid Li or by air oxidation under controlled conditions at 600-1000 C. These reactions converted the intermetallic layers to electrically insulating oxide-nitride or oxynitride layers. This coating method is applicable to reactor components. The liquid metal can be used over and over because only the solutes are consumed within the liquid metal. The technique can be applied to various shapes (e.g. inside or outside of tubes, complex geometrical shapes) because the coating is formed by liquid-phase reaction. This paper discusses initial results on the nature of the coatings (composition, thickness, adhesion, surface coverage) and their in situ electrical resistivity characteristics in liquid Li at high temperatures. (orig.)

  14. Melting in super-earths.

    Science.gov (United States)

    Stixrude, Lars

    2014-04-28

    We examine the possible extent of melting in rock-iron super-earths, focusing on those in the habitable zone. We consider the energetics of accretion and core formation, the timescale of cooling and its dependence on viscosity and partial melting, thermal regulation via the temperature dependence of viscosity, and the melting curves of rock and iron components at the ultra-high pressures characteristic of super-earths. We find that the efficiency of kinetic energy deposition during accretion increases with planetary mass; considering the likely role of giant impacts and core formation, we find that super-earths probably complete their accretionary phase in an entirely molten state. Considerations of thermal regulation lead us to propose model temperature profiles of super-earths that are controlled by silicate melting. We estimate melting curves of iron and rock components up to the extreme pressures characteristic of super-earth interiors based on existing experimental and ab initio results and scaling laws. We construct super-earth thermal models by solving the equations of mass conservation and hydrostatic equilibrium, together with equations of state of rock and iron components. We set the potential temperature at the core-mantle boundary and at the surface to the local silicate melting temperature. We find that ancient (∼4 Gyr) super-earths may be partially molten at the top and bottom of their mantles, and that mantle convection is sufficiently vigorous to sustain dynamo action over the whole range of super-earth masses.

  15. 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 ...

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  17. Blanket options for high-efficiency fusion power

    International Nuclear Information System (INIS)

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

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  18. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  19. Fusion blanket for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperature (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by Ar) utilizing Li 2 O for tritium breeding. In this design, approx. 60% of the fusion energy is deposited in the high-temperature interior. The maximum Ar temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  20. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1981-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 deg C) of conventional structural materials such as stainless steels. In this project 'two-zone' blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 deg C leading to an overall efficiency estimate of 55 to 60% for this reference case. (author)

  1. Fusion-reactor blanket-material safety-compatibility studies

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

    1982-11-01

    Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO 2 , Li 2 ZrO 3 , Li 2 SiO 3 , Li 4 SiO 4 and LiTiO 3 ) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li 7 Pb 2 alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li 17 Pb 83 alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li 17 Pb 83 alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns

  2. Nuclear characteristics of D-D fusion reactor blankets, (1)

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao; Seki, Yasushi.

    1977-01-01

    Fusion reactors operating on the deuterium (D-D) cycle are considered promising for their freedom from tritium breeding in the blanket. In this paper, neutronic and photonic calculations are undertaken covering several blanket models of the D-D fusion reactor, using presently available data, with a view to comparing the nuclear characteristics of these models, in particular, the nuclear heating rates and their spatial distributions. Nine models are taken up for the study, embodying various combinations of coolant, blanket, structural and reflector materials. About 10 MeV is found to be a typical value for the total nuclear energy deposition per source neutron in the models considered here. The realization of high energy gain is contingent upon finding a favorable combination of blanket composition and configuration. The resulting implications on the thermal design aspect are briefly discussed. (auth.)

  3. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

    A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

  4. Blast venting through blanket material in the HYLIFE ICF reactor

    International Nuclear Information System (INIS)

    Liu, J.C.; Peterson, P.F.; Schrock, V.E.

    1992-01-01

    This work presents a numerical study of blast venting through various blanket configurations in the HYLIFE ICF reactor design. The study uses TSUNAMI -- a multi-dimensional, high-resolution, shock capturing code -- to predict the momentum exchange and gas dynamics for blast venting in complex geometries. In addition, the study presents conservative predictions of wall loading by gas shock and impulse delivered to the protective liquid blanket. Configurations used in the study include both 2700 MJ and 350 MJ fusion yields per pulse for 5 meter and 3 meter radius reactor chambers. For the former, an annular jet array is used for the blanket geometry, while in the latter, both annular jet array as well as slab geometries are used. Results of the study indicate that blast venting and wall loading may be manageable in the HYLIFE-II design by a judicious choice of blanket configuration

  5. 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

  6. 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.

  7. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

    Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

    1982-01-01

    Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

  8. Recent developments in fusion first wall, blanket, and shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-01-01

    This brief overview of first wall, blanket and shield technology reviews the changes and trends in important design issues in first wall, blanket and shield design and related technology from the 1970's to the 1980's. The emphasis is on base technology rather than either systems engineering or materials development. The review is limited to the two primary confinement systems, tokamaks and mirrors, and production of electricity as the primary goal for development

  9. 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

  10. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-02-01

    A review is given of the features and potentialities of cold-blanket systems, with respect to plasma equilibrium, stability, and reactor technology. The treatment is concentrated on quasi-steady magnetized plasmas confined at moderately high beta values. The cold-blanket concept has specific potentialities as a fusion reactor, e.g. in connection with the desired densities and dimensions of full-scale systems, refuelling, as well as ash and impurity removal, and stability. (author)

  11. 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.

  12. 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.

  13. Overview of first wall/blanket/shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-04-01

    This brief overview of first wall, blanket, and shield technology focuses first on changes and trends in important design issues from the 1970's to the 1980's, then on current perceptions of critical issues in first wall, blanket, and shield design and related technology. The emphasis is on base technology rather than either systems engineering or materials development, on the two primary confinement systems, tokamaks and mirrors, and on production of electricity as the primary goal for development

  14. Applications of the Aqueous Self-Cooled Blanket concept

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.J.; Varsamis, G.; Wrisley, K.; Deutch, L.; Gierszewski, P.

    1986-01-01

    In this paper a novel water-cooled blanket concept is examined. This concept, designated the Aqueous Self-Cooled Blanket (ASCB), employs water with small amounts of dissolved fertile compounds as both the coolant and the breeding medium. The ASCB concept is reviewed and its application in three different contexts is examined: (1) power reactors; (2) near-term devices such as NET; and (3) fusion-fission hybrids

  15. Progress on DEMO blanket attachment concept with keys and pins

    International Nuclear Information System (INIS)

    Vizvary, Zsolt; Iglesias, Daniel; Cooper, David; Crowe, Robert; Riccardo, Valeria

    2015-01-01

    Highlights: • DEMO blanket attachment system with keys and pins (without using bolts). • Blanket segments are preloaded by progressively designed springs. • Blanket back plate flexibility has a major impact on spring design. • Mechanical analysis of other components indicates no unresolvable issues. • Thermal analysis indicates acceptable temperatures for the support system. - Abstract: The blanket attachment has to cope with gravity, thermal and electromagnetic loads, also it has to be installed and serviced by remote handling. Pre-stressed components suffer from stress relaxation in irradiated environments such as DEMO. To circumvent this problem pre-stressed component should be either avoided or shielded, and where possible keys and pins should be used. This strategy has been proposed for the DEMO multi-module segments (MMS). The blanket segments are held by two tapered keys each, designed to allow thermal expansions while providing contact with the vacuum vessel and to resist the poloidal and radial moments the latter being dominant at 9.1 MNm inboard and 15 MNm outboard. On the top of the blanket segment there is a pin which provides vertical support. At the bottom another vertical support has to lock them in position after installation and manage the pre-load on the segments. The pre-load is required to deal with the electromagnetic loads during disruption. This is provided by a set of springs, which require shielding as they are preloaded. These are sized to cope with the force (3 MN inboard, 1.4 MN outboard) due to halo currents and the toroidal moment which can reverse. Calculations show that the flexibility of the blanket segment itself plays a significant role in defining the required support system. The blanket segment acts as a preloaded spring and it has to be part of the attachment design as well.

  16. 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

  17. Fully developed magnetohydrodynamic flows in rectangular ducts with insulating walls

    International Nuclear Information System (INIS)

    Molokov, S.; Kernforschungszentrum Karlsruhe GmbH; Shishko, A.

    1993-10-01

    In the first part the effect of magnetic field inclination on the flow structure and the pressure drop is considered. The duct walls are insulating. An asymptotic solution to the problem at high Hartmann numbers is obtained. The results show that for a square duct the increase of the pressure gradient due to the field inclination is negligible (less than 10% for any angle). For blanket relevant values of inclination of up to 10 the deviation of the velocity profile from the slug profile is insignificant. The second part studies the flow in a duct with insulating walls parallel to the magnetic field, while the Hartmann walls are covered by an insulating coating. A new type of the boundary condition is derived, which takes into account finite coating resistance. The effect of the latter on the flow characteristics is studied. An exact solution to the problem is obtained and several approximate formulas for the pressure drop at high Hartmann numbers are presented. (orig./HP) [de

  18. 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)

  19. 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.)

  20. Studies on steps affecting tritium residence time in solid blanket

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

    For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

  1. Metal-insulator transitions

    Science.gov (United States)

    Imada, Masatoshi; Fujimori, Atsushi; Tokura, Yoshinori

    1998-10-01

    Metal-insulator transitions are accompanied by huge resistivity changes, even over tens of orders of magnitude, and are widely observed in condensed-matter systems. This article presents the observations and current understanding of the metal-insulator transition with a pedagogical introduction to the subject. Especially important are the transitions driven by correlation effects associated with the electron-electron interaction. The insulating phase caused by the correlation effects is categorized as the Mott Insulator. Near the transition point the metallic state shows fluctuations and orderings in the spin, charge, and orbital degrees of freedom. The properties of these metals are frequently quite different from those of ordinary metals, as measured by transport, optical, and magnetic probes. The review first describes theoretical approaches to the unusual metallic states and to the metal-insulator transition. The Fermi-liquid theory treats the correlations that can be adiabatically connected with the noninteracting picture. Strong-coupling models that do not require Fermi-liquid behavior have also been developed. Much work has also been done on the scaling theory of the transition. A central issue for this review is the evaluation of these approaches in simple theoretical systems such as the Hubbard model and t-J models. Another key issue is strong competition among various orderings as in the interplay of spin and orbital fluctuations. Experimentally, the unusual properties of the metallic state near the insulating transition have been most extensively studied in d-electron systems. In particular, there is revived interest in transition-metal oxides, motivated by the epoch-making findings of high-temperature superconductivity in cuprates and colossal magnetoresistance in manganites. The article reviews the rich phenomena of anomalous metallicity, taking as examples Ti, V, Cr, Mn, Fe, Co, Ni, Cu, and Ru compounds. The diverse phenomena include strong spin and

  2. Superconducting super collider

    International Nuclear Information System (INIS)

    Limon, P.J.

    1987-01-01

    The Superconducting Super Collider is to be a 20 TeV per beam proton-proton accelerator and collider. Physically the SCC will be 52 miles in circumference and slightly oval in shape. The use of superconducting magnets instead of conventional cuts the circumference from 180 miles to the 52 miles. The operating cost of the SCC per year is estimated to be about $200-250 million. A detailed cost estimate of the project is roughly $3 billion in 1986 dollars. For the big collider ring, the technical cost are dominated by the magnet system. That is why one must focus on the cost and design of the magnets. Presently, the process of site selection is underway. The major R and D efforts concern superconducting dipoles. The magnets use niobium-titanium as a conductor stabilized in a copper matrix. 10 figures

  3. 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)

  4. The Super DREAM Project

    Energy Technology Data Exchange (ETDEWEB)

    Wigmans, Richard [Texas Tech Univ., Lubbock, TX (United States)

    2017-09-25

    Despite the fact that DOE provided only a fraction of the requested funds, the goals we defined in the proposal on which award ER41783 was based were essentially all met. This was partially due to the fact that other funding agencies, which supported our collaborators (especially from Italy and Korea) contributed as well, and partially due to the effective solutions that were developed to compensate for the fact that the detector we had proposed to build had to be scaled down. The performance of the SuperDREAM calorimeter is better than anything that has been built or proposed so far. This has of course not gone unnoticed in the scientific community. Scientists who are preparing experiments for the proposed new generation of particle accelerators (FCCee, CPEC,..) are all very seriously considering the technology developed in this project. Several new collaborations have formed which aim to adapt the dual-readout calorimeter principles to the demands of a 4 environment. Preliminary measurements using silicon photomultipliers as light sensors have already been carried out. This type of readout would make it possible to operate this detector in a magnetic field, and it would also allow for a longitudinal segmentation into electromagnetic and hadronic sections, if so desired. In addition, SiPM readout would eliminate the need for “forests” of fibers sticking out of the rear end of the calorimeter (Figure 1), and obtain an arbitrary fine lateral segmentation, which might be very important for recognizing electrons inside jets. The improvements in our understanding of the fundamental structure of matter and the forces that govern its behavior have always hinged on the availability of detectors that make it possible to explore the possibilities of new, more powerful particle accelerators to the fullest extent. We believe that the SuperDREAM project has created a quantum leap in detector technology, which may turn out to be crucially important for future discoveries in

  5. Heat insulating plates

    Energy Technology Data Exchange (ETDEWEB)

    Allan, J.A.F.

    1976-10-28

    Micro-porous insulation plates are dealt with, for example, how they are used in the insulation of heat storage devices. Since one side of such plates is exposed to a temperature of over 700/sup 0/C, a shrinkage of the glass texture of the covering can occur, which can exceed the shrinkage of the inner micro-porous material, so that cracks and splits in the high temperature side of the covering can come about. The task of the invention is to design the plate in such a way as to prevent this from happening. For this purpose the plate is provided, according to invention specifications, with flutes, waves, ribs, waffle or grid patterns and the covering is set into the recesses originating from this.

  6. Green insulation: hemp fibers

    Energy Technology Data Exchange (ETDEWEB)

    Anon,

    2011-09-15

    Indian hemp (Cannabis indica) is known for its psychotropic values and it is banned in most countries. However, industrial hemp (Cannabis sativa) is known for its tough fibers. Several manufactures in Europe including, small niche players, have been marketing hemp insulation products for several years. Hemp is a low environmental impact material. Neither herbicide nor pesticide is used during the growth of hemp. The fibers are extracted in a waste-free and chemical-free mechanical process. Hemp can consume CO2 during its growth. In addition, hemp fiber can be disposed of harmlessly by composting or incineration at the end of its life. Hemp fibers are processed and treated only minimally to resist rot and fungal activity. There is little health risk when producing and installing the insulation, thanks to the absence of toxic additive. Its thermal resistance is comparable to mineral wool. But the development and marketing of hemp fibers may be restricted in North America.

  7. Compact vacuum insulation embodiments

    Science.gov (United States)

    Benson, D.K.; Potter, T.F.

    1992-04-28

    An ultra-thin compact vacuum insulation panel is comprised of two hard, but bendable metal wall sheets closely spaced apart from each other and welded around the edges to enclose a vacuum chamber. Glass or ceramic spacers hold the wall sheets apart. The spacers can be discrete spherical beads or monolithic sheets of glass or ceramic webs with nodules protruding therefrom to form essentially point' or line' contacts with the metal wall sheets. In the case of monolithic spacers that form line' contacts, two such spacers with the line contacts running perpendicular to each other form effectively point' contacts at the intersections. Corrugations accommodate bending and expansion, tubular insulated pipes and conduits, and preferred applications are also included. 26 figs.

  8. Compact vacuum insulation

    Science.gov (United States)

    Benson, D.K.; Potter, T.F.

    1993-01-05

    An ultra-thin compact vacuum insulation panel is comprised of two hard, but bendable metal wall sheets closely spaced apart from each other and welded around the edges to enclose a vacuum chamber. Glass or ceramic spacers hold the wall sheets apart. The spacers can be discrete spherical beads or monolithic sheets of glass or ceramic webs with nodules protruding therefrom to form essentially point'' or line'' contacts with the metal wall sheets. In the case of monolithic spacers that form line'' contacts, two such spacers with the line contacts running perpendicular to each other form effectively point'' contacts at the intersections. Corrugations accommodate bending and expansion, tubular insulated pipes and conduits, and preferred applications are also included.

  9. Insulating materials for optoelectronics

    International Nuclear Information System (INIS)

    Agullo-Lopez, F.

    1990-01-01

    Optoelectronics is an interdisciplinary field. Basic functions of an optoelectronic system include the generator of the optical signal, its transmission and handling and, finally, its detection, storage and display. A large variety of semiconductor and insulating materials are used or are being considered to perform those functions. The authors focus on insulating materials, mostly oxides. For signal generation, tunable solid state lasers, either vibronic or those based oon colour centres are briefly described, and their main operating parameters summarized. Reference is made to some developments on fiber and waveguide lasers. Relevant physical features of the silica fibres used for low-loss, long-band, optical transmission are reviewed, as well as present efforts to further reduce attenuation in the mid-infrared range. Particular attention is paid to photorefractive materials (LiNbO 3 , BGO, BSO, etc.), which are being investigated

  10. Insulated pipe clamp design

    International Nuclear Information System (INIS)

    Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

    1980-01-01

    Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. The design considerations and methods along with the development tests are presented. Special considerations to guard against adverse cracking of the insulation material, to maintain the clamp-pipe stiffness desired during a seismic event, to minimize clamp restraint on the pipe during normal pipe heatup, and to resist clamp rotation or spinning on the pipe are emphasized

  11. Insulated pipe clamp design

    International Nuclear Information System (INIS)

    Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

    1980-01-01

    Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. 5 refs

  12. Pourable Foam Insulation

    Science.gov (United States)

    Harvey, James A.; Butler, John M.; Chartoff, Richard P.

    1989-01-01

    Report describes search for polyisocyanurate/polyurethane foam insulation with superior characteristics. Discusses chemistry of current formulations. Tests of formulations, of individual ingredients and or alternative new formulations described. Search revealed commercially available formulations exhibiting increased thermal stability at temperatures up to 600 degree C, pours readily before curing, presents good appearance after curing, and remains securely bonded to aluminum at cryogenic temperatures. Total of 42 different formulations investigated, 10 found to meet requirements.

  13. Grassmann, super-Kac-Moody and super-derivation algebras

    International Nuclear Information System (INIS)

    Frappat, L.; Ragoucy, E.; Sorba, P.

    1989-05-01

    We study the cyclic cocycles of degree one on the Grassmann algebra and on the super-circle with N supersymmetries (i.e. the tensor product of the algebra of functions on the circle times a Grassmann algebra with N generators). They are related to central extensions of graded loop algebras (i.e. super-Kac-Moody algebras). The corresponding algebras of super-derivations have to be compatible with the cocycle characterizing the extension; we give a general method for determining these algebras and examine in particular the cases N = 1,2,3. We also discuss their relations with the Ademollo et al. algebras, and examine the possibility of defining new kinds of super-conformal algebras, which, for N > 1, generalize the N = 1 Ramond-Neveu-Schwarz algebra

  14. Electrically tuned super-capacitors

    OpenAIRE

    Chowdhury, Tazima S.; Grebel, Haim

    2015-01-01

    Fast charging and discharging of large amounts of electrical energy make super-capacitors ideal for short-term energy storage [1-5]. In its simplest form, the super-capacitor is an electrolytic capacitor made of an anode and a cathode immersed in an electrolyte. As for an ordinary capacitor, minimizing the charge separation distance and increasing the electrode area increase capacitance. In super-capacitors, charge separation is of nano-meter scale at each of the electrode interface (the Helm...

  15. The Super-Kamiokande detector

    International Nuclear Information System (INIS)

    Fukuda, S.; Fukuda, Y.; Hayakawa, T.; Ichihara, E.; Ishitsuka, M.; Itow, Y.; Kajita, T.; Kameda, J.; Kaneyuki, K.; Kasuga, S.; Kobayashi, K.; Kobayashi, Y.; Koshio, Y.; Miura, M.; Moriyama, S.; Nakahata, M.; Nakayama, S.; Namba, T.; Obayashi, Y.; Okada, A.; Oketa, M.; Okumura, K.; Oyabu, T.; Sakurai, N.; Shiozawa, M.; Suzuki, Y.; Takeuchi, Y.; Toshito, T.; Totsuka, Y.; Yamada, S.; Desai, S.; Earl, M.; Hong, J.T.; Kearns, E.; Masuzawa, M.; Messier, M.D.; Stone, J.L.; Sulak, L.R.; Walter, C.W.; Wang, W.; Scholberg, K.; Barszczak, T.; Casper, D.; Liu, D.W.; Gajewski, W.; Halverson, P.G.; Hsu, J.; Kropp, W.R.; Mine, S.; Price, L.R.; Reines, F.; Smy, M.; Sobel, H.W.; Vagins, M.R.; Ganezer, K.S.; Keig, W.E.; Ellsworth, R.W.; Tasaka, S.; Flanagan, J.W.; Kibayashi, A.; Learned, J.G.; Matsuno, S.; Stenger, V.J.; Hayato, Y.; Ishii, T.; Ichikawa, A.; Kanzaki, J.; Kobayashi, T.; Maruyama, T.; Nakamura, K.; Oyama, Y.; Sakai, A.; Sakuda, M.; Sasaki, O.; Echigo, S.; Iwashita, T.; Kohama, M.; Suzuki, A.T.; Hasegawa, M.; Inagaki, T.; Kato, I.; Maesaka, H.; Nakaya, T.; Nishikawa, K.; Yamamoto, S.; Haines, T.J.; Kim, B.K.; Sanford, R.; Svoboda, R.; Blaufuss, E.; Chen, M.L.; Conner, Z.; Goodman, J.A.; Guillian, E.; Sullivan, G.W.; Turcan, D.; Habig, A.; Ackerman, M.; Goebel, F.; Hill, J.; Jung, C.K.; Kato, T.; Kerr, D.; Malek, M.; Martens, K.; Mauger, C.; McGrew, C.; Sharkey, E.; Viren, B.; Yanagisawa, C.; Doki, W.; Inaba, S.; Ito, K.; Kirisawa, M.; Kitaguchi, M.; Mitsuda, C.; Miyano, K.; Saji, C.; Takahata, M.; Takahashi, M.; Higuchi, K.; Kajiyama, Y.; Kusano, A.; Nagashima, Y.; Nitta, K.; Takita, M.; Yamaguchi, T.; Yoshida, M.; Kim, H.I.; Kim, S.B.; Yoo, J.; Okazawa, H.; Etoh, M.; Fujita, K.; Gando, Y.; Hasegawa, A.; Hasegawa, T.; Hatakeyama, S.; Inoue, K.; Ishihara, K.; Iwamoto, T.; Koga, M.; Nishiyama, I.; Ogawa, H.; Shirai, J.; Suzuki, A.; Takayama, T.; Tsushima, F.; Koshiba, M.; Ichikawa, Y.; Hashimoto, T.; Hatakeyama, Y.; Koike, M.; Horiuchi, T.; Nemoto, M.; Nishijima, K.; Takeda, H.; Fujiyasu, H.; Futagami, T.; Ishino, H.; Kanaya, Y.; Morii, M.; Nishihama, H.; Nishimura, H.; Suzuki, T.; Watanabe, Y.; Kielczewska, D.; Golebiewska, U.; Berns, H.G.; Boyd, S.B.; Doyle, R.A.; George, J.S.; Stachyra, A.L.; Wai, L.L.; Wilkes, R.J.; Young, K.K.; Kobayashi, H.

    2003-01-01

    Super-Kamiokande is the world's largest water Cherenkov detector, with net mass 50,000 tons. During the period April, 1996 to July, 2001, Super-Kamiokande I collected 1678 live-days of data, observing neutrinos from the Sun, Earth's atmosphere, and the K2K long-baseline neutrino beam with high efficiency. These data provided crucial information for our current understanding of neutrino oscillations, as well as setting stringent limits on nucleon decay. In this paper, we describe the detector in detail, including its site, configuration, data acquisition equipment, online and offline software, and calibration systems which were used during Super-Kamiokande I

  16. 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)

  17. SuperB Progress Report: Detector

    International Nuclear Information System (INIS)

    Grauges, E.; Donvito, G.; Spinoso, V.; Manghisoni, M.; Re, V.; Traversi, G.; Eigen, G.; Fehlker, D.; Helleve, L.; Cheng, C.; Chivukula, A.; Doll, D.; Echenard, B.; Hitlin, D.; Ongmongkolkul, P.; Porter, F.; Rakitin, A.; Thomas, M.; Zhu, R.; Tatishvili, G.; Andreassen, R.; Fabby, C.; Meadows, B.; Simpson, A.; Sokoloff, M.; Tomko, K.; Fella, A.; Andreotti, M.; Baldini, W.; Calabrese, R.; Carassiti, V.; Cibinetto, G.; Cotta Ramusino, A.; Gianoli, A.; Luppi, E.; Munerato, M.; Santoro, V.; Tomassetti, L.; Stoker, D.; Bezshyyko, O.; Dolinska, G.; Arnaud, N.; Beigbeder, C.; Bogard, F.; Breton, D.; Burmistrov, L.; Charlet, D.; Maalmi, J.; Perez Perez, L.; Puill, V.; Stocchi, A.; Tocut, V.; Wallon, S.; Wormser, G.; Brown, D.

    2012-01-01

    This report describes the present status of the detector design for SuperB. It is one of four separate progress reports that, taken collectively, describe progress made on the SuperB Project since the publication of the SuperB Conceptual Design Report in 2007 and the Proceedings of SuperB Workshop VI in Valencia in 2008.

  18. SuperB Progress Report: Detector

    Energy Technology Data Exchange (ETDEWEB)

    Grauges, E.; /Barcelona U., ECM; Donvito, G.; Spinoso, V.; /INFN, Bari /Bari U.; Manghisoni, M.; Re, V.; Traversi, G.; /INFN, Pavia /Bergamo U., Ingengneria Dept.; Eigen, G.; Fehlker, D.; Helleve, L.; /Bergen U.; Carbone, A.; Di Sipio, R.; Gabrielli, A.; Galli, D.; Giorgi, F.; Marconi, U.; Perazzini, S.; Sbarra, C.; Vagnoni, V.; Valentinetti, S.; Villa, M.; Zoccoli, A.; /INFN, Bologna /Bologna U. /Caltech /Carleton U. /Cincinnati U. /INFN, CNAF /INFN, Ferrara /Ferrara U. /UC, Irvine /Taras Shevchenko U. /Orsay, LAL /LBL, Berkeley /UC, Berkeley /Frascati /INFN, Legnaro /Orsay, IPN /Maryland U. /McGill U. /INFN, Milan /Milan U. /INFN, Naples /Naples U. /Novosibirsk, IYF /INFN, Padua /Padua U. /INFN, Pavia /Pavia U. /INFN, Perugia /Perugia U. /INFN, Perugia /Caltech /INFN, Pisa /Pisa U. /Pisa, Scuola Normale Superiore /PNL, Richland /Queen Mary, U. of London /Rutherford /INFN, Rome /Rome U. /INFN, Rome2 /Rome U.,Tor Vergata /INFN, Rome3 /Rome III U. /SLAC /Tel Aviv U. /INFN, Turin /Turin U. /INFN, Padua /Trento U. /INFN, Trieste /Trieste U. /TRIUMF /British Columbia U. /Montreal U. /Victoria U.

    2012-02-14

    This report describes the present status of the detector design for SuperB. It is one of four separate progress reports that, taken collectively, describe progress made on the SuperB Project since the publication of the SuperB Conceptual Design Report in 2007 and the Proceedings of SuperB Workshop VI in Valencia in 2008.

  19. SuperB Progress Reports Accelerator

    CERN Document Server

    Biagini, Maria Enrica; Boscolo, M; Buonomo, B; Demma, T; Drago, A; Esposito, M; Guiducci, S; Mazzitelli, G; Pellegrino, L; Preger, M A; Raimondi, P; Ricci, R; Rotundo, U; Sanelli, C; Serio, M; Stella, A; Tomassini, S; Zobov, M; Bertsche, K; Brachman, A; Cai, Y; Chao, A; Chesnut, R; Donald, M.H; Field, C; Fisher, A; Kharakh, D; Krasnykh, A; Moffeit, K; Nosochkov, Y; Pivi, M; Seeman, J; Sullivan, M.K; Weathersby, S; Weidemann, A; Weisend, J; Wienands, U; Wittmer, W; Woods, M; Yocky, G; Bogomiagkov, A; Koop, I; Levichev, E; Nikitin, S; Okunev, I; Piminov, P; Sinyatkin, S; Shatilov, D; Vobly, P; Bosi, F; Liuzzo, S; Paoloni, E; Bonis, J; Chehab, R; Le Meur, G; Lepercq, P; Letellier-Cohen, F; Mercier, B; Poirier, F; Prevost, C; Rimbault, C; Touze, F; Variola, A; Bolzon, B; Brunetti, L; Jeremie, A; Baylac, M; Bourrion, O; De Conto, J M; Gomez, Y; Meot, F; Monseu, N; Tourres, D; Vescovi, C; Chanci, A; Napoly, O; Barber, D P; Bettoni, S; Quatraro, D

    2010-01-01

    This report details the present status of the Accelerator design for the SuperB Project. It is one of four separate progress reports that, taken collectively, describe progress made on the SuperB Project since the publication of the SuperB Conceptual Design Report in 2007 and the Proceedings of SuperB Workshop VI in Valencia in 2008.

  20. 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

  1. MHD pressure drop in ducts with imperfectly insulating coatings

    International Nuclear Information System (INIS)

    Malang, S.; Buehler, L.

    1994-08-01

    Liquid metal cooled blankets in fusion tokamak's are feasible only with electrically insulating coatings at the coolant channel walls. The requirements of such coatings are investigated and a simple analytical model is developed to determine the influence of imperfections in the coatings on the magneto-hydrodynamic pressure drop. This model is compared with the results of a 3D-MHD code based on the core flow approach. Both methods are in good agreement as long as the imperfections do not increase the pressure drop by more than 20%. The analytical model over-estimates the pressure drop for values larger than 20%. The importance of self-healing of coatings in case of cracking or flaking is quantified and an equation for the equilibrium conditions between the generation of imperfection and the healing of such spots is provided

  2. Reduction of heat insulation upon soaking of the insulation layer

    Science.gov (United States)

    Achtliger, J.

    1983-09-01

    Improved thermal protection of hollow masonry by introduction of a core insulation between the inner and outer shell is discussed. The thermal conductivity of insulation materials was determined in dry state and after soaking by water with different volume-related moisture contents. The interpolated thermal conductivity values from three measured values at 10 C average temperature are presented as a function of the pertinent moisture content. Fills of expanded polystyrene, perlite and granulated mineral fibers, insulating boards made of mineral fibers and in situ cellular plastics produced from urea-formaldehyde resin were investigated. Test results show a confirmation of thermal conductivity values for insulating materials in hollow masonry.

  3. Study on the LLFPs transmutation in a super-critical water-cooled fast reactor

    International Nuclear Information System (INIS)

    Lu Haoliang; Ishiwatari, Yuki; Oka, Yoshiaki

    2011-01-01

    Research highlights: → Transmutation of LLFPs with a super-criticial water cooled fast reactor. → Transmutation of iodine and cesium without the isotopic separation. → The transmuted isotope was mixed with UO 2 to reduce the effect of self-shielding. → A weak neutron moderator Al 2 O 3 was used to suppress the creation of 135 Cs from 133 Cs. - Abstract: The performance of the super-critical water-cooled fast reactor (Super FR) for the transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with the soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super FR. First region is in the blanket assembly due to the ZrH 1.7 layer which was utilized to slow down the fast neutrons to achieve a negative void reactivity. Second region is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected in the transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR or fast reactor. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered to avoid the separation. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe year and 2.79%/GWe year can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the yields from 11.8 and 6.2 1000 MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000 MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained in the Super FR. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super FR. It turns out that the

  4. Super boson-fermion correspondence

    International Nuclear Information System (INIS)

    Kac, V.G.; Leur van de, J.W.

    1987-01-01

    Since the pioneering work of Skyrme, the boson-fermion correspondence has been playing an increasingly important role in 2-dimensional quantum field theory. More recently, it has become an important ingredient in the work of the Kyoto school on the KP hierarchy of soliton equations. In the present paper we establish a super boson-fermion correspondence, having in mind its applications to super KP hierarchies

  5. Electrical insulating liquid: A review

    Directory of Open Access Journals (Sweden)

    Deba Kumar Mahanta

    2017-08-01

    Full Text Available Insulating liquid plays an important role for the life span of the transformer. Petroleum-based mineral oil has become dominant insulating liquid of transformer for more than a century for its excellent dielectric and cooling properties. However, the usage of petroleum-based mineral oil, derived from a nonrenewable energy source, has affected the environment for its nonbiodegradability property. Therefore, researchers direct their attention to renewable and biodegradable alternatives. Palm fatty acid ester, coconut oil, sunflower oil, etc. are considered as alternatives to replace mineral oil as transformer insulation liquid. This paper gives an extensive review of different liquid insulating materials used in a transformer. Characterization of different liquids as an insulating material has been discussed. An attempt has been made to classify different insulating liquids-based on different properties.

  6. 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.

  7. 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

  8. 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

  9. Super oil cracking update

    International Nuclear Information System (INIS)

    Mulraney, D.

    1997-01-01

    The conversion of residual fuel oil to usable middle distillates was discussed. The residue conversion processing paths are usually based on separation, carbon rejection, or hydrogen addition principles. Super Oil Cracking (SOC) uses a slurry catalyst system in a new, tubular reactor to achieve high levels of hydrothermal conversion. SOC can upgrade a variety of heavy, high metals residue feedstocks with high yields of middle distillates. The SOC products can also be further treated into feedstocks for FCC or hydrocracking. The SOC process can be incorporated easily into a refinery to obtain incremental residue conversion directly. It can also be integrated with other residue processes, acting as a demetallization and decarbonization step which results in enhanced overall conversion. The relative rate of coke formation and its handling are distinguishing characteristics between residue upgrading technologies. The SOC process operates at higher temperatures that other residue hydrocracking processes resulting in higher rates of thermal decomposition, thus preventing coke formation. SOC process can operate as a stand-alone upgrader or can be integrated with other bottoms processing steps to extend the refiner's range of options for increasing bottoms conversion.3 tabs., 14 figs

  10. Wideroe pre-accelerator for the SuperHILAC

    International Nuclear Information System (INIS)

    Staples, J.; Alonso, J.; Behrsing, G.; Clark, D.; Grunder, H.; Olivier, M.; Spence, D.; Yourd, R.

    1976-09-01

    In 1971 the Bevatron successfully accelerated low-intensity heavy ion beams up to neon to energies of 2.1 GeV/amu. More recently, beams up to argon have been accelerated using the SuperHILAC as an injector to the Bevatron--the Bevalac concept. With increasing scientific interest in high-energy high-intensity beams of heavier ions, plans to upgrade both the Bevatron vacuum system and the SuperHILAC ion sources and injectors have been formulated. A proposed new pre-accelerator based on an air-insulated Cockcroft-Walton and a Wideroe linac is presented. The Wideroe linac uses the design concepts established at UNILAC, modified for frequency and energy requirements. U 7 + from the ion source is accelerated from 12 keV/amu to 113 keV/amu and stripped to a mean charge state acceptable to the first tank of the SuperHILAC. The expected intensity improvement over the present pressurized injector is a factor of 100 at the highest masses. The physical modeling of the Wideroe linac structure will be kept to a minimum. Computer models predicting the characteristics of the structure have improved to the point where the probability of satisfactory performance is high

  11. 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.)

  12. Model SSC [Superconducting Super Collider] dipole magnet cryostat assembly at Fermilab

    International Nuclear Information System (INIS)

    Niemann, R.C.

    1989-03-01

    The Superconducting Super Collider (SSC) magnet development program includes the design, fabrication and testing of full length model dipole magnets. A result of the program has been the development of a magnet cryostat design. The cryostat subsystems consist of cold mass connection-slide, suspension, thermal shields, insulation, vacuum vessel and interconnections. Design details are presented along with model magnet production experience. 6 refs., 13 figs

  13. 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.)

  14. Computation Method Comparison for Th Based Seed-Blanket Cores

    International Nuclear Information System (INIS)

    Kolesnikov, S.; Galperin, A.; Shwageraus, E.

    2004-01-01

    This work compares two methods for calculating a given nuclear fuel cycle in the WASB configuration. Both methods use the ELCOS Code System (2-D transport code BOXER and 3-D nodal code SILWER) [4] are compared. In the first method, the cross-sections of the Seed and Blanket, needed for the 3-D nodal code are generated separately for each region by the 2-D transport code. In the second method, the cross-sections of the Seed and Blanket, needed for the 3-D nodal code are generated from Seed-Blanket Colorsets (Fig.1) calculated by the 2-D transport code. The evaluation of the error introduced by the first method is the main objective of the present study

  15. 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

  16. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Luke [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-30

    A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

  17. Heating an aquaculture pond with a solar pool blanket

    Energy Technology Data Exchange (ETDEWEB)

    Wisely, B; Holliday, J E; MacDonald, R E

    1982-01-01

    A floating solar blanket of laminated bubble plastic was used to heat a 0.11 ha seawater pond of 1.3 m depth. The covered pond maintained daily temperatures 6 to 9/sup 0/C above two controls. Local air temperatures averaged 14 to 19/sup 0/C. Oysters, prawns, seasquirts, and fish in the covered pond all survived. After three weeks, the blanket separated. This was the result of pond temperatures exceeding 30/sup 0/C, the maximum manufacturer's specification. Floating blankets fabricated to higher specifications would be useful for maintaining above-ambient temperatures in small ponds or tanks in temporary situations during cold winter months and might have a more permanent use.

  18. Solid breeder test blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ying, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)]. E-mail: ying@fusion.ucla.edu; Abdou, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Calderoni, P. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Sharafat, S. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Youssef, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); An, Z. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Abou-Sena, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Kim, E. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Reyes, S. [LANL, Livermore, CA (United States); Willms, S. [LANL, Los Alamos, NM (United States); Kurtz, R. [PNNL, Richland, WA (United States)

    2006-02-15

    This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration of fusion break-in phenomena and configuration scoping. Specific emphasis is placed on first wall structural response, evaluation of neutronic parameters, assessment of thermomechanical behavior and characterization of tritium release. The tests will be conducted with three unit cell arrays/sub-modules. The development approach includes: (1) design the unit cell/sub-module for low temperature operations and (2) refer to a reactor blanket design and use engineering scaling to reproduce key parameters under ITER wall loading conditions, so that phenomena under investigation can be measured at a reactor-like level.

  19. Direct LiT Electrolysis in a Metallic Fusion Blanket

    International Nuclear Information System (INIS)

    Olson, Luke

    2016-01-01

    A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

  20. The evolution of US helium-cooled blankets

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.

    1991-01-01

    This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors. (orig.)

  1. 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

  2. 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

  3. Cylindrical cryogenic calorimeter testing of six types of multilayer insulation systems

    Science.gov (United States)

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

    2018-01-01

    Extensive cryogenic thermal testing of more than 100 different multilayer insulation (MLI) specimens was performed over the last 20 years for the research and development of evacuated reflective thermal insulation systems. From this data library, 26 MLI systems plus several vacuum-only systems are selected for analysis and comparison. The test apparatus, methods, and results enabled the adoption of two new technical consensus standards under ASTM International. Materials tested include reflectors of aluminum foil or double-aluminized Mylar and spacers of fiberglass paper, polyester netting, silk netting, polyester fabric, or discrete polymer standoffs. The six types of MLI systems tested are listed as follows: Mylar/Paper, Foil/Paper, Mylar/Net, Mylar/Blanket, Mylar/Fabric, Mylar/Discrete. Also tested are vacuum-only systems with different cold surface materials/finishes including stainless steel, black, copper, and aluminum. Testing was performed between the boundary temperatures of 78 K and 293 K (and up to 350 K) using a thermally guarded one-meter-long cylindrical calorimeter (Cryostat-100) for absolute heat flow measurement. Cold vacuum pressures include the full range from 1 × 10-6 torr to 760 torr with nitrogen as the residual gas. System variations include number of layers from one to 80 layers, layer densities from 0.5 to 5 layers per millimeter, and installation techniques such layer-by-layer, blankets (multi-layer assemblies), sub-blankets, seaming, butt-joining, spiral wrapping, and roll-wrapping. Experimental thermal performance data for the different MLI systems are presented in terms of heat flux and effective thermal conductivity. Benchmark cryogenic-vacuum thermal performance curves for MLI are given for comparison with different insulation approaches for storage and transfer equipment, cryostats, launch vehicles, spacecraft, or science instruments.

  4. Liquid blanket MHD effects experimental results from LMEL facility at SWIP

    International Nuclear Information System (INIS)

    Xu Zengyu; Pan Chuanjie; Liu Yong; Pan Chuanhong; Reed, C.B.

    2007-01-01

    The self-cooled /helium-cooled liquid metal blanket concept is an attractive ITER and DEMO blanket candidate as it has low operating pressure, simplicity, and a convenient tritium breeding cycle. But MHD pressure drop remains a key issue, especially in ducts with flow channel inserts (FCI), where the reduction in MHD pressure drop is difficult to predict with existing tools, and there are no available experimental data to check current predictions. To understand well various kinds of MHD effects, it is important for us to analyze and understand FCI effects. In this paper, we present measurements of the MHD effects due to off normal power shutdown, two-dimensional effects due to channel velocity profiles, three-dimensional effects caused by manifolds, and surface/bulk instability effects as a result of insulator coating imperfections. These results were collected from the Liquid Metal Experimental Loop (LMEL) facility at Southwestern Institute of Physics, China and in collaboration with Argonne National Laboratory, US under an umbrella of the People's Republic of China/United States program of cooperation in magnetic fusion. Some results were observed for the first time, such as two dimensional effects and instabilities due to insulator coating imperfections. The experiments were conducted under the following conditions: a uniform magnetic field volume of 80 x 170 x 740 mm and a maximum value of magnetic field, B 0 , of 2 Tesla. The mean flow velocity v 0 was measured with an electromagnetic (EM) flow meter (error of 1.2%); a Liquid-metal Electro-magnetic Velocity Instrument (LEVI) was provided by Argonne National Laboratory. The flow was driven by two Electro-magnetic (EM) pumps (6.5+11.6 m3/h); the operating temperature was 85 centigrade degree due to self-heating by the EM pump and friction of the fluid against the loop piping. Experimental parameters were: Hartmann number, M, up to 3500, velocity v 0 up to 1.2 m/s under magnetic field, and B 0 =1.95 Tesla

  5. Magnetically insulated H- diodes

    International Nuclear Information System (INIS)

    Fisher, A.; Bystritskii, V.; Garate, E.; Prohaska, R.; Rostoker, N.

    1993-01-01

    At the Univ. of California, Irvine, the authors have been studying the production of intense H - beams using pulse power techniques for the past 7 years. Previously, current densities of H - ions for various diode designs at UCI have been a few A/cm 2 . Recently, they have developed diodes similar to the coaxial design of the Lebedev Physical Institute, Moscow, USSR, where current densities of up to 200 A/cm 2 were reported using nuclear activation of a carbon target. In experiments at UCI employing the coaxial diode, current densities of up to 35 A/cm 2 from a passive polyethylene cathode loaded with TiH 2 have been measured using a pinhole camera and CR-39 track recording plastic. The authors have also been working on a self-insulating, annular diode which can generate a directed beam of H - ions. In the annular diode experiments a plasma opening switch was used to provide a prepulse and a current path which self-insulated the diode. These experiments were done on the machine APEX, a 1 MV, 50 ns, 7 Ω pulseline with a unipolar negative prepulse of ∼ 100 kV and 400 ns duration. Currently, the authors are modifying the pulseline to include an external LC circuit which can generate a bipolar, 150 kV, 1 μs duration prepulse (similar prepulse characteristic as in the Lebedev Institute experiments cited above)

  6. Overview of the TFTR Lithium Blanket Module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

    The LBM (Lithium Blanket Module) is an approximately cubic module, about 80 cm on each side, with construction representative of a helium-cooled lithium oxide fusion reactor blanket module. Measurements of neutron transport and tritium breeding in the LBM will be made in irradiation programs first with a point-neutron source, and subsequently with the D-D and D-T fusion-neutron sources of the TFTR. This paper summarizes the objectives of the LBM program, the design, development and construction of the LBM, and progress in the experimental tests

  7. ITER blanket module shield block design and analysis

    International Nuclear Information System (INIS)

    Mitin, D.; Khomyakov, S.; Razmerov, A.; Strebkov, Yu.

    2008-01-01

    This paper presents the alternative design of the shield block cooling path for a typical ITER blanket module with a predominantly sequential flow circuit. A number of serious disadvantages have been observed for the reference design, where the parallel flow circuit is used, which is inherent in the majority of blanket modules. The paper discusses these disadvantages and demonstrates the benefit of the alternative design based on the detailed design and the technological, hydraulic, thermal, structural and strength analyses, conducted for module no. 17

  8. 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)

  9. 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

  10. 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)

  11. Limiter and first wall of the fusion reactor blanket

    International Nuclear Information System (INIS)

    Danilov, I.; Skladnov, K.; Kolganov, V.

    1994-01-01

    Previous designing of the first wall and limiter has allowed to determine their possible embodiment depending on the parameters and operation conditions of the blanket. As a rule limiter is a separate structure located on the plasma facing surface of the blanket assembly. Possible versions of the limiter/FW which may be considered: (1) limiters with mechanical attachment of the protective part; (2) limiters with the attachment with brazing; (3) limiters with common/separate cooling system; (4) limiter as a substitute of the FW. Generally the FW/limiter structure includes protective shield and its cooling system which consist of protective coating, heat accumulator, conductive layer and attachment locks

  12. Ceramic sphere-pac breeder design for fusion blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.J.; Sullivan, J.D.

    1991-01-01

    Randomly packed beds of ceramic spheres are a practical approach to surrounding fusion plasmas with tritium-breeding material. This paper examines the general properties of sphere-pac beds for application in fusion breeder blankets. The design considerations and models are reviewed for packing, tritium breeding and recovery, thermal conductivity, purge-gas pressure drop, mechanical behavior and fabrication. The design correlations are compared against available fusion ceramic data. Specific conclusions are that ternary (three-size) beds are not attractive for fusion blankets, and that the fusion spheres should be as large as possible subject primarily to packing constraints. (orig.)

  13. An aqueous lithium salt blanket option for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steiner, D.; Varsamis, G. (Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering and Engineering Physics); Deutsch, L.; Rathke, J. (Grumman Corp., Bethpage, NY (USA). Advanced Energy Systems); Gierszewski, P. (Canadian Fusion Fuels Technology Project (CFFTP), Mississauga, ON (Canada))

    1989-04-01

    An aqueous lithium salt blanket (ALSB) concept is proposed which could be the basis for either a power reactor blanket or a test module in an engineering test reactor. The design is based on an austenitic stainless steel structure, a beryllium multiplier, and a salt breeder concentration of about 32 g LiNO/sub 3/ per 100 cm/sup 3/ of H/sub 2/O. To limit tritium release rates, the salt breeder solution is separated from the water coolant circuit. The overall tritium system cost for a 2400 MW (fusion power) reactor is estimated to be 180 million Dollar US87 installed. (orig.).

  14. 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.)

  15. A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael

    2016-01-01

    'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

  16. Japanese contributions to the Japan-US workshop on blanket design/technology

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Seki, Yasushi; Minato, Akio; Kobayashi, Takeshi; Mori, Seiji; Kawasaki, Hiromitsu; Sumita, Kenji.

    1983-02-01

    This report describes Japanese papers presented at the Japan-US Workshop on Blanket Design/Technology which was held at Argonne National Laboratory, November 10 - 11, 1982. Overview of Fusion Experimental Reactor (FER), JAERI's activities related to first wall/blanket/shield, summary of FER blanket and its technology development issues and summary of activities at universities on fusion reactor blanket engineering are covered. (author)

  17. Insulation structure of thermonuclear device

    International Nuclear Information System (INIS)

    Suzuki, Takayuki; Usami, Saburo; Tsukamoto, Hideo; Kikuchi, Mitsuru

    1998-01-01

    The present invention provides an insulating structure of a thermonuclear device, in which insulation materials between toroidal coils are not broken even if superconductive toroidal coils are used. Namely, a tokamak type thermonuclear device of an insulating structure type comprises superconductive toroidal coils for confining plasmas arranged in a circular shape directing the center each at a predetermined angle, and the toroidal coils are insulated from each other. The insulation materials are formed by using a biaxially oriented fiber reinforced plastics. The contact surface of the toroidal coils and the insulating materials are arranged so that they are contact at a woven surface of the fiber reinforced plastics. Either or both of the contact surfaces of the fiber reinforced plastics and the toroidal coils are coated with a high molecular compound having a low friction coefficient. With such a constitution, since the interlayer shearing strength of the biaxially oriented fiber reinforced plastics is about 1/10 of the compression strength, the shearing stress exerted on the insulation material is reduced. Since a static friction coefficient on the contact surface is reduced to provide a structure causing slipping, shearing stress does not exceeds a predetermined limit. As a result, breakage of the insulation materials between the toroidal coils can be prevented. (I.S.)

  18. Plastic Materials for Insulating Applications.

    Science.gov (United States)

    Wang, S. F.; Grossman, S. J.

    1987-01-01

    Discusses the production and use of polymer materials as thermal insulators. Lists several materials that provide varying degrees of insulation. Describes the production of polymer foam and focuses on the major applications of polystyrene foam, polyurethane foam, and polyisocyanurate foam. (TW)

  19. Quantization of super Teichmueller spaces

    International Nuclear Information System (INIS)

    Aghaei, Nezhla

    2016-08-01

    The quantization of the Teichmueller spaces of Riemann surfaces has found important applications to conformal field theory and N=2 supersymmetric gauge theories. We construct a quantization of the Teichmueller spaces of super Riemann surfaces, using coordinates associated to the ideal triangulations of super Riemann surfaces. A new feature is the non-trivial dependence on the choice of a spin structure which can be encoded combinatorially in a certain refinement of the ideal triangulation. We construct a projective unitary representation of the groupoid of changes of refined ideal triangulations. Therefore, we demonstrate that the dependence of the resulting quantum theory on the choice of a triangulation is inessential. In the quantum Teichmueller theory, it was observed that the key object defining the Teichmueller theory has a close relation to the representation theory of the Borel half of U q (sl(2)). In our research we observed that the role of U q (sl(2)) is taken by quantum superalgebra U q (osp(1 vertical stroke 2)). A Borel half of U q (osp(1 vertical stroke 2)) is the super quantum plane. The canonical element of the Heisenberg double of the quantum super plane is evaluated in certain infinite dimensional representations on L 2 (R) x C 1 vertical stroke 1 and compared to the flip operator from the Teichmueller theory of super Riemann surfaces.

  20. Magnetically self-insulated transformers

    International Nuclear Information System (INIS)

    Novac, B.M.; Smith, I.R.; Brown, J.

    2002-01-01

    Magnetic insulation is the only practicable form of insulation for much equipment used in ultrahigh pulsed-power work, including transmission lines and plasma opening switches. It has not however so far been successfully exploited in the transformers that are necessarily involved, and the first proposed design that appeared more than 30 years ago raised apparently insuperable problems. The two novel arrangements for a magnetically insulated transformer described in this paper overcome the problems faced by the earlier designs and also offer considerable scope for development in a number of important areas. Theoretical justification is given for their insulating properties, and this is confirmed by proof-of-principle results obtained from a small-scale experimental prototype in which magnetic insulation was demonstrated at up to 100 kV. (author)

  1. Research on vacuum insulation for cryocables

    International Nuclear Information System (INIS)

    Graneau, P.

    1974-01-01

    Vacuum insulation, as compared with solid insulation, simplifies the construction of both resistive or superconducting cryogenic cables. The common vacuum space in the cable can furnish thermal insulation between the environment and the cryogenic coolant, provide electrical insulation between conductors, and establish thermal isolation between go- and return-coolant streams. The differences between solid and vacuum high voltage insulation are discussed, and research on the design, materials selection, and testing of vacuum insulated cryogenic cables is described

  2. Excitons in insulators

    International Nuclear Information System (INIS)

    Grasser, R.; Scharmann, A.

    1983-01-01

    This chapter investigates absorption, reflectivity, and intrinsic luminescence spectra of free and/or self-trapped (localized) excitons in alkali halides and rare gas solids. Introduces the concepts underlying the Wannier-Mott and Frenkel exciton models, two extreme pictures of an exciton in crystalline materials. Discusses the theoretical and experimental background; excitons in alkali halides; and excitons in rare gas solids. Shows that the intrinsic optical behavior of wide gap insulators in the range of the fundamental absorption edge is controlled by modified Wannier-Mott excitons. Finds that while that alkali halides only show free and relaxed molecular-like exciton emission, in rare gas crystals luminescence due to free, single and double centered localized excitons is observed. Indicates that the simultaneous existence of free and self-trapped excitons in these solid requires an energy barrier for self-trapping

  3. Mouse-resistant insulated covers keep pipes from freezing

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2010-01-15

    Fabric wellhead covers and insulated blankets are commonly used at well sites in the Wyoming coalbed methane field to keep surface pipes from freezing. These materials are often chewed up by mice who build nests close to the warm pipes. The mice attract rattlesnakes, a potentially serious problem for the workmen who check the wells daily. Kennon Products of Sheridan, Wyoming solved this problem by making a flexible covering material that has a coating of hardened guard plates that prevents mice from chewing through it. More than a hundred of Kennon's mouse-resistant wellhead covers have been used successfully in the gas fields for over a year. They can be installed in less than 30 minutes and cost only a fraction of what a fiberglass hut costs to purchase and install. Huts are being discouraged for use on federal lands because they alter the nesting patterns of eagles, who perch upon them to hunt rodents. Huts also trap methane gas, which is a potential safety hazard. Kennon's mouse-resistant wellhead covers are lower than the fiberglass huts and blend into the landscape. The company is working on camouflage colours to make wellheads less noticeable. In the future, the company plans to insulate water pipes. 1 fig.

  4. Siting the superconducting super collider

    International Nuclear Information System (INIS)

    Price, R.; Rooney, R.C.

    1988-01-01

    At the request of the Department of Energy, the National Academy of Sciences and the National Academy of Engineering established the Super Collider Site Evaluation Committee to evaluate the suitability of proposed sites for the Superconducting Super Collider. Thirty-six proposals were examined by the committee. Using the set of criteria announced by DOE in its Invitation for Site Proposals, the committee identified eight sites that merited inclusion on a ''best qualified list.'' The list represents the best collective judgment of 21 individuals, carefully chosen for their expertise and impartiality, after a detailed assessment of the proposals using 19 technical subcriteria and DOE's life cycle cost estimates. The sites, in alphabetical order, are: Arizona/Maricopa; Colorado; Illinois; Michigan/Stockbridge; New York/Rochester; North Carolina; Tennessee; and Texas/Dallas-Fort Worth. The evaluation of these sites and the Superconducting Super Collider are discussed in this book

  5. 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

  6. Summary of the target-blanket breakout group

    Energy Technology Data Exchange (ETDEWEB)

    Capiello, M.; Bell, C. [Los Alamos National Laboratory, NM (United States); Barthold, W.

    1995-10-01

    This breakout group discussed a number of topics and issues pertaining to target and blanket concepts for accelerator-driven systems. This major component area is one marked by a broad spectrum of technical approaches. It is therefore less defined than other major component areas such as the accelerator and is at an earlier stage of technical needs and task specification. The working group did reach a number of general conclusions and recommendations that are summarized. The Conference and the Target/Blanket Breakout Group provided a first opportunity for people working on a variety of missions and concepts to get together and exchange information. A number of subcritical systems applicable for a spectrum of missions were proposed at the Conference and discussed in the Breakout Group. Missions included plutonium disposition, energy production, waste destruction, isotope production, and neutron scattering. The Target/Blanket Breakout Group also defined areas where parameters and data should be addressed as target/blanket design activities become more detailed and sophisticated.

  7. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

    1985-01-01

    Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented

  8. First-wall/blanket materials selection for STARFIRE tokamak reactor

    International Nuclear Information System (INIS)

    Smith, D.L.; Mattas, R.F.; Clemmer, R.G.; Davis, J.W.

    1980-01-01

    The development of the reference STARFIRE first-wall/blanket design involved numerous trade-offs in the materials selection process for the breeding material, coolant structure, neutron multiplier, and reflector. The major parameters and properties that impact materials selection and design criteria are reviewed

  9. Thermal-hydraulic analysis of low activity fusion blanket designs

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.; Yu, W.S.

    1977-01-01

    The heat transfer aspects of fusion blankets are considered where: (a) conduction and (b) boiling and condensation are the dominant heat transfer mechanisms. In some cases, unique heat transfer problems arise and additional heat transfer data and analyses may be required

  10. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    Kleefeldt, K.

    1985-01-01

    The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m 2 . Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

  11. 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)

  12. 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.

  13. 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

  14. Tritium inventory in Li2ZrO3 blanket

    International Nuclear Information System (INIS)

    Nishikawa, M.; Baba, A.

    1998-01-01

    Recently, we have presented the way to estimate the tritium inventory in a solid breeder blanket considering effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reactions. It is reported in our previous paper that the estimated tritium inventory for a LiAlO 2 blanket agrees well with data observed in various in situ experiments when the effective diffusivity of tritium from the EXOTIC-6 experiment is used and that the better agreement is obtained when existence of some water vapor is assumed in the purge gas. The same way as used for a LiAlO 2 blanket is applied to a Li 2 ZrO 3 blanket in this study and the estimated tritium inventory shows a good agreement with data obtained in such in situ experiments as MOZART, EXOTIC-6 and TRINE experiments. (orig.)

  15. 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

  16. Quantisation of super Teichmueller theory

    International Nuclear Information System (INIS)

    Aghaei, Nezhla; Hamburg Univ.; Pawelkiewicz, Michal; Techner, Joerg

    2015-12-01

    We construct a quantisation of the Teichmueller spaces of super Riemann surfaces using coordinates associated to ideal triangulations of super Riemann surfaces. A new feature is the non-trivial dependence on the choice of a spin structure which can be encoded combinatorially in a certain refinement of the ideal triangulation. By constructing a projective unitary representation of the groupoid of changes of refined ideal triangulations we demonstrate that the dependence of the resulting quantum theory on the choice of a triangulation is inessential.

  17. A note on the super AKNS equations

    International Nuclear Information System (INIS)

    Li Yishen; Zhang Lining.

    1986-10-01

    We find some relationships between the usual AKNS scheme with the super one, when its elements take value from the Grassmann algebra on a two-dimensional vector space. The solutions of these super AKNS equations are discussed. (author)

  18. 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

  19. Remote handling assessment of attachment concepts for DEMO blanket segments

    Energy Technology Data Exchange (ETDEWEB)

    Iglesias, Daniel, E-mail: daniel.iglesias@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Bastow, Roger; Cooper, Dave; Crowe, Robert; Middleton-Gear, Dave [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Sibois, Romain [VTT, Technical Research Centre of Finland, Industrial Systems, ROViR, Tampere (Finland); Carloni, Dario [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT) (Germany); Vizvary, Zsolt; Crofts, Oliver [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Harman, Jon [EFDA Close Support Unit Garching, Boltzmannstaße 2, D-85748 Garching bei München (Germany); Loving, Antony [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2015-10-15

    Highlights: • Challenges are identified for the remote handling of blanket segments’ attachments. • Two attachment design approaches are assessed for remote handling (RH) feasibility. • An alternative is proposed, which potentially simplifies and speeds-up RH operations. • Up to three different assemblies are proposed for the remote handling of the attachments. • Proposed integrated design of upper port is compatible with the attachment systems. - Abstract: The replacement strategy of the massive Multi-Module Blanket Segments (MMS) is a key driver in the design of several DEMO systems. These include the blankets themselves, the vacuum vessel (VV) and its ports and the Remote Maintenance System (RMS). Common challenges to any blanket attachment system have been identified, such as the need for applying a preload to the MMS manifold, the effects of the decay heat and several uncertainties related to permanent deformations when removing the blanket segments after service. The WP12 kinematics of the MMS in-vessel transportation was adapted to the requirements of each of the supports during 2013 and 2014 design activities. The RM equipment envisaged for handling attachments and earth connections may be composed of up to three different assemblies. An In-Vessel Mover at the divertor level handles the lower support and earth bonding, and could stabilize the MMS during transportation. A Shield Plug crane with a 6 DoF manipulator operates the upper attachment and earth straps. And a Vertical Maintenance Crane is responsible for the in-vessel MMS transportation and can handle the removable upper support pins. A final proposal is presented which can potentially reduce the number of required systems, at the same time that speeds-up the RMS global operations.

  20. 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

  1. Evaluation of compost blankets for erosion control from disturbed lands.

    Science.gov (United States)

    Bhattarai, Rabin; Kalita, Prasanta K; Yatsu, Shotaro; Howard, Heidi R; Svendsen, Niels G

    2011-03-01

    Soil erosion due to water and wind results in the loss of valuable top soil and causes land degradation and environmental quality problems. Site specific best management practices (BMP) are needed to curb erosion and sediment control and in turn, increase productivity of lands and sustain environmental quality. The aim of this study was to investigate the effectiveness of three different types of biodegradable erosion control blankets- fine compost, mulch, and 50-50 mixture of compost and mulch, for soil erosion control under field and laboratory-scale experiments. Quantitative analysis was conducted by comparing the sediment load in the runoff collected from sloped and tilled plots in the field and in the laboratory with the erosion control blankets. The field plots had an average slope of 3.5% and experiments were conducted under natural rainfall conditions, while the laboratory experiments were conducted at 4, 8 and 16% slopes under simulated rainfall conditions. Results obtained from the field experiments indicated that the 50-50 mixture of compost and mulch provides the best erosion control measures as compared to using either the compost or the mulch blanket alone. Laboratory results under simulated rains indicated that both mulch cover and the 50-50 mixture of mulch and compost cover provided better erosion control measures compared to using the compost alone. Although these results indicate that the 50-50 mixtures and the mulch in laboratory experiments are the best measures among the three erosion control blankets, all three types of blankets provide very effective erosion control measures from bare-soil surface. Results of this study can be used in controlling erosion and sediment from disturbed lands with compost mulch application. Testing different mixture ratios and types of mulch and composts, and their efficiencies in retaining various soil nutrients may provide more quantitative data for developing erosion control plans. Copyright © 2010 Elsevier

  2. Structural analysis under the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Majumdar, S.

    1985-01-01

    Structural design procedures followed in the Blanket Comparison and Selection Study are briefly reviewed. The American Society of Mechanical Engineers Boilers and Pressure Vessels Code, Section III, Code Case N47 has been used as a design guide. Its relevance to fusion reactor applications, however, is open to question and needs to be evaluated in the future. The primary structural problem encountered in tokamak blanket designs is the high thermal stress due to surface heat flux, with fatigue being an additional concern for pulsed systems. The conflicting requirements of long erosion life and high surface heat flux capability imply that some form of stress relief in the first-wall region will be necessary. Simplified stress and fatigue crack growth analyses are presented to show that the use of orthogonally grooved first wall may be a potential solution for mitigating the thermal stress problem. A comparison of three structural alloys on the basis of both grooved and nongrooved first-wall designs is also presented. Other structural problems encountered in tokamak designs include stresses due to plasma disruptions, and magnetohydrodynamic (MHD) pressure drop in liquid-metal-cooled systems. In particular, it is shown that the maximum stress in the side wall of a uniform duct generated by MHD pressure drop cannot be reduced by increasing the wall thickness or by decreasing the span. In contract to tokamak blankets, tandem mirror blankets are far less severely stressed because of a much lower surface heat flux, coolant pressure, and also because of their axisymmetric geometry. Both blankets, however, will require detailed structural dynamics analysis to verify their ability to withstand seismic loadings if the heavy 17Li-83Pb is used as a coolant

  3. 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)

  4. The super-classical-Boussinesq hierarchy and its super-Hamiltonian structure

    International Nuclear Information System (INIS)

    Si-Xing, Tao; Tie-Cheng, Xia

    2010-01-01

    Based on the constructed Lie superalgebra, the super-classical-Boussinesq hierarchy is obtained. Then, its super-Hamiltonian structure is obtained by making use of super-trace identity. Furthermore, the super-classical-Boussinesq hierarchy is also integrable in the sense of Liouville. (general)

  5. Lightweight, Thermally Insulating Structural Panels

    Science.gov (United States)

    Eisen, Howard J.; Hickey, Gregory; Wen, Liang-Chi; Layman, William E.; Rainen, Richard A.; Birur, Gajanana C.

    1996-01-01

    Lightweight, thermally insulating panels that also serve as structural members developed. Honeycomb-core panel filled with low-thermal-conductivity, opacified silica aerogel preventing convection and minimizes internal radiation. Copper coating on face sheets reduces radiation. Overall thermal conductivities of panels smaller than state-of-art commercial non-structurally-supporting foam and fibrous insulations. On Earth, panels suitable for use in low-air-pressure environments in which lightweight, compact, structurally supporting insulation needed; for example, aboard high-altitude aircraft or in partially evacuated panels in refrigerators.

  6. Reflecting variable opening insulating panel

    International Nuclear Information System (INIS)

    Nungesser, W.T.

    1976-01-01

    A description is given of a reflecting variable opening insulating panel assembly, comprising a static panel assembly of reflecting insulation sheets forming a cavity along one side of the panel and a movable panel opening out by sliding from the cavity of the static panel, and a locking device for holding the movable panel in a position extending from the cavity of the static panel. This can apply to a nuclear reactor of which the base might require maintenance and periodical checking and for which it is desirable to have available certain processes for the partial dismantling of the insulation [fr

  7. Research and development of super light water reactors and super fast reactors in Japan

    International Nuclear Information System (INIS)

    Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

    2011-01-01

    Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

  8. Supergrassmannians, super τ-functions and strings

    International Nuclear Information System (INIS)

    Dolgikh, S.N.; Schwarz, A.S.

    1989-03-01

    Recently, infinite-dimensional grassmannians and their supergeneralizations were used to study conformal two-dimensional fields and strings. In particular, the super Mumford form (holomorphic square root from the superstring measure on moduli space) was expressed through super analog of Sato τ-function. In this paper we present results of supergrassmannians and super τ-functions. 8 refs

  9. Study on the temperature control mechanism of the tritium breeding blanket for CFETR

    Science.gov (United States)

    Liu, Changle; Qiu, Yang; Zhang, Jie; Zhang, Jianzhong; Li, Lei; Yao, Damao; Li, Guoqiang; Gao, Xiang; Wu, Songtao; Wan, Yuanxi

    2017-12-01

    The Chinese fusion engineering testing reactor (CFETR) will demonstrate tritium self- sufficiency using a tritium breeding blanket for the tritium fuel cycle. The temperature control mechanism (TCM) involves the tritium production of the breeding blanket and has an impact on tritium self-sufficiency. In this letter, the CFETR tritium target is addressed according to its missions. TCM research on the neutronics and thermal hydraulics issues for the CFETR blanket is presented. The key concerns regarding the blanket design for tritium production under temperature field control are depicted. A systematic theory on the TCM is established based on a multiplier blanket model. In particular, a closed-loop method is developed for the mechanism with universal function solutions, which is employed in the CFETR blanket design activity for tritium production. A tritium accumulation phenomenon is found close to the coolant in the blanket interior, which has a very important impact on current blanket concepts using water coolant inside the blanket. In addition, an optimal tritium breeding ratio (TBR) method based on the TCM is proposed, combined with thermal hydraulics and finite element technology. Meanwhile, the energy gain factor is adopted to estimate neutron heat deposition, which is a key parameter relating to the blanket TBR calculations, considering the structural factors. This work will benefit breeding blanket engineering for the CFETR reactor in the future.

  10. Super-resolution Phase Tomography

    KAUST Repository

    Depeursinge, Christian; Cotte, Yann; Toy, Fatih; Jourdain, Pascal; Boss, Daiel; Marquet, Pierre; Magistretti, Pierre J.

    2013-01-01

    Digital Holographic Microscopy (DHM) yields reconstructed complex wavefields. It allows synthesizing the aperture of a virtual microscope up to 2π, offering super-resolution phase images. Live images of micro-organisms and neurons with resolution less than 100 nm are presented.

  11. Super-resolution Phase Tomography

    KAUST Repository

    Depeursinge, Christian

    2013-04-21

    Digital Holographic Microscopy (DHM) yields reconstructed complex wavefields. It allows synthesizing the aperture of a virtual microscope up to 2π, offering super-resolution phase images. Live images of micro-organisms and neurons with resolution less than 100 nm are presented.

  12. Super Virasoro algebra and solvable supersymmetric quantum field theories

    International Nuclear Information System (INIS)

    Yamanaka, Itaru; Sasaki, Ryu.

    1987-09-01

    Interesting and deep relationships between super Virasoro algebras and super soliton systems (super KdV, super mKdV and super sine-Gordon equations) are investigated at both classical and quantum levels. An infinite set of conserved quantities responsible for solvability is characterized by super Virasoro algebras only. Several members of the infinite set of conserved quantities are derived explicitly. (author)

  13. 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)

  14. A New Generation of Building Insulation by Foaming Polymer Blend Materials with CO2

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Arthur [Industrial Science & Technology Network, Inc., Lancaster, PA (United States); Domszy, Roman [Industrial Science & Technology Network, Inc., Lancaster, PA (United States); Yang, Jeff [Industrial Science & Technology Network, Inc., Lancaster, PA (United States)

    2016-03-30

    Advanced thermal insulation is among the most effective technologies in transforming our nation’s energy system and contributing to DOE’s stated goal of 50% less building energy consumption by 2030. The installation of an advanced thermal insulation would prevent energy waste without the need for any maintenance, and ISTN conservatively estimates that the commercialization of such a new technology would contribute to annual U.S. energy savings of 0.361 Quads and $8 billion in annual economic savings. The key challenge to improving building insulation is to maintain and surpass the industry standard of R-5 per inch insulation value in a cost-competitive manner. Improvements in R-value without cost-efficiency are not likely to impact the market given the cost-sensitive nature of the construction industry (insulation is already the lowest-cost component of the building envelope). However, significantly higher insulating value at competitive costs is extremely appealing to the market given the greater potential to save on energy consumption and costs over the long-term. Thus, our goal is to develop a super-thermal insulation with 50% greater insulation value (R-9 to R-10 per inch) and manufacturing costs that are equal on a per-R-value basis (< $0.70/ft2).

  15. Molecular dewetting on insulators

    International Nuclear Information System (INIS)

    Burke, S A; Topple, J M; Gruetter, P

    2009-01-01

    Recent attention given to the growth and morphology of organic thin films with regard to organic electronics has led to the observation of dewetting (a transition from layer(s) to islands) of molecular deposits in many of these systems. Dewetting is a much studied phenomenon in the formation of polymer and liquid films, but its observation in thin films of the 'small' molecules typical of organic electronics requires additional consideration of the structure of the interface between the molecular film and the substrate. This review covers some key concepts related to dewetting and molecular film growth. In particular, the origins of different growth modes and the thickness dependent interactions which give rise to dewetting are discussed in terms of surface energies and the disjoining pressure. Characteristics of molecular systems which may lead to these conditions, including the formation of metastable interface structures and commensurate-incommensurate phase transitions, are also discussed. Brief descriptions of some experimental techniques which have been used to study molecular dewetting are given as well. Examples of molecule-on-insulator systems which undergo dewetting are described in some detail, specifically perylene derivatives on alkali halides, C 60 on alkali halides, and the technologically important system of pentacene on SiO 2 . These examples point to some possible predicting factors for the occurrence of dewetting, most importantly the formation of an interface layer which differs from the bulk crystal structure. (topical review)

  16. Molecular dewetting on insulators.

    Science.gov (United States)

    Burke, S A; Topple, J M; Grütter, P

    2009-10-21

    Recent attention given to the growth and morphology of organic thin films with regard to organic electronics has led to the observation of dewetting (a transition from layer(s) to islands) of molecular deposits in many of these systems. Dewetting is a much studied phenomenon in the formation of polymer and liquid films, but its observation in thin films of the 'small' molecules typical of organic electronics requires additional consideration of the structure of the interface between the molecular film and the substrate. This review covers some key concepts related to dewetting and molecular film growth. In particular, the origins of different growth modes and the thickness dependent interactions which give rise to dewetting are discussed in terms of surface energies and the disjoining pressure. Characteristics of molecular systems which may lead to these conditions, including the formation of metastable interface structures and commensurate-incommensurate phase transitions, are also discussed. Brief descriptions of some experimental techniques which have been used to study molecular dewetting are given as well. Examples of molecule-on-insulator systems which undergo dewetting are described in some detail, specifically perylene derivatives on alkali halides, C(60) on alkali halides, and the technologically important system of pentacene on SiO(2). These examples point to some possible predicting factors for the occurrence of dewetting, most importantly the formation of an interface layer which differs from the bulk crystal structure.

  17. Insulating fcc YH

    International Nuclear Information System (INIS)

    Molen, S. J. van der; Nagengast, D. G.; Gogh, A. T. M. van; Kalkman, J.; Kooij, E. S.; Rector, J. H.; Griessen, R.

    2001-01-01

    We study the structural, optical, and electrical properties of Mg z Y 1-z switchable mirrors upon hydrogenation. It is found that the alloys disproportionate into essentially pure YH 3-δ and MgH 2 with the crystal structure of YH 3-δ dependent on the Mg concentration z. For 0 3-δ are observed, whereas for z≥0.1 only cubic YH 3-δ is present. Interestingly, cubic YH 3-δ is expanded compared to YH 2 , in disagreement with theoretical predictions. From optical and electrical measurements we conclude that cubic YH 3-δ is a transparent insulator with properties similar to hexagonal YH 3-δ . Our results are inconsistent with calculations predicting fcc YH 3-δ to be metallic, but they are in good agreement with recent GW calculations on both hcp and fcc YH 3 . Finally, we find an increase in the effective band gap of the hydrided Mg z Y 1-z alloys with increasing z. Possibly this is due to quantum confinement effects in the small YH 3 clusters

  18. Topological Insulator Nanowires and Nanoribbons

    KAUST Repository

    Kong, Desheng; Randel, Jason C.; Peng, Hailin; Cha, Judy J.; Meister, Stefan; Lai, Keji; Chen, Yulin; Shen, Zhi-Xun; Manoharan, Hari C.; Cui, Yi

    2010-01-01

    Recent theoretical calculations and photoemission spectroscopy measurements on the bulk Bi2Se3 material show that it is a three-dimensional topological insulator possessing conductive surface states with nondegenerate spins, attractive

  19. Measure Guideline: Basement Insulation Basics

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, R.; Mantha, P.; Puttagunta, S.

    2012-10-01

    This guideline is intended to describe good practices for insulating basements in new and existing homes, and is intended to be a practical resources for building contractors, designers, and also to homeowners.

  20. Metal-insulator-semiconductor photodetectors.

    Science.gov (United States)

    Lin, Chu-Hsuan; Liu, Chee Wee

    2010-01-01

    The major radiation of the sun can be roughly divided into three regions: ultraviolet, visible, and infrared light. Detection in these three regions is important to human beings. The metal-insulator-semiconductor photodetector, with a simpler process than the pn-junction photodetector and a lower dark current than the MSM photodetector, has been developed for light detection in these three regions. Ideal UV photodetectors with high UV-to-visible rejection ratio could be demonstrated with III-V metal-insulator-semiconductor UV photodetectors. The visible-light detection and near-infrared optical communications have been implemented with Si and Ge metal-insulator-semiconductor photodetectors. For mid- and long-wavelength infrared detection, metal-insulator-semiconductor SiGe/Si quantum dot infrared photodetectors have been developed, and the detection spectrum covers atmospheric transmission windows.

  1. Metal-Insulator-Semiconductor Photodetectors

    Directory of Open Access Journals (Sweden)

    Chu-Hsuan Lin

    2010-09-01

    Full Text Available The major radiation of the Sun can be roughly divided into three regions: ultraviolet, visible, and infrared light. Detection in these three regions is important to human beings. The metal-insulator-semiconductor photodetector, with a simpler process than the pn-junction photodetector and a lower dark current than the MSM photodetector, has been developed for light detection in these three regions. Ideal UV photodetectors with high UV-to-visible rejection ratio could be demonstrated with III-V metal-insulator-semiconductor UV photodetectors. The visible-light detection and near-infrared optical communications have been implemented with Si and Ge metal-insulator-semiconductor photodetectors. For mid- and long-wavelength infrared detection, metal-insulator-semiconductor SiGe/Si quantum dot infrared photodetectors have been developed, and the detection spectrum covers atmospheric transmission windows.

  2. Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.; Embrechts, M.J.

    1987-11-01

    The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs

  3. 16 CFR 460.18 - Insulation ads.

    Science.gov (United States)

    2010-01-01

    ... Commercial Practices FEDERAL TRADE COMMISSION TRADE REGULATION RULES LABELING AND ADVERTISING OF HOME INSULATION § 460.18 Insulation ads. (a) If your ad gives an R-value, you must give the type of insulation and... your ad gives a price, you must give the type of insulation, the R-value at a specific thickness, the...

  4. Cryogenic foam insulation: Abstracted publications

    Science.gov (United States)

    Williamson, F. R.

    1977-01-01

    A group of documents were chosen and abstracted which contain information on the properties of foam materials and on the use of foams as thermal insulation at cryogenic temperatures. The properties include thermal properties, mechanical properties, and compatibility properties with oxygen and other cryogenic fluids. Uses of foams include applications as thermal insulation for spacecraft propellant tanks, and for liquefied natural gas storage tanks and pipelines.

  5. Superconducting super collider second generation dipole magnet cryostat design

    International Nuclear Information System (INIS)

    Niemann, R.C.; Bossert, R.C.; Carson, J.A.; Engler, N.H.; Gonczy, J.D.; Larson, E.T.; Nicol, T.H.; Ohmori, T.

    1988-12-01

    The Superconducting Super Collider, a planned colliding beam particle physics research facility, requires /approximately/10,000 superconducting devices for the control of high energy particle beams. The /approximately/7,500 collider ring superconducting dipole magnets require cryostats that are functional, cryogenically efficient, mass producible and cost effective. A second generation cryostat design has been developed utilizing the experiences gained during the construction, installation and operation of several full length first generation dipole magnet models. The nature of the cryostat improvements is presented. Considered are the connections between the magnet cold mass and its supports, cryogenic supports, cold mass axial anchor, thermal shields, insulation, vacuum vessel and interconnections. The details of the improvements are enumerated and the abstracted results of available component and system evaluations are presented. 8 refs., 11 figs

  6. Novel load responsive multilayer insulation with high in-atmosphere and on-orbit thermal performance

    Science.gov (United States)

    Dye, S.; Kopelove, A.; Mills, G. L.

    2012-04-01

    Aerospace cryogenic systems require lightweight, high performance thermal insulation to preserve cryopropellants both pre-launch and on-orbit. Current technologies have difficulty meeting all requirements, and advances in insulation would benefit cryogenic upper stage launch vehicles, LH2 fueled aircraft and ground vehicles, and provide capabilities for sub-cooled cryogens for space-borne instruments and orbital fuel depots. This paper reports the further development of load responsive multilayer insulation (LRMLI) that has a lightweight integrated vacuum shell and provides high thermal performance both in-air and on-orbit. LRMLI is being developed by Quest Product Development and Ball Aerospace under NASA contract, with prototypes designed, built, installed and successfully tested. A 3-layer LRMLI blanket (0.63 cm thick, 77 K cold, 295 K hot) had a measured heat leak of 6.6 W/m2 in vacuum and 40.6 W/m2 in air at one atmosphere. In-air LRMLI has an 18× advantage over Spray On Foam Insulation (SOFI) in heat leak per thickness and a 16× advantage over aerogel. On-orbit LRMLI has a 78× lower heat leak than SOFI per thickness and 6× lower heat leak than aerogel. The Phase II development of LRMLI is reported with a modular, flexible, thin vacuum shell and improved on-orbit performance. Structural and thermal analysis and testing results are presented. LRMLI mass and thermal performance is compared to SOFI, aerogel and MLI over SOFI.

  7. Electrically insulating films deposited on V-4%Cr-4%Ti by reactive CVD

    International Nuclear Information System (INIS)

    Park, J.H.

    1998-04-01

    In the design of liquid-metal blankets for magnetic fusion reactors, corrosion resistance of structural materials and the magnetohydrodynamic forces and their influence on thermal hydraulics and corrosion are major concerns. Electrically insulating CaO films deposited on V-4%Cr-4%Ti exhibit high-ohmic insulator behavior even though a small amount of vanadium from the alloy become incorporated into the film. However, when vanadium concentration in the film is > 15 wt.%, the film becomes conductive. When the vanadium concentration is high in localized areas, a calcium vanadate phase that exhibits semiconductor behavior can form. The objective of this study is to evaluate electrically insulating films that were deposited on V-4%Cr-4%Ti by a reactive chemical vapor deposition (CVD) method. To this end, CaO and Ca-V-O coatings were produced on vanadium alloys by CVD and by a metallic-vapor process to investigate the electrical resistance of the coatings. The authors found that the Ca-V-O films exhibited insulator behavior when the ratio of calcium concentration to vanadium concentration R in the film > 0.9, and semiconductor or conductor behavior when R 0.98 were exposed in liquid lithium. Based on these studies, they conclude that semiconductor behavior occurs if a conductive calcium vanadate phase is present in localized regions in the CaO coating

  8. Avalanches in insulating gases

    International Nuclear Information System (INIS)

    Verhaart, H.F.A.

    1982-01-01

    Avalanches of charged particles in gases are often studied with the ''electrical method'', the measurement of the waveform of the current in the external circuit. In this thesis a substantial improvement of the time resolution of the measuring setup, to be used for the electrical method, is reported. The avalanche is started by an N 2 -laser with a pulse duration of only 0.6 ns. With this laser it is possible to release a high number of primary electrons (some 10 8 ) which makes it possible to obtain sizeable signals, even at low E/p values. With the setup it is possible to analyze current waveforms with a time resolution down to 1.4 ns, determined by both the laser and the measuring system. Furthermore it is possible to distinguish between the current caused by the electrons and the current caused by the ions in the avalanche and to monitor these currents simultaneously. Avalanche currents are measured in N 2 , CO 2 , O 2 , H 2 O, air of varying humidity, SF 6 and SF 6 /N 2 mixtures. Depending on the nature of the gas and the experimental conditions, processes as diffusion, ionization, attachment, detachment, conversion and secondary emission are observed. Values of parameters with which these processes can be described, are derived from an analysis of the current waveforms. For this analysis already published theories and new theories described in this thesis are used. The drift velocity of both the electrons and the ions could be easily determined from measured avalanche currents. Special attention is paid to avalanches in air becasue of the practical importance of air insulation. (Auth.)

  9. APT 3He target/blanket. Topical report

    International Nuclear Information System (INIS)

    1995-03-01

    The 3 He target/blanket (T/B) preconceptual design for the 3/8-Goal facility is based on a 1000-MeV, 200-mA accelerator to produce a high-intensity proton beam that is expanded and then strikes one of two T/B modules. Each module consists of a centralized neutron source made of tungsten and lead, a proton beam backstop region made of zirconium and lead, and a moderator made of D 2 O. Helium-3 gas is circulated through the neutron source region and the blanket to create tritium through neutron capture. The gas is continually processed to extract the tritium with an online separation process

  10. Improved modules for the blanket of RTO/RC ITER

    International Nuclear Information System (INIS)

    Elio, F.; Ioki, K.; Cardella, A.

    2000-01-01

    This paper describes innovative design aspects that are considered to optimise the blanket modules for the reduced technical objective/reduced cost international thermonuclear experimental reactor. The blanket modules have a vertical straight profile facing the plasma, and the first wall is built in small and flat panels. Copper may be applied only in front of the first row of cooling passages. The radial cooling of the shield block avoids a complex by-pass at the back and opens up the possibility to use cast instead of forged steel. Slits in the shield block and in the first wall reduce the electromagnetic forces enough to allow the support of the modules on the vessel and the mechanical attachment of the first wall panels

  11. 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

  12. Blanket handling concepts for future fusion power plants

    International Nuclear Information System (INIS)

    Bogusch, E.; Gottfried, R.; Maisonnier, D.

    2003-01-01

    In the frame of the power plant conceptual studies (PPCS) launched by the European Commission, two main blanket handling concepts have been investigated with respect to engineering feasibility and the impact on the plant availability and on cost: the large module handling concept (LMHC) and the large sector handling concept (LSHC). The LMHC has been considered as the reference handling concept while the LSHC has been considered as an attractive alternative to the LMHC due to its potential of smaller replacement times and hence increasing the plant availability. Although no principle feasibility issue has been identified, a number of engineering issues have been highlighted for the LSHC that would require considerable efforts for their resolution. Since its availability of about 77% based on a replacement time for all the internals of about 4.2 months is slightly lower than for the LMHC, the LMHC remains the reference blanket replacement concept for a conceptual reactor

  13. 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.

  14. 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

  15. Stability properties of cold blanket systems for current driven modes

    International Nuclear Information System (INIS)

    Ohlsson, D.

    1977-12-01

    The stability problem of the boundary regions of cold blanket systems with induced currents parallel to the lines of force is formulated. Particular interest is focused on two types of modes: first electrostatic modes driven by the combined effects of a transverse resistivity gradient due to a spatially non-uniform electron temperature and a longitudinal current, second electromagnetic kink like modes driven by the torque arising from a transverse current density gradient and magnetic field perturbations. It is found that the combination of various dissipative and neutral gas effects introduces strong stabilizing effects within specific parameter ranges. For particular steady-state models investigated it is shown that these effects become of importance in laboratory plasmas at relatively high densities, low temperatures and moderate magnetic field strengths. Stability diagrams based on specific steady-state cold plasma blanket models will be presented

  16. APT {sup 3}He target/blanket. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    The {sup 3}He target/blanket (T/B) preconceptual design for the 3/8-Goal facility is based on a 1000-MeV, 200-mA accelerator to produce a high-intensity proton beam that is expanded and then strikes one of two T/B modules. Each module consists of a centralized neutron source made of tungsten and lead, a proton beam backstop region made of zirconium and lead, and a moderator made of D{sub 2}O. Helium-3 gas is circulated through the neutron source region and the blanket to create tritium through neutron capture. The gas is continually processed to extract the tritium with an online separation process.

  17. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    Jassby, D.L.; Leinoff, S.

    1979-12-01

    A conceptual design has been carried out for an Engineering Test Station (ETS) which will provide structural support and utilities/instrumentation services for blanket modules positioned adjacent to the vacuum vessel of the TFTR (Tokamak Fusion Test Reactor). The ETS is supported independently from the Test Cell floor. The ETS module support platform is constructed of fiberglass to eliminate electromagnetic interaction with the pulsed tokamak fields. The ETS can hold blanket modules with dimensions up to 78 cm in width, 85 cm in height, and 105 cm in depth, and with a weight up to 4000 kg. Interfaces for all utility and instrumentation requirements are made via a shield plug in the TFTR igloo shielding. The modules are readily installed or removed by means of TFTR remote handling equipment

  18. 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

  19. 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.

  20. Tritium breeding blanket device of D-T reactors

    International Nuclear Information System (INIS)

    Chevereau, G.

    1984-01-01

    This blanket device uses solid tritium breeding materials as those which include, in a known manner, near a neutron breeding plasma, a neutron multiplier medium and a tritium breeding medium, cooled by a cooling fluid circulation. This device is characterized by the fact that the association of the multiplier media and the tritium breeding media is realized by pellet alternated piling up of each of those both media, help in close contact on all their lateral surfaces [fr

  1. 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

  2. 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

  3. Evaluation of US demo helium-cooled blanket options

    International Nuclear Information System (INIS)

    Wong, C.P.C.; McQuillan, B.W.; Schleicher, R.W.

    1995-10-01

    A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed

  4. 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.)

  5. 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

  6. Choice of economical optimum blanket of hybrid reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blinkin, V L; Novikov, V M

    1981-01-01

    The economical effectiveness of symbiotic power systems depends on the choice of the correlation between energy production and fissile fuel production in blankets of controlled thermonuclear fusion reactor (CTR), what is investigated here. It is shown that the optimum value of this correlation essentially depends on the ratio between the specific costs for energy production in hybrid thermonuclear reactors and that in fission reactors as part of the symbiotic system.

  7. 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.)

  8. 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

  9. Comparison of inventory of tritium in various ceramic breeder blankets

    International Nuclear Information System (INIS)

    Nishikawa, M.; Beloglazov, S.; Nakashima, N.; Hashimoto, K.; Enoeda, M.

    2002-01-01

    It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of breeder material, absorption of water vapor into bulk of the grain, and adsorption of water on surface of the grain, together with the isotope exchange reaction between hydrogen in purge gas and tritium on surface of breeder material and the isotope exchange reaction between water vapor in purge gas and tritium on surface, for estimation of the tritium inventory in a uniform ceramic breeder blanket under the steady-state condition. It has been also pointed out by the present authors that the water formation reaction on the surface of ceramic breeder materials at introduction of hydrogen can give effect on behavior of bred tritium and lithium transfer in blanket. The tritium inventory for various ceramic breeder blankets are compared in this study basing on adsorption capacity, absorption capacity, isotope exchange capacity, and isotope exchange reactions on the Li 2 O, LiAlO 2 , Li 2 ZrO 3 , Li 4 SiO 4 and Li 2 TiO 3 surface experimentally obtained by the present authors. Effect of each mass transfer steps on the shape of release curve of bred tritium at change of the operational conditions is also discussed from the observation at out pile experiment in KUR. (orig.)

  10. 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

  11. 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

  12. NOEL: a no-leak fusion blanket concept

    International Nuclear Information System (INIS)

    Powell, J.R.; Yu, W.S.; Fillo, J.A.; Horn, F.L.; Makowitz, H.

    1980-01-01

    Analysis and tests of a no-leak fusion blanket concept (NOEL-NO External Leak) are described. Coolant cannot leak into the plasma chamber even if large through-cracks develop in the first wall. Blanket modules contain a two-phase material, A, that is solid (several cm thick) on the inside of the module shell, and liquid in the interior. The solid layer is maintained by imbedded tubes carrying a coolant, B, below the freezing point of A. Most of the 14-MeV neutron energy is deposited as heat in the module interior. The thermal energy flow from the module interior to the shell keeps the interior liquid. Pressure on the liquid A interior is greater than the pressure on B, so that B cannot leak out if failures occur in coolant tubes. Liquid A cannot leak into the plasma chamber through first wall cracks because of the intervening frozen layer. The thermal hydraulics and neutronics of NOEL blankets have been investigated for various metallic (e.g., Li, Pb 2 , LiPb, Pb) and fused salt choices for material A

  13. 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.

  14. Measurements relevant to simulating subcriticality in ADS facilities with blanket

    International Nuclear Information System (INIS)

    Titarenko, Yu. E.; Batyaev, V.F.; Borovlev, S.P.; Gladkikh, N.G.; Igumnov, M.M.; Legostaev, V.O.; Karpikhin, E.I.; Konev, V.N.; Kushnerev, Yu.T.; Popkov, V.N.; Ryazhsky, V.I.; Spiridonov, V.G.; Chernyavsky, E.V.; Shvedov, O.V.

    2009-10-01

    The work presents the results of determining the blanket subcriticality for a zero-power heavy water reactor MAKET at the Institute for Theoretical and Experimental Physics, Moscow. The blanket is hexagonal lattice made of 36 90%-enriched 235U fuel rods spaced 173mm apart. The subcriticality was varied from ∼0.3% to 5% by adjusting the heavy water level. The subcriticality values were calibrated using the dependence of reactivity on heavy water level. The pulsed neutron source technique was used to measure the temporal dependence of neutron field at different blanket points for the calibrated subcriticality values. The subciticality values obtained in terms of the 'inverse clock' formulae using the decay constants of the measured dependences proved to differ from the calibrated subcriticalities by not more than 7% at the average. The MCNP code-aided simulations of the experiment made has given the calibrated keff values at prescribed heavy water levels and led to the neutron field decay constants at given points, which differ on the average from their experimental values by not more than 7% too. (author)

  15. 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

  16. 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.

  17. Mirror hybrid reactor blanket and power conversion system conceptual design

    International Nuclear Information System (INIS)

    Schultz, K.R.; Backus, G.A.; Baxi, C.B.; Dee, J.B.; Estrine, E.A.; Rao, R.; Veca, A.R.

    1976-01-01

    The conceptual design of the blanket and power conversion system for a gas-cooled mirror hybrid fusion-fission reactor is presented. The designs of the fuel, blanket module and power conversion system are based on existing gas-cooled fission reactor technology that has been developed at General Atomic Company. The uranium silicide fuel is contained in Inconel-clad rods and is cooled by helium gas. The fuel is contained in 16 spherical segment modules which surround the fusion plasma. The hot helium is used to raise steam for a conventional steam cycle turbine generator. The details of the method of support for the massive blanket modules and helium ducts remain to be determined. Nevertheless, the conceptual design appears to be technically feasible with existing gas-cooled technology. A preliminary safety analysis shows that with the development of a satisfactory method of primary coolant circuit containment and support, the hybrid reactor could be licensed under existing Nuclear Regulatory Commission regulations

  18. Corrosion characteristics of an aqueous self-cooled fusion blanket

    International Nuclear Information System (INIS)

    Bogaerts, W.F.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Jackson, D.

    1986-01-01

    A novel aqueous self-cooled blanket concept (ASCB) has recently been proposed. This blanket concept, as applied to a MARS-like tandem mirror reactor, consists of disks of spiraling tubes of Zircaloy-4 housed in a structural container of vanadium alloy (V-15 Ti-5 Cr). The Zircaloy tubes are cooled by a mixture of light and heavy water with 9 g of LiOH per 100 cm 3 of water dissolved in the coolant. A major issue for the feasibility of the integrated blanket coil concept is the chemical compatibility of the coolant and Zircaloy. Initial corrosion tests have been undertaken in order to resolve this question. Results clearly show that successful alloy heats can be prepared, for which corrosion problems will probably not be the limiting factor of the ASCB design concept. As is quite well known from fission engineering studies, small variations in the alloy compositions or in the metallurgical structure may, however, be able to cause significant alterations in the oxidation or corrosion rates. Further tests will be necessary to resolve the remaining uncertainties and to determine the behavior of successful alloy heats in the presence of trace impurities in order to address the sensitivity to localized corrosion phenomena such as pitting, stress corrosion cracking, and intergranular attack

  19. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

    Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

  20. Super-stable Poissonian structures

    International Nuclear Information System (INIS)

    Eliazar, Iddo

    2012-01-01

    In this paper we characterize classes of Poisson processes whose statistical structures are super-stable. We consider a flow generated by a one-dimensional ordinary differential equation, and an ensemble of particles ‘surfing’ the flow. The particles start from random initial positions, and are propagated along the flow by stochastic ‘wave processes’ with general statistics and general cross correlations. Setting the initial positions to be Poisson processes, we characterize the classes of Poisson processes that render the particles’ positions—at all times, and invariantly with respect to the wave processes—statistically identical to their initial positions. These Poisson processes are termed ‘super-stable’ and facilitate the generalization of the notion of stationary distributions far beyond the realm of Markov dynamics. (paper)

  1. Super-stable Poissonian structures

    Science.gov (United States)

    Eliazar, Iddo

    2012-10-01

    In this paper we characterize classes of Poisson processes whose statistical structures are super-stable. We consider a flow generated by a one-dimensional ordinary differential equation, and an ensemble of particles ‘surfing’ the flow. The particles start from random initial positions, and are propagated along the flow by stochastic ‘wave processes’ with general statistics and general cross correlations. Setting the initial positions to be Poisson processes, we characterize the classes of Poisson processes that render the particles’ positions—at all times, and invariantly with respect to the wave processes—statistically identical to their initial positions. These Poisson processes are termed ‘super-stable’ and facilitate the generalization of the notion of stationary distributions far beyond the realm of Markov dynamics.

  2. swot: Super W Of Theta

    Science.gov (United States)

    Coupon, Jean; Leauthaud, Alexie; Kilbinger, Martin; Medezinski, Elinor

    2017-07-01

    SWOT (Super W Of Theta) computes two-point statistics for very large data sets, based on “divide and conquer” algorithms, mainly, but not limited to data storage in binary trees, approximation at large scale, parellelization (open MPI), and bootstrap and jackknife resampling methods “on the fly”. It currently supports projected and 3D galaxy auto and cross correlations, galaxy-galaxy lensing, and weighted histograms.

  3. BEWARE OF...SUPER GLUES!!

    CERN Multimedia

    2006-01-01

    What happened? A number of accidents have occurred with the use of 'Super Glues'. Some individuals have suffered injuries - severe irritation, or skin bonded together - through getting glue on their face and in their eyes. What are the hazards associated with glues? 'Super Glues' (i.e. cyanoacrylates): Are harmful if swallowed and are chemical irritants to the eyes, respiratory system and skin. Present the risk of polymerization (hardening) leading to skin damage. Be careful ! 'Super Glues' can bond to skin and eyes in seconds. Note: Other glues, resins and hardeners are also chemicals and as such can cause serious damage to the skin, eyes, respiratory or digestive tract. (For example: some components can be toxic, harmful, corrosive, sensitizing agents, etc.). How to prevent accidents in the future? Read the Material Safety Data Sheet (MSDS) for all of the glues you work with. Check the label on the container to find out which of the materials you work with are hazardous. Wear the right Per...

  4. 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

  5. Passive Collecting of Solar Radiation Energy using Transparent Thermal Insulators, Energetic Efficiency of Transparent Thermal Insulators

    Directory of Open Access Journals (Sweden)

    Smajo Sulejmanovic

    2014-11-01

    Full Text Available This paper explains passive collection of solar radiation energy using transparent thermal insulators. Transparent thermal insulators are transparent for sunlight, at the same time those are very good thermal insulators. Transparent thermal insulators can be placed instead of standard conventional thermal insulators and additionally transparent insulators can capture solar radiation, transform it into heat and save heat just as standard insulators. Using transparent insulators would lead to reduce in usage of fossil fuels and would help protection of an environment and reduce effects of global warming, etc.

  6. DEMO relevance of the test blanket modules in ITER-Application to the European test blanket modules

    International Nuclear Information System (INIS)

    Magnani, E.; Gabriel, F.; Boccaccini, L.V.; Li-Puma, A.

    2010-01-01

    Test blanket module (TBM) testing programme in ITER as a support to DEMO design is a very important step on the road map to commercial fusion reactors although it is an ambitious task. Finding as much as possible DEMO relevant tests in view of the future DEMO blanket design is therefore a major goal since ITER and DEMO environment and loading conditions are different. To clarify and quantify the meaning of the DEMO relevance, criteria using a structural, functional and behavioural representation of the breeding blanket acting as a system are investigated. Then, a three-step strategy is proposed to carry out TBM DEMO relevant tests associated with a TBM design modification strategy. Key parameters should intensively be used as target for TBM characterization and numerical code validation. When assessing the relevance, on the other hand, not only the actual difference between DEMO and ITER values should be considered, but also whether the analyzed phenomena have a threshold and a range of applicability, as numerical simulations are usually permitted within these limits. The proposed methodology is at the end applied to the design of the HCLL TBM breeding unit configuration.

  7. Continuous fine pattern formation by screen-offset printing using a silicone blanket

    Science.gov (United States)

    Nomura, Ken-ichi; Kusaka, Yasuyuki; Ushijima, Hirobumi; Nagase, Kazuro; Ikedo, Hiroaki; Mitsui, Ryosuke; Takahashi, Seiya; Nakajima, Shin-ichiro; Iwata, Shiro

    2014-09-01

    Screen-offset printing combines screen-printing on a silicone blanket with transference of the print from the blanket to a substrate. The blanket absorbs organic solvents in the ink, and therefore, the ink does not disperse through the material. This prevents blurring and allows fine patterns with widths of a few tens of micrometres to be produced. However, continuous printing deteriorates the pattern’s shape, which may be a result of decay in the absorption abilities of the blanket. Thus, we have developed a new technique for refreshing the blanket by substituting high-boiling-point solvents present on the blanket surface with low-boiling-point solvents. We analyse the efficacy of this technique, and demonstrate continuous fine pattern formation for 100 screen-offset printing processes.

  8. Continuous fine pattern formation by screen-offset printing using a silicone blanket

    International Nuclear Information System (INIS)

    Nomura, Ken-ichi; Kusaka, Yasuyuki; Ushijima, Hirobumi; Nagase, Kazuro; Ikedo, Hiroaki; Mitsui, Ryosuke; Takahashi, Seiya; Nakajima, Shin-ichiro; Iwata, Shiro

    2014-01-01

    Screen-offset printing combines screen-printing on a silicone blanket with transference of the print from the blanket to a substrate. The blanket absorbs organic solvents in the ink, and therefore, the ink does not disperse through the material. This prevents blurring and allows fine patterns with widths of a few tens of micrometres to be produced. However, continuous printing deteriorates the pattern’s shape, which may be a result of decay in the absorption abilities of the blanket. Thus, we have developed a new technique for refreshing the blanket by substituting high-boiling-point solvents present on the blanket surface with low-boiling-point solvents. We analyse the efficacy of this technique, and demonstrate continuous fine pattern formation for 100 screen-offset printing processes. (paper)

  9. Applications of the aqueous self-cooled blanket (ASCB) concept to the Next European Torus (NET)

    International Nuclear Information System (INIS)

    Embrechts, M.J.; Bogaerts, W.; Cardella, A.; Chazalon, M.; Danner, W.; Dinner, P.; Libin, B.

    1987-01-01

    The Aqueous Self-Cooled Blanket Concept (ASCB) leads to a low-technology blanket design that relies on just structural material and coolant with small amounts of lithium compound dissolved in the coolant to provide for tritium production. The application of the ASCB concept in NET is being considered as a driver blanket that would operate at low temperature and low pressure and provide a reliable environment for machine operation during the technology phase. Shielding and tritium production are the primary objectives for such a low-technology blanket. Net tritium breeding is not a design requirement per se for a driver blanket for NET. A DEMO relevant ASCB based blanket test module with (local) tritium self-sufficiency and energy recovery as primary objectives might also be tested in NET if future developments confirm their viability

  10. Positron Annihilation in Insulating Materials

    International Nuclear Information System (INIS)

    Asoka-Kumar, P; Sterne, PA

    2002-01-01

    We describe positron results from a wide range of insulating materials. We have completed positron experiments on a range of zeolite-y samples, KDP crystals, alkali halides and laser damaged SiO 2 . Present theoretical understanding of positron behavior in insulators is incomplete and our combined theoretical and experimental approach is aimed at developing a predictive understanding of positrons and positronium annihilation characteristics in insulators. Results from alkali halides and alkaline-earth halides show that positrons annihilate with only the halide ions, with no apparent contribution from the alkali or alkaline-earth cations. This contradicts the results of our existing theory for metals, which predicts roughly equal annihilation contributions from cation and anion. We also present result obtained using Munich positron microprobe on laser damaged SiO 2 samples

  11. Floquet topological insulators for sound

    Science.gov (United States)

    Fleury, Romain; Khanikaev, Alexander B.; Alù, Andrea

    2016-06-01

    The unique conduction properties of condensed matter systems with topological order have recently inspired a quest for the similar effects in classical wave phenomena. Acoustic topological insulators, in particular, hold the promise to revolutionize our ability to control sound, allowing for large isolation in the bulk and broadband one-way transport along their edges, with topological immunity against structural defects and disorder. So far, these fascinating properties have been obtained relying on moving media, which may introduce noise and absorption losses, hindering the practical potential of topological acoustics. Here we overcome these limitations by modulating in time the acoustic properties of a lattice of resonators, introducing the concept of acoustic Floquet topological insulators. We show that acoustic waves provide a fertile ground to apply the anomalous physics of Floquet topological insulators, and demonstrate their relevance for a wide range of acoustic applications, including broadband acoustic isolation and topologically protected, nonreciprocal acoustic emitters.

  12. Fermi surfaces in Kondo insulators

    Science.gov (United States)

    Liu, Hsu; Hartstein, Máté; Wallace, Gregory J.; Davies, Alexander J.; Ciomaga Hatnean, Monica; Johannes, Michelle D.; Shitsevalova, Natalya; Balakrishnan, Geetha; Sebastian, Suchitra E.

    2018-04-01

    We report magnetic quantum oscillations measured using torque magnetisation in the Kondo insulator YbB12 and discuss the potential origin of the underlying Fermi surface. Observed quantum oscillations as well as complementary quantities such as a finite linear specific heat capacity in YbB12 exhibit similarities with the Kondo insulator SmB6, yet also crucial differences. Small heavy Fermi sections are observed in YbB12 with similarities to the neighbouring heavy fermion semimetallic Fermi surface, in contrast to large light Fermi surface sections in SmB6 which are more similar to the conduction electron Fermi surface. A rich spectrum of theoretical models is suggested to explain the origin across different Kondo insulating families of a bulk Fermi surface potentially from novel itinerant quasiparticles that couple to magnetic fields, yet do not couple to weak DC electric fields.

  13. Topological insulators and topological superconductors

    CERN Document Server

    Bernevig, Andrei B

    2013-01-01

    This graduate-level textbook is the first pedagogical synthesis of the field of topological insulators and superconductors, one of the most exciting areas of research in condensed matter physics. Presenting the latest developments, while providing all the calculations necessary for a self-contained and complete description of the discipline, it is ideal for graduate students and researchers preparing to work in this area, and it will be an essential reference both within and outside the classroom. The book begins with simple concepts such as Berry phases, Dirac fermions, Hall conductance and its link to topology, and the Hofstadter problem of lattice electrons in a magnetic field. It moves on to explain topological phases of matter such as Chern insulators, two- and three-dimensional topological insulators, and Majorana p-wave wires. Additionally, the book covers zero modes on vortices in topological superconductors, time-reversal topological superconductors, and topological responses/field theory and topolo...

  14. First Super-Earth Atmosphere Analysed

    Science.gov (United States)

    2010-12-01

    The atmosphere around a super-Earth exoplanet has been analysed for the first time by an international team of astronomers using ESO's Very Large Telescope. The planet, which is known as GJ 1214b, was studied as it passed in front of its parent star and some of the starlight passed through the planet's atmosphere. We now know that the atmosphere is either mostly water in the form of steam or is dominated by thick clouds or hazes. The results will appear in the 2 December 2010 issue of the journal Nature. The planet GJ 1214b was confirmed in 2009 using the HARPS instrument on ESO's 3.6-metre telescope in Chile (eso0950) [1]. Initial findings suggested that this planet had an atmosphere, which has now been confirmed and studied in detail by an international team of astronomers, led by Jacob Bean (Harvard-Smithsonian Center for Astrophysics), using the FORS instrument on ESO's Very Large Telescope. "This is the first super-Earth to have its atmosphere analysed. We've reached a real milestone on the road toward characterising these worlds," said Bean. GJ 1214b has a radius of about 2.6 times that of the Earth and is about 6.5 times as massive, putting it squarely into the class of exoplanets known as super-Earths. Its host star lies about 40 light-years from Earth in the constellation of Ophiuchus (the Serpent Bearer). It is a faint star [2], but it is also small, which means that the size of the planet is large compared to the stellar disc, making it relatively easy to study [3]. The planet travels across the disc of its parent star once every 38 hours as it orbits at a distance of only two million kilometres: about seventy times closer than the Earth orbits the Sun. To study the atmosphere, the team observed the light coming from the star as the planet passed in front of it [4]. During these transits, some of the starlight passes through the planet's atmosphere and, depending on the chemical composition and weather on the planet, specific wavelengths of light are

  15. 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

  16. 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.

  17. 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.

  18. Structural design study of tritium breeding blanket with a lead layer as a neutron multiplier

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kitamura, Kazunori; Minato, Akio; Sakamoto, Hiroki; Yamamoto, Takashi

    1980-12-01

    Thermal and structural design study of a tritium breeding blanket with a lead layer for a International Tokamak Reactor (INTOR) is carried out. Tube in shell type blanket with a lead layer is found to be promising. The volume fraction of structural material in the lead layer can be small enough to keep the neutron multiplication effect of lead. Reasonable value of shell effect is attainable due to lead layer in the front part of the blanket. (author)

  19. Design of the breeder units in the new HCPB modular blanket concept and material requirements

    International Nuclear Information System (INIS)

    Boccaccini, L.V.; Fischer, U.; Hermsmeyer, S.; Reimann, J.; Xu, Z.; Koehly, C.

    2004-01-01

    ; according to the experience from the old design no major problems are expected concerning stress levels and gap formation in the pebble beds; in fact, the situation should be more favourable due to the reduced dimensions (max. 20 cm) of the beds that should minimise ratcheting and particle flow phenomena. In respect to tritium extraction, the most favourable features of the HCPB concept (e.g. an overall low partial pressure of tritium in the beds that minimise permeation into the main coolant system) can be kept in the new design; a complication could be the necessity to provide each cell with a system of tubes to inlet the purge helium in the front part of the beds or to divide the purge flow for Be and CB. The modular design of the new HCPB blanket, that makes the breeder units almost independent on the structural design of the box, opens interesting possibilities in the development of these units. The present design can be optimised on the basis of the results of the present R and D programme on Be and CB. In addition, new requirements could appear to improve the design performances or manufacturing: i.e. in respect to filling procedures of the beds (use of pre-packed breeder units) or the necessity of insulating layer to thermally decouple the breeder units from the box or the selection of new materials with a better compatibility with Be and CB at high temperatures as protection of the steel structure. (author)

  20. Topological insulators fundamentals and perspectives

    CERN Document Server

    Ortmann, Frank; Valenzuela, Sergio O

    2015-01-01

    There are only few discoveries and new technologies in physical sciences that have the potential to dramatically alter and revolutionize our electronic world. Topological insulators are one of them. The present book for the first time provides a full overview and in-depth knowledge about this hot topic in materials science and condensed matter physics. Techniques such as angle-resolved photoemission spectrometry (ARPES), advanced solid-state Nuclear Magnetic Resonance (NMR) or scanning-tunnel microscopy (STM) together with key principles of topological insulators such as spin-locked electronic