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Sample records for armored blanket development

  1. Flexible armored blanket development

    Energy Technology Data Exchange (ETDEWEB)

    Roth, E.S.

    1978-05-01

    An exploratory development contract was undertaken on December 23, 1977 which had as its purpose the development and demonstration of a flexible armored blanket design suitable for providing ballistic protection to nuclear weapons during shipment. Objectives were to design and fabricate a prototype blanket which will conform to the weapon shape, is troop-handleable in the field, and which, singly or in multiple layers, can defeat a range of kinetic energy armor piercing (AP) ammunition potentially capable of damaging the critical portion of the nuclear weapon. Following empirical testing, including the firing of threat ammunition under controlled laboratory and field test conditions, materials were selected and assembled into two blanket designs, each weighing approximately 54 kg/m{sup 2} (11 lbs/ft{sup 2}) and estimated to cost from $111 to $180 per ft{sup 2} in production. A firing demonstration to evidence blanket performance against terrorist/light infantry weapons, heavy infantry weapons, and aircraft cannon was conducted for representatives of the DOD and interested Sandia employees on April 12, 1978. The blankets performed better than anticipated defeating bullets up to 7.62 mm x 51 mm AP with one layer and projectiles up to 23 mm HEI with two layers. Based on these preliminary tests it is recommended that development work be continued with the following objectives: (1) the selection by the DOD of priority applications, (2) the specific design and fabrication of sufficient quantities of armored blankets for field testing, (3) the evaluation of the blankets by DOD operational units, with reports to Sandia Laboratories to enable final design.

  2. Development of a New Armor Steel and its Ballistic Performance

    OpenAIRE

    S. Hakan Atapek

    2013-01-01

    In this study, a boron added armor steel was developed according to standard rolled homogenous armor steel, MIL-A-12560, and metallographic-fractographic examinations were carried out to understand its deformation characteristics and perforation mode after interaction with a 7.62 mm armor piercing projectile. The microstructure of the developed steel was characterized by light and scanning electron microscope to evaluate its matrix after application of several heat treatments consisting of au...

  3. Development of Solid Breeder Blanket at JAERI

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been performing blanket development based on the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. The blanket development consists of out-pile R and D, In-pile R and D, TBM Neutronics and TPR Tests and Tritium Recovery System R and D. Based on the achievements of element technology development, the R and D program is now stepping to the engineering testing phase, in which scalable mockup tests will be performed for obtaining engineering data unique to the specific structure of the components, with the objective to define the fabrication specification of test blanket modules for ITER. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI

  4. Development of a New Armor Steel and its Ballistic Performance

    Directory of Open Access Journals (Sweden)

    S. Hakan Atapek

    2013-05-01

    Full Text Available In this study, a boron added armor steel was developed according to standard rolled homogenous armor steel, MIL-A-12560, and metallographic-fractographic examinations were carried out to understand its deformation characteristics and perforation mode after interaction with a 7.62 mm armor piercing projectile. The microstructure of the developed steel was characterized by light and scanning electron microscope to evaluate its matrix after application of several heat treatments consisting of austenization, quenching and tempering. The mechanical properties of the developed steel were determined by tensile test at room temperature and notched impact test at -40 ºC. The ballistic performance of developed steel was determined by its V50 ballistic protection limit according to MIL-STD-662F standard and it was found to be higher than that of MIL-A-12560 steel. After perforation deformation induced adiabatic shear bands, that have an important role on the crack nucleation, were observed close to the penetration in the etched steel and perforation occurred by typical ductile hole enlargement with certain radial flows.Defence Science Journal, 2013, 63(3, pp.271-277, DOI:http://dx.doi.org/10.14429/dsj.63.1341

  5. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    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)

  6. 坦克装甲车辆装甲防护发展研究%Armor Protection Development of Tank & Armored Vehicle

    Institute of Scientific and Technical Information of China (English)

    房凌晖; 郑翔玉; 马丽; 汪伦根

    2014-01-01

    This article first discusses the advances and the applications of main armor materials,including armour steel,aluminum alloys,titanium alloy,ceramic and composite materials. Then,the basical theory and research status of some typical armor protection technologies are introduced in detail,such as explosive reactive armor,composite armor,grille armor,electromagnetic armor and so on. And on this basis,the future development trend of armor protection technology is also discussed,which is important to improve the protection ability and viability of armored vehicle.%阐述了装甲钢、铝合金、钛合金、陶瓷和复合材料等主流装甲防护材料的研究现状和应用情况,并对爆炸反应装甲、复合装甲、格栅装甲和电磁装甲等几种典型装甲防护技术的基本原理和研究现状进行了介绍;分析了未来装甲防护技术的发展方向和趋势,对提高装甲车辆的防护能力和战场生存能力具有重要意义。

  7. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li17Pb83 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 Li17Pb83 blankets. (author)

  8. Development of the Helium Cooled Lithium Lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, G., E-mail: giacomo.aiello@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aubert, J.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The HCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • The new design has been developed with the aim to capitalize on TBM experience in ITER. • A new attachment system for the modules has been proposed. - Abstract: The Helium Cooled Lithium Lead (HCLL) blanket is one of the candidate European blanket concepts selected for the DEMOnstration fusion power plant that should follow ITER. In a fusion power plant, the blanket is one of the key components because of its impact on the plant performance, availability, safety and economics. In 2012, the European Fusion Development Agreement (EFDA) agency issued new specifications for DEMO: this paper describes the work performed to adapt the previous 2007 HCLL-DEMO blanket design to those specifications. A new segmentation has been defined assuming straight surfaces for all blanket modules. Following the Multi Module Segment (MMS) option, all modules are attached to a common back supporting structure which also serves as manifold for Helium and PbLi distribution. A detailed CAD design of the central outboard module has been defined. Thermo-hydraulic and thermo-mechanical analyses on of the First Wall and Breeder Zone have been carried out. For the attachment of the modules to the common backplate, a new solution based on the use of Tie Rods, derived from the design of the corresponding HCLL Test Blanket Module for ITER, has been proposed. This paper also identifies the priorities for further development of the HCLL blanket design.

  9. Development of the water cooled lithium lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aiello, G.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The WCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • Preliminary CAD design of the equatorial outboard module of the WCLL blanket has been developed for DEMO. • Finite elements analyses have been carried out in order to assess the module thermal behavior in the straight part of the module. - Abstract: The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant. In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications. Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant. A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood. The priorities for further development of the WCLL blanket concept are identified in the paper.

  10. Armor Development from Decapitated Flash Flood Bores in Supply-Limited Flume Experiments

    Science.gov (United States)

    Goodwin, K.; Rhodes, R.; Johnson, J. P.

    2013-12-01

    In rivers assumed to have quasi-normal flow, three main processes have been used to explain bed surface armoring: i) selective entrainment and transport of smaller grains, ii) limited supply of smaller grain sizes, and iii) equal mobility of grains of different sizes, which develops through natural feedbacks such that larger, less mobile grains are enriched on the surface relative to smaller grains. Flash flood-dominated river channels in arid environments often completely lack surface armoring, yet it is unclear whether increased sediment supply or transport of all grain sizes prevents armor development. In order to examine armor development in an end-member case of non-normal flow, we conducted a series of laboratory experiments using flash flood bores. The flume is 33.5 m long, 0.5 m wide, 0.8 m tall, and capable of creating reproducible flood bores by raising a high-speed computerized lift gate and releasing impounded water. For each experiment, the gate was quickly lowered as soon as the flood bore traveled the length of the flume, 'decapitating' the bore from subsequent flow, to better isolate the effects of the bore alone on entrainment and transport. Sediment was not fed into the upstream end of the flume and only sourced from the gravel bed (2 mm to 40 mm), resulting in supply-limited experimental conditions. In response to repeated flood bores, the surface grain size distribution rapidly coarsened. We interpret that kinetic sieving was the dominant cause of surface armoring in these experiments. LiDAR scans of the bed topography from before and after each bore show increased surface roughness due to grain size changes, but small surface elevation changes due to relatively limited erosion. Digital gravelometry from photographs taken after each bore show increased armoring, while sediment transported out the downstream end of the flume tended to be as coarse or coarser than the bed surface. Travel distances of three sizes of RFID-tagged tracer clasts show

  11. Fusion blanket materials development and recent R and D activities

    International Nuclear Information System (INIS)

    Development of structural materials plays an important role in the feasibility of fusion power plant. The candidate structural materials for future fusion reactors are Reduced Activation Ferritic Martensitic (RAFM) steel, nano structured ODS Steel, vanadium alloys and SiC/SiCf composite etc. RAFM steel is presently considered as the structural material for Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) because of its high void swelling resistance and improved thermal properties compared to austenitic steel. Development of RAFM steel in India is being carried out in full swing in collaboration with various research laboratories and steel industries. This paper presents an overview of the Indian activities on fusion blanket materials and describes in brief the efforts made to develop IN-RAFM steel as structural material for the LLCB TBM. In future, due to enhanced properties of vanadium base alloy and nano structured materials like ODS RAFMS, RAFM steel may be replaced by these materials for its application in DEMO relevant fusion reactor. Future R and D activities will be specifically towards the development of these structural materials for fusion reactor

  12. Development of a remote handling system for replacement of armor tiles in the Fusion Experimental Reactor

    International Nuclear Information System (INIS)

    The armor tiles of the Fusion Experimental Reactor (FER) planned by JAERI are categorized as scheduled maintenance components, since they are damaged by severe heat and particle loads from the plasma during operation. A remote handling system is thus required to replace a large number of tiles rapidly in the highly activated reactor. However, the simple teaching-playback method cannot be adapted to this system because of deflection of the tiles caused by thermal deformation and so on. We have developed a control system using visual feedback control to adapt to this deflection and an end-effector for a single arm. We confirm their performance in tests. (orig.)

  13. Integrated development of light armored vehicles based on wargaming simulators

    Science.gov (United States)

    Palmarini, Marc; Rapanotti, John

    2004-08-01

    Vehicles are evolving into vehicle networks through improved sensors, computers and communications. Unless carefully planned, these complex systems can result in excessive crew workload and difficulty in optimizing the use of the vehicle. To overcome these problems, a war-gaming simulator is being developed as a common platform to integrate contributions from three different groups. The simulator, OneSAF, is used to integrate simplified models of technology and natural phenomena from scientists and engineers with tactics and doctrine from the military and analyzed in detail by operations analysts. This approach ensures the modelling of processes known to be important regardless of the level of information available about the system. Vehicle survivability can be improved as well with better sensors, computers and countermeasures to detect and avoid or destroy threats. To improve threat detection and reliability, Defensive Aids Suite (DAS) designs are based on three complementary sensor technologies including: acoustics, visible and infrared optics and radar. Both active armour and softkill countermeasures are considered. In a typical scenario, a search radar, providing continuous hemispherical coverage, detects and classifies the threat and cues a tracking radar. Data from the tracking radar is processed and an explosive grenade is launched to destroy or deflect the threat. The angle of attack and velocity from the search radar can be used by the soft-kill system to carry out an infrared search and track or an illuminated range-gated scan for the threat platform. Upon detection, obscuration, countermanoeuvres and counterfire can be used against the threat. The sensor suite is completed by acoustic detection of muzzle blast and shock waves. Automation and networking at the platoon level contribute to improved vehicle survivability. Sensor data fusion is essential in avoiding catastrophic failure of the DAS. The modular DAS components can be used with Light Armoured

  14. ITER breeding blanket module design and analysis

    International Nuclear Information System (INIS)

    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. Development and analysis of fusion breeder blanket neutronics. Progress report, November 1, 1983-October 31, 1984

    International Nuclear Information System (INIS)

    The following activities are briefly described: (a) the IBM versions of the computer codes FORSS, PUFF-II, ONETRAN, TWOTRAN-II, and DOT4.3 were obtained from the Radiation Shielding Information Center (RSIC) and have been implemented on the UCLA local computer, the IBM 3033; (b) mathematical and computational models to describe the time-dependent transport and inventory of tritium in individual components of a fusion reactor system have been developed; (c) extensive cross-section sensitivity and uncertainty analysis was carried out to evaluate an estimate for the uncertainty associated with the TBR (both from 6Li and 7Li, individually) in four of the leading blanket concepts (the Li2O/HT-9 helium-cooled blanket, the 17Li-83Pb/PCA self-cooled blanket, the LiAlO2/He/FS/Be blanket, and the flibe/He/FS/Be blanket); (d) as far as the TBR obtain able in various blanket concepts is concerned, a comparative analysis was carried out to estimate the change in TBR in a particular blanket module when placed in a tokamak machine [R (first wall) approx. 2 m] as opposed to adopting the same blanket in a mirror machine [R (first wall) approx. 50 cm] with the same wall loading

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-15

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

  18. Breeding blanket for DEMO

    International Nuclear Information System (INIS)

    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 being investigated within the framework of the European Test-Blanket Development Programme. (orig.)

  19. Breeding blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Proust, E. (Commissariat a l' Energie Atomique (CEA), DRN/DMT/SERMA, CE, Saclay (France)); Anzidei, L. (ENEA/FUS, C.R.E., Frascati (Italy)); Casini, G. (Commission of the European Communities, Joint Research Center, Ispara (Italy)); Dalle Donne, M. (Kernforschungszentrum Karlsruhe GmbH (Germany)); Giancarli, L. (Commissariat a l' Energie Atomique (CEA), DRN/DMT/SERMA, CE, Saclay (France)); Malang, S. (Kernforschungszentrum Karlsruhe GmbH (Germany))

    1993-03-01

    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 being investigated within the framework of the European Test-Blanket Development Programme. (orig.)

  20. Breeding blanket for Demo

    Energy Technology Data Exchange (ETDEWEB)

    Proust, E.; Giancarli, L. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Anzidei, L. [ENEA, Frascati (Italy). Centro Ricerche Energia; Casini, G. [Commission of the European Communities, Ispra (Italy). Joint Research Centre; Dalle Donne, M.; Malang, S. [Kernforschungszentrum Karlsruhe GmbH (Germany)

    1992-12-31

    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.

  1. Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Ezato, Koichiro; Seki, Yohji; Yoshikawa, Akira; Tsuru, Daigo; Akiba, Masato [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan)

    2012-08-15

    The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

  2. Composite armor, armor system and vehicle including armor system

    Science.gov (United States)

    Chu, Henry S.; Jones, Warren F.; Lacy, Jeffrey M.; Thinnes, Gary L.

    2013-01-01

    Composite armor panels are disclosed. Each panel comprises a plurality of functional layers comprising at least an outermost layer, an intermediate layer and a base layer. An armor system incorporating armor panels is also disclosed. Armor panels are mounted on carriages movably secured to adjacent rails of a rail system. Each panel may be moved on its associated rail and into partially overlapping relationship with another panel on an adjacent rail for protection against incoming ordnance from various directions. The rail system may be configured as at least a part of a ring, and be disposed about a hatch on a vehicle. Vehicles including an armor system are also disclosed.

  3. Tritium processing for the European test blanket systems: current status of the design and development strategy

    International Nuclear Information System (INIS)

    Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant test blanket module and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. (authors)

  4. Tritium processing for the European test blanket systems: current status of the design and development strategy

    Energy Technology Data Exchange (ETDEWEB)

    Ricapito, I.; Calderoni, P.; Poitevin, Y. [Fusion for Energy, Barcelona (Spain); Aiello, A.; Utili, M. [ENEA, Camugnano (Italy); Demange, D. [Karlsruhe Institute of Technology - KIT, Karlsruhe (Germany)

    2015-03-15

    Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant test blanket module and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. (authors)

  5. Development of ITER shielding blanket prototype mockup by HIP bonding

    International Nuclear Information System (INIS)

    A prototype (∼900H x 1700W x 350T mm) of the ITER shielding blanket module has been fabricated following the previous successful fabrication of a small-scale (∼500H x 400W x 150T mm) and mid-scale (∼800H x 500W x 350T mm) mock-ups. This prototype incorporates most of key design features essential to the fabrication of the ITER shielding blanket module such as 1) the first wall heat sink made of Al2O3 dispersion strengthened Cu (DSCu) with built-in SS316L coolant tubes bonded to a massive SS316LN shield block, 2) toroidally curved first wall with a radius of 5106 mm while straight in poloidal direction, 3) coolant channels oriented in poloidal direction in the first wall and in toroidal direction in the shield block, 4) the first wall coolant channel routing to avoid the interference with the front access holes, 5) coolant channels drilled through the forged SS316LN-IG shield block, and 6) four front access holes of 30 mm in diameter penetrated through the first wall and the shield block. For the joining method, especially for the first wall/side wall parts and the shield block, the solid HIP (Hot Isostatic Pressing) process was applied. It is difficult to apply conventional joining methods such as field welding, brazing, explosion bonding and mechanical one-axial diffusion bonding to a wide area bonding because sufficient mechanical strengths can not be obtained and excessive deformations occurs. In order to solve these fabrication issues, HIP bonding was applied. The first wall stainless steel (SS) coolant tubes of 10 mm in inner diameter and l mm in thickness were sandwiched by semi-circular grooved DSCu plates at the first wall and the front region of the side wall, and by semi-circular grooved SS plates at the back region of the side wall. After assembling of these first wall/side wall parts with the shield block, they were simultaneously bonded by single step HIP in order to minimize thermal effects on the mechanical properties and to reduce the number

  6. Development of refractory armored silicon carbide by infrared transient liquid phase processing

    Science.gov (United States)

    Hinoki, Tatsuya; Snead, Lance L.; Blue, Craig A.

    2005-12-01

    Tungsten (W) and molybdenum (Mo) were coated on silicon carbide (SiC) for use as a refractory armor using a high power plasma arc lamp at powers up to 23.5 MW/m 2 in an argon flow environment. Both tungsten powder and molybdenum powder melted and formed coating layers on silicon carbide within a few seconds. The effect of substrate pre-treatment (vapor deposition of titanium (Ti) and tungsten, and annealing) and sample heating conditions on microstructure of the coating and coating/substrate interface were investigated. The microstructure was observed by scanning electron microscopy (SEM) and optical microscopy (OM). The mechanical properties of the coated materials were evaluated by four-point flexural tests. A strong tungsten coating was successfully applied to the silicon carbide substrate. Tungsten vapor deposition and pre-heating at 5.2 MW/m 2 made for a refractory layer containing no cracks propagating into the silicon carbide substrate. The tungsten coating was formed without the thick reaction layer. For this study, small tungsten carbide grains were observed adjacent to the interface in all conditions. In addition, relatively large, widely scattered tungsten carbide grains and a eutectic structure of tungsten and silicon were observed through the thickness in the coatings formed at lower powers and longer heating times. The strength of the silicon carbide substrate was somewhat decreased as a result of the processing. Vapor deposition of tungsten prior to powder coating helped prevent this degradation. In contrast, molybdenum coating was more challenging than tungsten coating due to the larger coefficient of thermal expansion (CTE) mismatch as compared to tungsten and silicon carbide. From this work it is concluded that refractory armoring of silicon carbide by Infrared Transient Liquid Phase Processing is possible. The tungsten armored silicon carbide samples proved uniform, strong, and capable of withstanding thermal fatigue testing.

  7. Development of refractory armored silicon carbide by infrared transient liquid phase processing

    International Nuclear Information System (INIS)

    Tungsten (W) and molybdenum (Mo) were coated on silicon carbide (SiC) for use as a refractory armor using a high power plasma arc lamp at powers up to 23.5 MW/m2 in an argon flow environment. Both tungsten powder and molybdenum powder melted and formed coating layers on silicon carbide within a few seconds. The effect of substrate pre-treatment (vapor deposition of titanium (Ti) and tungsten, and annealing) and sample heating conditions on microstructure of the coating and coating/substrate interface were investigated. The microstructure was observed by scanning electron microscopy (SEM) and optical microscopy (OM). The mechanical properties of the coated materials were evaluated by four-point flexural tests. A strong tungsten coating was successfully applied to the silicon carbide substrate. Tungsten vapor deposition and pre-heating at 5.2 MW/m2 made for a refractory layer containing no cracks propagating into the silicon carbide substrate. The tungsten coating was formed without the thick reaction layer. For this study, small tungsten carbide grains were observed adjacent to the interface in all conditions. In addition, relatively large, widely scattered tungsten carbide grains and a eutectic structure of tungsten and silicon were observed through the thickness in the coatings formed at lower powers and longer heating times. The strength of the silicon carbide substrate was somewhat decreased as a result of the processing. Vapor deposition of tungsten prior to powder coating helped prevent this degradation. In contrast, molybdenum coating was more challenging than tungsten coating due to the larger coefficient of thermal expansion (CTE) mismatch as compared to tungsten and silicon carbide. From this work it is concluded that refractory armoring of silicon carbide by Infrared Transient Liquid Phase Processing is possible. The tungsten armored silicon carbide samples proved uniform, strong, and capable of withstanding thermal fatigue testing

  8. Coastal Erosion Armoring 2005

    Data.gov (United States)

    California Department of Resources — Coastal armoring along the coast of California, created to provide a database of all existing coastal armoring based on data available at the time of creation....

  9. Distribution of bog and heath in a Newfoundland blanket bog complex: topographic limits on the hydrological processes governing blanket bog development

    Directory of Open Access Journals (Sweden)

    P. A. Graniero

    1999-01-01

    Full Text Available This research quantified the role of topography and hydrological processes within and, hence, the development of, blanket bogs. Topographic characteristics were derived from digital elevation models (DEMs developed for the surface and underlying substrate at three blanket bog sites on the southeastern lobe of the Avalon Peninsula, Newfoundland. A multinomial logit (MNL model of the probability of bog occurrence was constructed in terms of relevant topographic characteristics. The resulting model was then used to investigate the probabilistic boundary conditions of bog occurrence within the landscape. Under average curvatures for the sites studied, substrate slopes up to 0.065 favoured blanket bog development. However, steeper slopes could, theoretically, be occupied by blanked bog where water is concentrated by convergent curvatures or large contributing areas. Near community boundaries, bog and heath communities both occupied similar topographic conditions. Since these boundary locations are capable of supporting the hydrological conditions necessary for bog development, the heath is likely to be encroached upon by bog.

  10. Water-cooled, fire boom blanket, test and evaluation for system prototype development

    International Nuclear Information System (INIS)

    Initial development of actively cooled fire booms indicated that water-cooled barriers could withstand direct oil fire for several hours with little damage if cooling water were continuously supplied. Despite these early promising developments, it was realized that to build reliable full-scale system for Navy host salvage booms would require several development tests and lengthy evaluations. In this experiment several types of water-cooled fire blankets were tested at the Oil and Hazardous Materials Simulated Test Tank (OHMSETT). After the burn test the blankets were inspected for damage and additional tests were conducted to determine handling characteristics for deployment, recovery, cleaning and maintenance. Test results showed that water-cooled fire boom blankets can be used on conventional offshore oil containment booms to extend their use for controlling large floating-oil marine fires. Results also demonstrated the importance of using thermoset rubber coated fabrics in the host boom to maintain sufficient reserve seam strength at elevated temperatures. The suitability of passively cooled covers should be investigated to protect equipment and boom from indirect fire exposure. 1 ref., 2 tabs., 8 figs

  11. Requirements for a helium-cooled blanket heat removal system development facility for fusion reactor research

    International Nuclear Information System (INIS)

    Existing and potential design problems associated with the helium-cooled blanket assemblies of experimental, demonstration and hybrid reactor designs considered in the Magnetic Fusion Energy (MFE) Program were assessed. It was observed that a balanced program of design, analysis and experimentation would be required to develop, verify and qualify these designs and those of related hardware and equipment. To respond to the potential experimental requirements of the first-generation reactors (the EPRs and possibly the hybrid concept), the need for a helium test facility was identified. It was determined that this facility should have the capacity for recirculating 100,000 kg/hr of helium at 70 atm and 6000C and should have 3 MW of electrical power available for simulating neutron heating. No radioactive material or processes should be used to facilitate ''hands-on'' experimentation and development. The general types of testing anticipated in this facility would include: (1) thermal and coolant flow performance of the blanket and other components in the primary cooling circuit; (2) structural adequacy of the blanket and first wall including vibration considerations; (3) capability for accommodating safety/off-normal conditions. Existing facilities worldwide were surveyed. It was determined that a number of facilities exist in foreign nations for performing the anticipated experiments. However, no large helium gas flow loop exists within the USA. Consequently, it is recommended that a helium thermal-hydraulic blanket test facility be planned and build on a schedule that will meet the unique design development and verification needs of the fusion program. This report provides the rationale and preliminary scoping of the operational characteristics and requirements for such a facility

  12. Imaging body armor.

    Science.gov (United States)

    Harcke, H Theodore; Schauer, David A; Harris, Robert M; Campman, Steven C; Lonergan, Gael J

    2002-04-01

    This study examined the feasibility of performing radiographic studies on patients wearing standard-issue body armor. The Kevlar helmet, fragmentation vest, demining suit sleeve, and armor plate were studied with plain film and computed tomography in a simulated casualty situation. We found that the military helmet contains metal screws and metal clips in the headband, but diagnostic computed tomographic images can be obtained. Kevlar, the principal component of soft armor, has favorable photon attenuation characteristics. Plate armor of composite material also did not limit radiographic studies. Therefore, when medically advantageous, patients can be examined radiographically while wearing standard military body armor. Civilian emergency rooms should be aware of these observations because law enforcement officers wear similar protective armor. PMID:11977874

  13. Note on Armor Steel Design: A proposition

    Directory of Open Access Journals (Sweden)

    S. K. Potay

    1982-07-01

    Full Text Available A theoretical approach to the ballistic penetration and armor development has been made and a single phase material with a computer evaluated optimum hardness has been proposed to give a superior performance as an armour.

  14. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 2. Detailed version

    International Nuclear Information System (INIS)

    A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary. Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated RandD-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required RandD-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.)

  15. Development of Reduced Activation Ferritic-Martensitic Steels and fabrication technologies for Indian test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Baldev [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Jayakumar, T., E-mail: tjk@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2011-10-01

    For the development of Reduced Activation Ferritic-Martensitic Steel (RAFMS), for the Indian Test Blanket Module for ITER, a 3-phase programme has been adopted. The first phase consists of melting and detailed characterization of a laboratory scale heat conforming to Eurofer 97 composition, to demonstrate the capability of the Indian industry for producing fusion grade steel. In the second phase which is currently in progress, the chemical composition will be optimized with respect to tungsten and tantalum for better combination of mechanical properties. Characterization of the optimized commercial scale India-specific RAFM steel will be carried out in the third phase. The first phase of the programme has been successfully completed and the tensile, impact and creep properties are comparable with Eurofer 97. Laser and electron beam welding parameters have been optimized and welding consumables were developed for Narrow Gap - Gas Tungsten Arc welding and for laser-hybrid welding.

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

    International Nuclear Information System (INIS)

    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)

  17. ITER blanket, shield and material data base

    International Nuclear Information System (INIS)

    As part of the summary of the Conceptual Design Activities (CDA) for the International Thermonuclear Experimental Reactor (ITER), this document describes the ITER blanket, shield, and material data base. Part A, ''ITER Blanket and Shield Conceptual Design'', discusses the need for ITER of a tritium breeding blanket to supply most of the tritium for the fuel cycle of the device. Blanket and shield combined must be designed to operate at a neutron wall loading of 1MW/m2, and to provide adequate shielding of the magnets to meet the neutron energy fluence goal of 3MWa/m2 at the first wall. After a summary of the conceptual design, the following topics are elaborated upon: (1) function, design requirement, and critical issues; (2) material selection; (3) blanket and shield segmentation; (4) blanket design description; (5) design analysis; (6) shield; (7) radiation streaming analysis; and (8) a summary of benchmark calculations. Part B, ''ITER Materials Evaluation and Data Base'', treats the compilation and assessment of the available materials data base used for the selection of the appropriate materials for all major components of ITER, including (i) structural materials for the first wall, (ii) Tritium breeding materials for the blanket, (iii) plasma facing materials for the divertor and first wall armor, and (4) electric insulators for use in the blanket and divertor. Refs, figs and tabs

  18. 75 FR 25208 - Announcement of Body Armor Research Needs Meeting

    Science.gov (United States)

    2010-05-07

    ... discuss test methods, technologies, and R&D that can significantly improve ballistic protection through... Identify/define research needs for body armor Develop performance standards Demonstrate performance...

  19. Nanotechnology for armor: hype, facts, and future

    Science.gov (United States)

    Maher, Mick

    2012-06-01

    Over the past two decades, nanotechnology has offered the promise of revolutionary performance improvements over existing armor materials. During that time there was substantial effort and resources put into developing the material technology and supporting theories, with only limited emphasis placed on understanding the ballistic event, mechanisms that drive armor performance, and the dependent nature of the threat. As a result, this large investment in nanotechnology for armor has not produced improved performance on the ballistics testing range, and armor nanotechnologies have never been fielded. No matter what the platform, armor systems have several functions that they have to perform in order to function properly. In order to defeat a threat, armor systems are designed to: deform/deflect the threat; dissipate energy; and prevent residual debris penetration. To date there is no definitive answer as to what material properties drive the system behavior of these functions at high rates in response to a specific threat, making the adaptation of nanotechnology that much harder. However, these functions are now being considered with respect to the material system and armor mechanism being utilized, and nanotechnology is beginning to be shown as an effective means of improving performance. When looking at the materials being used today, there are examples of nanotechnology making inroads into today's latest systems. Nano-particles are being used to manipulate grain boundaries in both metals and ceramics to tailor performance. Composite materials are utilizing nanotechnology to enhance basic material properties and enhance the system level behaviors to high rate events. While the anticipated revolution never occurred, nanotechnology is beginning to be utilized as an enabler in the latest armor performance improvements.

  20. Physical Model Development and Benchmarking for MHD Flows in Blanket Design

    International Nuclear Information System (INIS)

    An advanced simulation environment to model incompressible MHD flows relevant to blanket conditions in fusion reactors has been developed at HyPerComp in research collaboration with TEXCEL. The goals of this phase-II project are two-fold: The first is the incorporation of crucial physical phenomena such as induced magnetic field modeling, and extending the capabilities beyond fluid flow prediction to model heat transfer with natural convection and mass transfer including tritium transport and permeation. The second is the design of a sequence of benchmark tests to establish code competence for several classes of physical phenomena in isolation as well as in select (termed here as 'canonical',) combinations. No previous attempts to develop such a comprehensive MHD modeling capability exist in the literature, and this study represents essentially uncharted territory. During the course of this Phase-II project, a significant breakthrough was achieved in modeling liquid metal flows at high Hartmann numbers. We developed a unique mathematical technique to accurately compute the fluid flow in complex geometries at extremely high Hartmann numbers (10,000 and greater), thus extending the state of the art of liquid metal MHD modeling relevant to fusion reactors at the present time. These developments have been published in noted international journals. A sequence of theoretical and experimental results was used to verify and validate the results obtained. The code was applied to a complete DCLL module simulation study with promising results.

  1. Physical Model Development and Benchmarking for MHD Flows in Blanket Design

    Energy Technology Data Exchange (ETDEWEB)

    Ramakanth Munipalli; P.-Y.Huang; C.Chandler; C.Rowell; M.-J.Ni; N.Morley; S.Smolentsev; M.Abdou

    2008-06-05

    An advanced simulation environment to model incompressible MHD flows relevant to blanket conditions in fusion reactors has been developed at HyPerComp in research collaboration with TEXCEL. The goals of this phase-II project are two-fold: The first is the incorporation of crucial physical phenomena such as induced magnetic field modeling, and extending the capabilities beyond fluid flow prediction to model heat transfer with natural convection and mass transfer including tritium transport and permeation. The second is the design of a sequence of benchmark tests to establish code competence for several classes of physical phenomena in isolation as well as in select (termed here as “canonical”,) combinations. No previous attempts to develop such a comprehensive MHD modeling capability exist in the literature, and this study represents essentially uncharted territory. During the course of this Phase-II project, a significant breakthrough was achieved in modeling liquid metal flows at high Hartmann numbers. We developed a unique mathematical technique to accurately compute the fluid flow in complex geometries at extremely high Hartmann numbers (10,000 and greater), thus extending the state of the art of liquid metal MHD modeling relevant to fusion reactors at the present time. These developments have been published in noted international journals. A sequence of theoretical and experimental results was used to verify and validate the results obtained. The code was applied to a complete DCLL module simulation study with promising results.

  2. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 2

    International Nuclear Information System (INIS)

    The BOT (Breeder Outside Tube) Helium Cooled Solid Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test blankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of the test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.)

  3. Development of pipe welding, cutting and inspection tools for the ITER blanket

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Ito, Akira; Taguchi, Kou; Takiguchi, Yuji; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-07-01

    In D-T burning reactors such as International Thermonuclear Experimental Reactor (ITER), an internal access welding/cutting of blanket cooling pipe with bend sections is inevitably required because of spatial constraint due to nuclear shield and available port opening space. For this purpose, internal access pipe welding/cutting/inspection tools for manifolds and branch pipes are being developed according to the agreement of the ITER R and D task (T329). A design concept of welding/cutting processing head with a flexible optical fiber has been developed and the basic feasibility studies on welding, cutting and rewelding are performed using stainless steel plate (SS316L). In the same way, a design concept of inspection head with a non-destructive inspection probe (including a leak-testing probe) has been developed and the basic characteristic tests are performed using welded stainless steel pipes. In this report, the details of welding/cutting/inspection heads for manifolds and branch pipes are described, together with the basic experiment results relating to the welding/cutting and inspection. In addition, details of a composite type optical fiber, which can transmit both the high-power YAG laser and visible rays, is described. (author)

  4. Development of pipe welding, cutting and inspection tools for the ITER blanket

    International Nuclear Information System (INIS)

    In D-T burning reactors such as International Thermonuclear Experimental Reactor (ITER), an internal access welding/cutting of blanket cooling pipe with bend sections is inevitably required because of spatial constraint due to nuclear shield and available port opening space. For this purpose, internal access pipe welding/cutting/inspection tools for manifolds and branch pipes are being developed according to the agreement of the ITER R and D task (T329). A design concept of welding/cutting processing head with a flexible optical fiber has been developed and the basic feasibility studies on welding, cutting and rewelding are performed using stainless steel plate (SS316L). In the same way, a design concept of inspection head with a non-destructive inspection probe (including a leak-testing probe) has been developed and the basic characteristic tests are performed using welded stainless steel pipes. In this report, the details of welding/cutting/inspection heads for manifolds and branch pipes are described, together with the basic experiment results relating to the welding/cutting and inspection. In addition, details of a composite type optical fiber, which can transmit both the high-power YAG laser and visible rays, is described. (author)

  5. 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.

  6. Embedded-monolith armor

    Energy Technology Data Exchange (ETDEWEB)

    McElfresh, Michael W.; Groves, Scott E; Moffet, Mitchell L.; Martin, Louis P.

    2016-07-19

    A lightweight armor system utilizing a face section having a multiplicity of monoliths embedded in a matrix supported on low density foam. The face section is supported with a strong stiff backing plate. The backing plate is mounted on a spall plate.

  7. Joining of Tungsten Armor Using Functional Gradients

    International Nuclear Information System (INIS)

    The joining of low thermal expansion armor materials such as tungsten to high thermal expansion heat sink materials has been a major problem in plasma facing component (PFC) development. Conventional planar bonding techniques have been unable to withstand the high thermal induced stresses resulting from fabrication and high heat flux testing. During this investigation, innovative functional gradient joints produced using vacuum plasma spray forming techniques have been developed for joining tungsten armor to copper alloy heat sinks. A model was developed to select the optimum gradient architecture. Based on the modeling effort, a 2mm copper rich gradient was selected. Vacuum plasma pray parameters and procedures were then developed to produce the functional gradient joint. Using these techniques, dual cooling channel, medium scale mockups (32mm wide x 400mm length) were produced with vacuum plasma spray formed tungsten armor. The thickness of the tungsten armor was up to 5mm thick. No evidence of debonding at the interface between the heat sink and the vacuum plasma sprayed material was observed.

  8. AC-130H Gunship Armor Upgrade Project

    Energy Technology Data Exchange (ETDEWEB)

    Shell, T.E.; Landingham, R.L.

    1990-09-19

    This report covers the test methods and equipment for testing aircraft armor both hard and soft. The hard armor are the typical ceramic type while the soft armor are various types of layered composite materials. 10 figs. (JEF)

  9. Survivability Armor Ballistic Laboratory (SABL)

    Data.gov (United States)

    Federal Laboratory Consortium — The SABL provides independent analysis, ballistic testing, data collection, data reduction and qualification of current and advanced armors. Capabilities: The SABL...

  10. Preliminary Study on Melting and Reaction with Liquid Metal Breeders for Developing the Korean Test Blanket Module in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. W.; Yoon, J. S.; Kim, S. K.; Lee, E. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, H. G. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the liquid TBM. In the Korean liquid TBM and breeder blanket, liquid lithium (Li) and lead-lithium (PbLi) are considered as breeders. Related research has been performed: an Experimental Loop for a Liquid breeder (ELLI) constructed to develop an electromagnetic (EM) pump for circulating the liquid breeder, a magnetohydrodynamic (MHD) experiment, and a flow corrosion test. In the ELLI, Pb-15.7Li, where Li is 15.7 at % (called PbLi hereafter), is used as the breeding material. It was purchased from Stachow Metall Company, Germany, and its impurities are shown in Table 1. An EM pump circulates the material in the loop with a maximum flow rate of 60 lpm. The operating pressure and temperature in the loop are 0.4 MPa and 300 .deg. C, respectively, and the maximum operating pressure and temperature are 0.5 MPa and 550 .deg. C Before the loop operation, the melting and solidifying temperatures of the PbLi were measured for ascertaining whether it will show a consistent value for the many cycles of heating and cooling at various conditions of the loop operation. We can also investigate the contamination of PbLi according to the cyclic use. Of the liquid type breeder materials, PbLi is much safer than Li itself, as liquid metal can be ignited when it meets with water or air. There is still a concern regarding the use of PbLi, and it has not been fully proven whether it will react with water or air when it is in a molten state, as it contains lithium. Therefore, reaction tests of Li and PbLi with air and water were performed for safety reasons using the prepared test chamber

  11. Oxide fuel element and blanket element development programs. Quarterly progress report, April-June 1978

    International Nuclear Information System (INIS)

    Approval-in-principle has been granted for run beyond breach experiment XY-2, which will incorporate an F11A series rod. Fuel microstructures and operating parameters have been tabulated for 118 specimens from the F20 power to melt experiment. Retained gas measurements have been compiled indicating 36-50 μl/gm in this high power fuel. Topical report GEFR-00367 was prepared describing F20 results. Preparation of the Test Design Description for axial blanket experiment AB-1 is proceeding on schedule (for Cycle 2 irradiation). The safety analysis calculations, showing no fuel melting nor sodium boiling in design-basis upsets, have been completed

  12. Armored garment for protecting

    Science.gov (United States)

    Purvis, James W.; Jones, II, Jack F.; Whinery, Larry D.; Brazfield, Richard; Lawrie, Catherine; Lawrie, David; Preece, Dale S.

    2009-08-11

    A lightweight, armored protective garment for protecting an arm or leg from blast superheated gases, blast overpressure shock, shrapnel, and spall from a explosive device, such as a Rocket Propelled Grenade (RPG) or a roadside Improvised Explosive Device (IED). The garment has a ballistic sleeve made of a ballistic fabric, such as an aramid fiber (e.g., KEVLAR.RTM.) cloth, that prevents thermal burns from the blast superheated gases, while providing some protection from fragments. Additionally, the garment has two or more rigid armor inserts that cover the upper and lower arm and protect against high-velocity projectiles, shrapnel and spall. The rigid inserts can be made of multiple plies of a carbon/epoxy composite laminate. The combination of 6 layers of KEVLAR.RTM. fabric and 28 plies of carbon/epoxy laminate inserts (with the inserts being sandwiched in-between the KEVLAR.RTM. layers), can meet the level IIIA fragmentation minimum V.sub.50 requirements for the US Interceptor Outer Tactical Vest.

  13. Progress on DCLL Blanket Concept

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Abdou, M.; Katoh, Yutai; Kurtz, Richard J.; Lumsdaine, A.; Marriott, Edward P.; Merrill, Brad; Morley, Neil; Pint, Bruce A.; Sawan, M.; Smolentsev, S.; Williams, Brian; Willms, Scott; Youssef, M.

    2013-09-01

    Under the US Fusion Nuclear Science and Technology Development program, we have selected the Dual Coolant Lead Lithium concept (DCLL) as a reference blanket, which has the potential to be a high performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. The self-cooled breeder PbLi is circulated for power conversion and for tritium breeding. A SiC-based flow channel insert (FCI) is used as a means for magnetohydrodynamic pressure drop reduction from the circulating liquid PbLi and as a thermal insulator to separate the high-temperature PbLi (~700°C) from the helium-cooled RAF/M steel structure. We are making progress on related R&D needs to address critical Fusion Nuclear Science and Facility (FNSF) and DEMO blanket development issues. When performing the function as the Interface Coordinator for the DCLL blanket concept, we had been developing the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. We had estimated the necessary ancillary equipment that will be needed at the ITER site and a detailed safety impact report has been prepared. This provided additional understanding of the DCLL blanket concept in preparation for the FNSF and DEMO. This paper will be a summary report on the progress of the DCLL TBM design and R&Ds for the DCLL blanket concept.

  14. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

  15. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    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

  16. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    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

  17. Development of the Symbolic Manipulator Laboratory modeling package for the kinematic design and optimization of the Future Armor Rearm System robot. Ammunition Logistics Program

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, S.; Jansen, J.F.; Kress, R.L.; Babcock, S.M. [Oak Ridge National Lab., TN (United States); Dubey, R.V. [Tennessee Univ., Knoxville, TN (United States). Dept. of Mechanical and Aerospace Engineering

    1992-08-01

    A new program package, Symbolic Manipulator Laboratory (SML), for the automatic generation of both kinematic and static manipulator models in symbolic form is presented. Critical design parameters may be identified and optimized using symbolic models as shown in the sample application presented for the Future Armor Rearm System (FARS) arm. The computer-aided development of the symbolic models yields equations with reduced numerical complexity. Important considerations have been placed on the closed form solutions simplification and on the user friendly operation. The main emphasis of this research is the development of a methodology which is implemented in a computer program capable of generating symbolic kinematic and static forces models of manipulators. The fact that the models are obtained trigonometrically reduced is among the most significant results of this work and the most difficult to implement. Mathematica, a commercial program that allows symbolic manipulation, is used to implement the program package. SML is written such that the user can change any of the subroutines or create new ones easily. To assist the user, an on-line help has been written to make of SML a user friendly package. Some sample applications are presented. The design and optimization of the 5-degrees-of-freedom (DOF) FARS manipulator using SML is discussed. Finally, the kinematic and static models of two different 7-DOF manipulators are calculated symbolically.

  18. Development of the Symbolic Manipulator Laboratory modeling package for the kinematic design and optimization of the Future Armor Rearm System robot

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, S.; Jansen, J.F.; Kress, R.L.; Babcock, S.M. (Oak Ridge National Lab., TN (United States)); Dubey, R.V. (Tennessee Univ., Knoxville, TN (United States). Dept. of Mechanical and Aerospace Engineering)

    1992-08-01

    A new program package, Symbolic Manipulator Laboratory (SML), for the automatic generation of both kinematic and static manipulator models in symbolic form is presented. Critical design parameters may be identified and optimized using symbolic models as shown in the sample application presented for the Future Armor Rearm System (FARS) arm. The computer-aided development of the symbolic models yields equations with reduced numerical complexity. Important considerations have been placed on the closed form solutions simplification and on the user friendly operation. The main emphasis of this research is the development of a methodology which is implemented in a computer program capable of generating symbolic kinematic and static forces models of manipulators. The fact that the models are obtained trigonometrically reduced is among the most significant results of this work and the most difficult to implement. Mathematica, a commercial program that allows symbolic manipulation, is used to implement the program package. SML is written such that the user can change any of the subroutines or create new ones easily. To assist the user, an on-line help has been written to make of SML a user friendly package. Some sample applications are presented. The design and optimization of the 5-degrees-of-freedom (DOF) FARS manipulator using SML is discussed. Finally, the kinematic and static models of two different 7-DOF manipulators are calculated symbolically.

  19. Systematic methodology for estimating direct capital costs for blanket tritium processing systems

    International Nuclear Information System (INIS)

    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

  20. Blast-Resistant Improvement of Sandwich Armor Structure with Aluminum Foam Composite

    OpenAIRE

    Shu Yang; Chang Qi

    2013-01-01

    Sandwich armor structures with aluminum foam can be utilized to protect a military vehicle from harmful blast load such as a landmine explosion. In this paper, a system-level dynamic finite element model is developed to simulate the blast event and to evaluate the blast-resistant performance of the sandwich armor structure. It is found that a sandwich armor structure with only aluminum foam is capable of mitigating crew injuries under a moderate blast load. However, a severe blast load causes...

  1. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

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

  2. Design study of an armor tile handling manipulator for the Fusion Experimental Reactor

    International Nuclear Information System (INIS)

    A conceptual design of the Fusion Experimental Reactor (FER), which is a D-T burning reactor following on JT-60 in Japan, has been developed by Japan Atomic Energy Research Institute (JAERI). In FER, a rail-mounted vehicle concept is planned to be adopted for in-vessel maintenance, such as maintenance of divertor plates and armor tiles. Advantages of this concept are the high stiffness of the rail as a base structure for maintenance and the high mobility of the vehicle along the rail. Twin armor tile handling manipulators installed on both sides of the vehicle have been designed. The respective manipulators for armor tile handling have 8 degrees of freedom in order to have access to any place of the first wall and to go through the horizontal port by operating manipulator joints. If the two types of manipulators for divertor plates and armor tiles are installed on the vehicle and the divertor handling manipulator carries a case filled with armor tiles, the replacement time of armor tiles will be reduced. In FER, moreover, maintenance of armor tiles, which is a scheduled maintenance, is planned to be carried out by the autonomous control using position sensors etc. In order to accumulate the data base for the development of the autonomous control of the manipulator in armor tile maintenance, the present paper describes basic mechanical characteristics (stress, deflection and natural frequency) of the armor tile handling manipulator calculated by static stress and dynamic eigenvalue analyses. (orig.)

  3. Fast breeder reactor blanket management: comparison of LMFBR and GCFR blankets

    International Nuclear Information System (INIS)

    The economic performance of the fast breeder reactor blanket, considering different fuel management schemes was studied. To perform this, the investigation started with a standard reactor physics calculation. Then, two economic models for evaluation of the economic performance of the radial blanket were developed. These models formed the basis of a computer code, ECOBLAN, which computes the net economic gain and the levelized fuel cost due to the radial blanket. The net gain in terms of dollars and $/kgHM-y and the levelized fuel cost in mills/kWhe were obtained as a function of blanket thickness and a residence time of the fuel in the blanket. A LMFBR and a GCFR were the reactor models considered in this study. The optimum radial blanket of a GCFR consists of two rows, that of a LMFBR consists of three rows. Regarding the different fuel management schemes, the fixed blanket was found to be more favorable than reshuffled blanket. Out-in and in-out reshuffled blanket offer almost the same net gain. In all the cases, the burnup calculated for the fuel was found to be less than the acceptable limit. There is an optimum residence time for the fuel in the blanket which depends on the position of the fuel in the blanket and the fuel management scheme studied. As expected, except for very short residence times (less than 2.5 years), the radial blanket is a net income producer. There is no significant difference between the economic performance of the blanket of a LMFBR and a GCFR

  4. Estimation of Peak Wave Stresses in Slender Complex Concrete Armor Units

    DEFF Research Database (Denmark)

    Howell, G.L.; Burcharth, H. F.; Rhee, Joon R

    1991-01-01

    Recent methods for the structural design of concrete armor units divide the forces into static loads, impact loads, and wave or pulsating loads. Physical model technology is being developed at several laboratories to measure wave loads on model armor units. While this technology represents...

  5. Negligible heat strain in armored vehicle officers wearing personal body armor

    Directory of Open Access Journals (Sweden)

    Hunt Andrew P

    2011-07-01

    Full Text Available Abstract Objectives This study evaluated the heat strain experienced by armored vehicle officers (AVOs wearing personal body armor (PBA in a sub-tropical climate. Methods Twelve male AVOs, aged 35-58 years, undertook an eight hour shift while wearing PBA. Heart rate and core temperature were monitored continuously. Urine specific gravity (USG was measured before and after, and with any urination during the shift. Results Heart rate indicated an intermittent and low-intensity nature of the work. USG revealed six AVOs were dehydrated from pre through post shift, and two others became dehydrated. Core temperature averaged 37.4 ± 0.3°C, with maximum's of 37.7 ± 0.2°C. Conclusions Despite increased age, body mass, and poor hydration practices, and Wet-Bulb Globe Temperatures in excess of 30°C; the intermittent nature and low intensity of the work prevented excessive heat strain from developing.

  6. The state of the art report on the development of manufacturing technology of fusion reactor FW blanket and mock-up in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Jeong, Y. H.; Baek, J. H.; Kim, J. H.; Kim, H. G

    2004-08-15

    The joining technology of first wall blanket has been developed by JAERI in collaboration with Kawasaki Heavy Industry, Isuau Motors and University of Tsukuba in Japan. A variety of joining technologies including HIP, brazing, casing and friction welding was applied to the manufacturing of SS/SS and Cu/SS joint. In Be/Cu joining, it was emphasized to find the optimal HIP temperature lower than 650 .deg. C in order to avoid excessive SS sensitization because the joining of Be tile to Cu heat sink is a final processing step in the manufacturing of FW blanket. The selected HIP condition were 620 .deg. C, 150MPa and 2hr with Cu interlayer. Sample tests for joints was completed by 1995. The small scale mockup was manufactured and its performance was qualified by end of 2000. From 2001, the manufacturing and the characterization has been carried out for the larger scale mockup.

  7. 76 FR 33746 - Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Liquefied Natural...

    Science.gov (United States)

    2011-06-09

    ..., which granted the Dow Chemical Company blanket authorization to export up to an amount equivalent to 390....'' \\7\\ \\7\\ The Dow Chemical Company, DOE/FE Order No. 2859, issued October 5, 2010. Additionally...) Texas LNG Holdings, LLC, a Delaware limited liability company and wholly-owned subsidiary of The...

  8. Low activity aluminum blanket

    Energy Technology Data Exchange (ETDEWEB)

    Benenati, R.; Tichler, P.; Powell, J.R.

    1976-03-01

    The basic design of the breeding blanket consists of cylindrical aluminium canisters filled with a ceramic bed of moderating, shielding, and breeding materials all suitably cooled. A technical analysis of the blanket for an EPR design is given. Activation studies are presented. The effect of pulsed magnetic fields on module structure is investigated. (MOW)

  9. Current status of technology development for fabrication of Indian Test Blanket Module (TBM) of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, T., E-mail: tjk@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam 603102 (India); Rajendra Kumar, E. [TBM Division, Institute for Plasma Research (IPR), Bhat, Gandhinagar 382428 (India)

    2014-10-15

    Highlights: • Status of technology developments for Indian TBM to be installed in ITER is presented. • Procedure development for EB, laser and laser-hybrid welding of RAFM steel presented. • Filler wires for RAFM steel for TIG, NG-TIG and laser-hybrid welding have been developed. • Feasibility of production of channel plate by HIP technology has been demonstrated. - Abstract: Ever since India decided to install its Lead-Lithium Ceramic Breeder (LLCB) TBM in ITER, various technologies for fabrication of Indian TBM are being pursued by IPR and IGCAR, in collaboration with various research laboratories in India. Welding consumables for joining India specific RAFM steels (IN-RAFMS), procedures for hot isostatic pressing, electron beam welding, laser and laser-hybrid welding have been developed. Considering the complex nature and limited access available for inspection, innovative inspection procedures that involved use of phased array ultrasonic and C-scan imaging are also being pursued. This paper presents the current status of these developments and provides a roadmap for the future activities planned in realizing Indian TBM for testing in ITER.

  10. COMPUTER SIMULATION OF PENETRATION IN MATERIALS IN SOFT ARMOR PLATES

    OpenAIRE

    Parga Landa, B.; Hernández Olivares, F.

    1988-01-01

    This paper presents a newly developed computer code able to analyze the penetration mechanics into Composite Materials in soft armor plates. It can be used to calculate the ballistic curve of Composite Material soft armors as well as the impact force, the tension in each layer, displacements and velocities of the plate and projectile, stresses and strains in the plate and the damaged area. The program is based on the moment conservation law. The fibres are assumed to be linear elastic up to f...

  11. Effectiveness of eye armor during blast loading.

    Science.gov (United States)

    Bailoor, Shantanu; Bhardwaj, Rajneesh; Nguyen, Thao D

    2015-11-01

    experienced by the unprotected eye after 0.2 ms of impact of blast wave, for lower as well as higher charge mass. The present model provides fundamental insights of flow and pressure fields in the ocular region, which helps to explain the effectiveness of the eye armor. Since the measurements of these fields are not trivial, the computational model aids in better understanding of development of PBI.

  12. Comparison of coating processes in the development of aluminum-based barriers for blanket applications

    Energy Technology Data Exchange (ETDEWEB)

    Wulf, Sven-Erik, E-mail: sven-erik.wulf@kit.edu; Krauss, Wolfgang; Konys, Jürgen

    2014-10-15

    Highlights: •Electrochemical processes ECA and ECX are suitable for Al deposition on RAFM steels. •ECA and ECX are able to produce thin Al layers with adjustable thicknesses. •All aluminization processes need a subsequent heat treatment. •Scales made by ECA or ECX exhibit reduced thicknesses compared to HDA. •ECX provides higher flexibility compared to ECA to produce scales on RAFM steels. -- Abstract: Reduced activation ferritic-martensitic steels (RAFM), e.g. Eurofer 97, are envisaged in future fusion technology as structural material, which will be in direct contact with a flowing liquid lead–lithium melt serving as breeder material. Aluminum-based barrier layers had proven their ability to protect the structural material from corrosion attack in flowing Pb–15.7Li and to reduce tritium permeation into the coolant. Coming from scales produced by hot dipping aluminization (HDA), the development of processes based on electrochemical methods to produce defined aluminum-based scales on RAFM steels gained attention in research during the last years. Two different electrochemical processes are proposed: The first one, referred to as ECA process, is based on the electrodeposition of aluminum from volatile, metal-organic electrolytes. The other process called ECX is based on ionic liquids. All three processes exhibit specific characteristics, for example in the field of processability, control of coating thicknesses (low activation criteria) and heat treatment behavior. The aim of this article is to compare these different coating processes critically, whereby the focus is on the comparison of ECA and ECX processes. New results for ECX-process will be presented and occurring development needs for the future will be discussed.

  13. Success Story of Armored Multitask Vehicle Design

    OpenAIRE

    ALANKAYA, Veysel

    2015-01-01

    In this study; an armored vehicle design including whole process from preliminary studies to armor design is presented in order of defining requirements depending on the multitasking performance, selecting armor threats for different parts of the fuselage, preliminary design stages for different arrangements, procurement phases and performance tests. Ballistic evaluation of main areas depending on the vulnerability capabilities are studied by means of no add-on protection for personnel is nec...

  14. Divertor and gas blanket impurity control study

    Energy Technology Data Exchange (ETDEWEB)

    El Derini, Z; Stacey, Jr, W M

    1979-04-01

    A simple calculational model for the transport of particles across the scrap off region between the plasma and the wall in the presence of a divertor or a gas blanket has been developed. The model departs from previous work in including: (a) the entire impurity transport as well as its effect on the energy balance equations; (b) the recycling neutrals from the divertor, and (c) the reflected neutrals from the wall. Results obtained with this model show how the steady state impurity level in the plasma depends on the divertor parameters such as the neutral backflow from the divertor, the particle residence time and the scrape off thickness; and on the gas blanket parameters such as the neutral source strength and the gas blanket thickness. The variation of the divertor or gas blanket performance as a function of the heat and particle fluxes escaping from the plasma, the wall material and the cross field diffusion is examined and numerical examples are given.

  15. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this project advanced acoustic blankets for improved low frequency interior noise control in aircraft will be developed and demonstrated. The improved...

  16. River-bed armoring as a granular segregation phenomenon

    CERN Document Server

    Ferdowsi, Behrooz; Houssais, Morgane; Jerolmack, Douglas J

    2016-01-01

    Gravel-river beds typically have an "armored" layer of coarse grains on the surface, which acts to protect finer particles underneath from erosion. River bed-load transport is a kind of dense granular flow, and such flows are known to vertically segregate grains. The contribution of granular physics to river-bed armoring, however, has not been investigated. Here we examine these connections in a laboratory river with bimodal sediment size, by tracking the motion of particles from the surface to deep inside the bed, and find that armor develops by two distinct mechanisms. Bed-load transport in the near-surface layer drives rapid segregation, with a vertical advection rate proportional to the granular shear rate. Creeping grains beneath the bed-load layer give rise to slow but persistent segregation, which is diffusion dominated and insensitive to shear rate. We verify these findings with a continuum phenomenological model and discrete element method simulations. Our results suggest that river beds armor by gra...

  17. Development of a control system for a heavy object handling manipulator. Application to a remote maintenance system for ITER blanket module

    International Nuclear Information System (INIS)

    This paper describes a control system for the heavy object handling manipulator. It has been developed for the blanket module remote maintenance system of ITER (International Thermonuclear Fusion Experimental Reactor). A rail-mounted vehicle-type manipulator is proposed for the precise handling of a blanket module which is about 4 tons in weight. Basically, this manipulator is controlled by teaching-playback technique. When grasping or releasing the module, the manipulator sags and the position of the end-effector changes about 50 [mm]. Applying only the usual teaching-playback control makes the smooth operation of setting/removing modules to/from the vacuum vessel wall difficult due to this position change. To solve this proper problem of heavy object handling manipulator, we have developed a system which uses motion patterns generated from two kinds of teaching points. These motion patterns for setting/removing heavy objects are generated by combining teaching points for positioning the manipulator with and without grasping the object. When these motion patterns are applied, the manipulator can transfer the object's weight smoothly at the setting/removing point. This developed system has been applied to the real-scale mock-up of the vehicle manipulator and through the actual module setting/removing experiments, we have verified its effectiveness and realized smooth maintenance operation. (author)

  18. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    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

  19. Development of the driving panel and control system in armored vehicle%装甲车辆驾驶员仪表和指令系统的发展方向

    Institute of Scientific and Technical Information of China (English)

    刘北生; 刘勇; 梁瑞雪

    2001-01-01

    该文用人一机一环境系统工程的理论和方法,分析了坦克装甲车辆仪表和指令系统发展的特点和规律。指出多媒体、智能化是未来坦克装甲车辆仪表和指令系统的发展方向。%In this paper the development of the panel and control system in tank and armored vehicle is ana1ysised with the theory and method of man-machine-environment system engineering. It points out that the system ' s developing trendency is multimedia and intellectualized.

  20. Development of Blanket Using Milk Protein Composite Fibers%牛奶蛋白复合纤维毛毯的开发

    Institute of Scientific and Technical Information of China (English)

    厉巽巽; 胡忠娟; 陈玉霜; 陈金霞

    2015-01-01

    以牛奶蛋白复合纤维为毛绒纱,涤纶纤维为底布,制备出牛奶蛋白复合纤维毛毯,详细阐述该毛毯加工过程中的关键技术,包括整经、编织、预定形、剖绒、染色、柔软整理和烘干定形。最后对成品的面料参数及性能进行了测试与表征。结果表明:使用牛奶蛋白复合纤维和涤纶可以开发出牛奶蛋白复合纤维毛毯,并且达到了色泽鲜艳、保暖等效果。%The paper develops a new kind of blanket by using milk protein composite fibers as plush and polyester fiber as foundation fabric, and also introduces in detail the key technologies including beaming, knitting, pre-setting, cutting, dyeing, soft finishing and drying setting, then tests and characterizes the finished products’ parameters and performance. The results show that the developed blanket possesses the expected properties of bright color and heat retention by using milk protein composite fibers and polyester fiber.

  1. Ballistic performance and microstructure of four armor ceramics

    NARCIS (Netherlands)

    Abadjieva, E.; Carton, E.P.

    2013-01-01

    The ballistic behavior of four different armor ceramic materials with thicknesses varying from 3 mm to 14 mm has been investigated. These are two types of alumina Al2O3 armor grades and two types of SiC armor grades produced by different armor ceramic producers. The ballistic study has been performe

  2. The effects of posture, body armor, and other equipment on rifleman lethality

    OpenAIRE

    Kramlich, Gary R.

    2005-01-01

    How does body armor and posture affect Soldier marksmanship? The Interceptor Body Armor (IBA) has significantly improved Soldier combat survivability, but in what ways does it change rifleman lethality? Moreover, can we model these effects so as to develop better tactics and operational plans? This study quantifies the effects of Soldier equipment on lethality through multi-factor logistic regression using data from range experiments with the 1st Brigade, 1st Infantry Division (Mechanized), a...

  3. Single-Layer Cubipod Armored Breakwaters in Punta Langosteira (Spain)

    OpenAIRE

    Corredor, Antonio; Santos, Moisés; Peña, Enrique; Maciñeira, Enrique; GÓMEZ MARTÍN, MARÍA ESTHER; Medina, Josep R.

    2014-01-01

    This paper describes the design process, hydraulic stability tests and construction of two single-layer Cubipod armored breakwaters in the Port of Punta Langosteira (A Coruña, Spain), located on the Atlantic coast of Spain, the first single-layer armors of randomly placed massive concrete armor units. The environmental, geotechnical, economic and logistic conditions favored randomly-placed Cubipods in single-layer armoring. 3D hydraulic stability tests of single-layer Cubipod armored breakwat...

  4. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    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.)

  5. Blast-Resistant Improvement of Sandwich Armor Structure with Aluminum Foam Composite

    Directory of Open Access Journals (Sweden)

    Shu Yang

    2013-01-01

    Full Text Available Sandwich armor structures with aluminum foam can be utilized to protect a military vehicle from harmful blast load such as a landmine explosion. In this paper, a system-level dynamic finite element model is developed to simulate the blast event and to evaluate the blast-resistant performance of the sandwich armor structure. It is found that a sandwich armor structure with only aluminum foam is capable of mitigating crew injuries under a moderate blast load. However, a severe blast load causes force enhancement and results in much worse crew injury. An isolating layer between the aluminum foam and the vehicle floor is introduced to remediate this drawback. The results show that the blast-resistant capability of the innovative sandwich armor structure with the isolating layer increases remarkably.

  6. High power density self-cooled lithium-vanadium blanket.

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Majumdar, S.; Smith, D.

    1999-07-01

    A self-cooled lithium-vanadium blanket concept capable of operating with 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading has been developed. The blanket has liquid lithium as the tritium breeder and the coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because it can accommodate high heat loads. Also, it has good mechanical properties at high temperatures, high neutron fluence capability, low degradation under neutron irradiation, good compatibility with the blanket materials, low decay heat, low waste disposal rating, and adequate strength to accommodate the electromagnetic loads during plasma disruption events. Self-healing electrical insulator (CaO) is utilized to reduce the MHD pressure drop. A poloidal coolant flow with high velocity at the first wall is used to reduce the peak temperature of the vanadium structure and to accommodate high surface heat flux. The blanket has a simple blanket configuration and low coolant pressure to reduce the fabrication cost, to improve the blanket reliability, and to increase confidence in the blanket performance. Spectral shifter, moderator, and reflector are utilized to improve the blanket shielding capability and energy multiplication, and to reduce the radial blanket thickness. Natural lithium is used to avoid extra cost related to the lithium enrichment process.

  7. Lightweight IMM PV Flexible Blanket Assembly

    Science.gov (United States)

    Spence, Brian

    2015-01-01

    Deployable Space Systems (DSS) has developed an inverted metamorphic multijunction (IMM) photovoltaic (PV) integrated modular blanket assembly (IMBA) that can be rolled or z-folded. This IMM PV IMBA technology enables a revolutionary flexible PV blanket assembly that provides high specific power, exceptional stowed packaging efficiency, and high-voltage operation capability. DSS's technology also accommodates standard third-generation triple junction (ZTJ) PV device technologies to provide significantly improved performance over the current state of the art. This SBIR project demonstrated prototype, flight-like IMM PV IMBA panel assemblies specifically developed, designed, and optimized for NASA's high-voltage solar array missions.

  8. Boron Rich Solids Sensors, Ultra High Temperature Ceramics, Thermoelectrics, Armor

    CERN Document Server

    Orlovskaya, Nina

    2011-01-01

    The objective of this book is to discuss the current status of research and development of boron-rich solids as sensors, ultra-high temperature ceramics, thermoelectrics, and armor. Novel biological and chemical sensors made of stiff and light-weight boron-rich solids are very exciting and efficient for applications in medical diagnoses, environmental surveillance and the detection of pathogen and biological/chemical terrorism agents. Ultra-high temperature ceramic composites exhibit excellent oxidation and corrosion resistance for hypersonic vehicle applications. Boron-rich solids are also promising candidates for high-temperature thermoelectric conversion. Armor is another very important application of boron-rich solids, since most of them exhibit very high hardness, which makes them perfect candidates with high resistance to ballistic impact. The following topical areas are presented: •boron-rich solids: science and technology; •synthesis and sintering strategies of boron rich solids; •microcantileve...

  9. A new world find of european scale armor

    Directory of Open Access Journals (Sweden)

    Rogers, Hugh C.

    1999-12-01

    Full Text Available A group of armor scales found in New Mexico (USA is critically examined from an archeological and historical point of view. The rarity of the find and its importance are demostrated, both for the history of Spanish exploration in New Mexico and for the development of scale armor in Europe.

    Estudio crítico, desde un punto de vista histórico y arqueológico, de un grupo de placas de armadura hallados en Nuevo Méjico (USA. La rareza del hallazgo es manifiesta para la historia de las expediciones españolas en Nuevo Méjico y para la evolución de las armaduras de placas en Europa.

  10. Armoring and vertical sorting in aeolian dune fields

    Science.gov (United States)

    Gao, Xin; Narteau, Clément; Rozier, Olivier

    2016-04-01

    Unlike ripples, there are only few numerical studies on grain-size segregation at the scale of dunes in aeolian environments. Here we use a cellular automaton model to analyze vertical sorting in granular mixtures under steady unidirectional flow conditions. We investigate the feedbacks between dune growth and the segregation mechanisms by varying the size of coarse grains and their proportion within the bed. We systematically observe the development of a horizontal layer of coarse grains at the top of which sorted bed forms may grow by amalgamation. The formation of such an armor layer controls the overall sediment transport and availability. The emergence of dunes and the transition from barchan to transverse dune fields depend only on the grain size distribution of the initial sediment layer. As confirmed by observation, this result indicates that armor layers should be present in most arid deserts, where they are likely to control dune morphodynamics.

  11. 75 FR 51482 - Woven Electric Blankets From China

    Science.gov (United States)

    2010-08-20

    ... publishing the notice in the Federal Register of March 11, 2010 (75 FR 11557). The hearing was held in... COMMISSION Woven Electric Blankets From China Determination On the basis of the record \\1\\ developed in the... United States is materially injured by reason of imports from China of woven electric blankets,...

  12. Ballistic protection performance of curved armor systems with or without debondings/delaminations

    International Nuclear Information System (INIS)

    Highlights: • Influence of pre-existing defect in an armor system on its ballistic performance. • Development of finite element models for the ballistic performance of armor systems. • Prediction of the ballistic limit and back face deformation of curved armor systems. - Abstract: In order to discern how pre-existing defects such as single or multiple debondings/delaminations in a curved armor system may affect its ballistic protection performance, two-dimensional axial finite element models were generated using the commercial software ANSYS/Autodyn. The armor systems considered in this investigation are composed of boron carbide front component and Kevlar/epoxy backing component. They are assumed to be perfectly bonded at the interface without defects. The parametric study shows that for the cases considered, the maximum back face deformation of a curved armor system with or without defects is more sensitive to its curvature, material properties of the ceramic front component, and pre-existing defect size and location than the ballistic limit velocity. Additionally, both the ballistic limit velocity and maximum back face deformation are significantly affected by the backing component thickness, front/backing component thickness ratio and the number of delaminations

  13. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    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)

  14. Neutronic analysis of a dual He/LiPb coolant breeding blanket for DEMO

    OpenAIRE

    Catalán, J.P.; Ogando Serrano, Francisco; Sanz Gonzalo, Javier; Palermo, I.; Veredas, G.; Gómez Ros, J. M.; Sedano, L

    2010-01-01

    A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO_FUS based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils ...

  15. ITER blanket manifold system: Integration, assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Alex, E-mail: alex.martin@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Dellopoulos, George [F4E, EU ITER Domestic Agency, Barcelona (Spain); Edwards, Paul; Furmanek, Andreas; Gicquel, Stefan; Macklin, Brian [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Martin, Patrick [RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Merola, Mario; Norman, Mark; Raffray, Rene [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    Highlights: •The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. •-The blanket manifold system has been redesigned to improve leak detection and localization. •-The redesign of the blanket manifold system into a system based on individual pipes has proven to be a major engineering challenge. -- Abstract: The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned.

  16. An assessment of the base blanket for ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-31

    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.

  17. An assessment of the base blanket for ITER

    Energy Technology Data Exchange (ETDEWEB)

    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. ITER blanket manifold system: Integration, assembly and maintenance

    International Nuclear Information System (INIS)

    Highlights: •The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. •-The blanket manifold system has been redesigned to improve leak detection and localization. •-The redesign of the blanket manifold system into a system based on individual pipes has proven to be a major engineering challenge. -- Abstract: The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned

  19. LMFBR Blanket Physics Project progress report No. 6

    International Nuclear Information System (INIS)

    Progress is summarized in experimental and analytical investigations of the neutronics and photonics of benchmark mockups of LMFBR blankets. During the reporting period work was devoted primarily to a wide range of analytical/numerical investigations, including blanket fuel management/economics studies, evaluation of improved blanket designs, and assessment of state-of-the-art methods for gamma heating calculations. Experimental work included preparations for resumption of MIT Reactor operations, primarily fabrication of improved steel reflector assemblies for blanket mockups, and development of an improved radiophotoluminescent readout device for LiF thermoluminescent detectors. The most significant finding was that the neutronic and economic performance of radial blanket assemblies are essentially independent of core size (rating) for radially-power-flattened cores. Hence the methodology and results of current experiments and calculations should be valid for the large commercial LMFBR's of the future

  20. LMFBR Blanket Physics Project progress report No. 6

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J. (ed.)

    1975-06-30

    Progress is summarized in experimental and analytical investigations of the neutronics and photonics of benchmark mockups of LMFBR blankets. During the reporting period work was devoted primarily to a wide range of analytical/numerical investigations, including blanket fuel management/economics studies, evaluation of improved blanket designs, and assessment of state-of-the-art methods for gamma heating calculations. Experimental work included preparations for resumption of MIT Reactor operations, primarily fabrication of improved steel reflector assemblies for blanket mockups, and development of an improved radiophotoluminescent readout device for LiF thermoluminescent detectors. The most significant finding was that the neutronic and economic performance of radial blanket assemblies are essentially independent of core size (rating) for radially-power-flattened cores. Hence the methodology and results of current experiments and calculations should be valid for the large commercial LMFBR's of the future.

  1. Advanced Polymer For Multilayer Insulating Blankets

    Science.gov (United States)

    Haghighat, R. Ross; Shepp, Allan

    1996-01-01

    Polymer resisting degradation by monatomic oxygen undergoing commercial development under trade name "Aorimide" ("atomic-oxygen-resistant imidazole"). Intended for use in thermal blankets for spacecraft in low orbit, useful on Earth in outdoor applications in which sunlight and ozone degrades other plastics. Also used, for example, to make threads and to make films coated with metals for reflectivity.

  2. Fidget Blankets: A Sensory Stimulation Outreach Program.

    Science.gov (United States)

    Kroustos, Kelly Reilly; Trautwein, Heidi; Kerns, Rachel; Sobota, Kristen Finley

    2016-01-01

    Behavioral and Psychological Symptoms of Dementia (BPSD) include behaviors such as aberrant motor behavior, agitation, anxiety, apathy, delusions, depression, disinhibition, elation, hallucinations, irritability, and sleep or appetite changes. A student-led project to provide sensory stimulation in the form of "fidget blankets" developed into a community outreach program. The goal was to decrease the use of antipsychotics used for BPSD.

  3. Water-cooled blanket concepts for the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    The primary goal of the Blanket Comparison and Selection Study (BCSS) was to select a limited number of blanket concepts for fusion power reactors, to serve as the focus for the U.S. Department of Energy blanket research and development program. The concepts considered most seriously by the BCSS can be grouped for discussion purposes by coolant: liquid metals and alloys, pressurized water, helium, and nitrate salts. Concepts using pressurized water as the coolant are discussed. Water-cooled concepts using liquid breeders-lithium and 17Li-83Pb (LiPb)-have severe fundamental safety problems. The use of lithium and water in the blanket was considered unacceptable. Initial results of tests at Hanford Engineering Development Laboratory using steam injected into molten LiPb indicate that use of LiPb and water together in a blanket is a very serious concern from the safety standpoint. Key issues for water-cooled blankets with solid tritium breeders (Li2O, or a ternary oxide such as LiAlO2) were identified and examined: reliability against leaks, control of tritium permeation into the coolant, retention of breeder physical integrity, breeder temperature predictability, determination of allowable temperature limits for breeders, and 6Li burnup effects (for LiAlO2). The BCSS's final rankings and associated rationale for all water-cooled concepts are examined. Key issues and factors for tokamak and tandem mirror reactor versions of water-cooled solid breeder concepts are discussed. The reference design for the top-ranked concept-LiAlO2 breeder, ferritic steel structure, and beryllium neutron multiplier-is presented. Finally, some general conclusions for water-cooled blanket concepts are drawn based on the study's results

  4. Dynamic chamber armor behavior in IFE and MFE

    International Nuclear Information System (INIS)

    The chamber wall armor is subject to demanding conditions in both inertial fusion energy (IFE) and magnetic fusion energy (MFE) chambers. This paper assesses the requirements on armor imposed by the operating conditions in IFE and MFE, including energy deposition density, time of deposition and frequencies, and discusses their impact on the performance of the candidate armor materials

  5. Resonance and streaming of armored microbubbles

    Science.gov (United States)

    Spelman, Tamsin; Bertin, Nicolas; Stephen, Olivier; Marmottant, Philippe; Lauga, Eric

    2015-11-01

    A new experimental technique involves building a hollow capsule which partially encompasses a microbubble, creating an ``armored microbubble'' with long lifespan. Under acoustic actuation, such bubble produces net streaming flows. In order to theoretically model the induced flow, we first extend classical models of free bubbles to describe the streaming flow around a spherical body for any known axisymmetric shape oscillation. A potential flow model is then employed to determine the resonance modes of the armored microbubble. We finally use a more detailed viscous model to calculate the surface shape oscillations at the experimental driving frequency, and from this we predict the generated streaming flows.

  6. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    The blanket of the present invention can keep the temperature of breeding materials within a predetermined range even if the breeding materials are consumed and the amount of heat generated from the breeding materials is reduced, thereby enabling to release tritium stably. That is, a neutron incident amount control means is disposed to the blanket for controlling the amount of neutrons incident to the breeding materials. Alternatively, a material to form hollow layers are disposed to the periphery of the breeding materials. With such constitution, the neutron incident amount control means enables to control the incident amount of neutrons from plasmas to the breeding materials, thereby enabling to suppress the change of the amount of heat generated in the breeding materials. In addition, the hollow layers formed at the periphery of the breeding materials enables selective filling of fluids having different heat transfer characteristics thereby enabling to control heat resistance between the breeding materials and cooling tubes. Accordingly, temperature of the breeding materials can be kept constant even in any of the cases. (I.S.)

  7. Formation of Temperature Mode in Internal Spaces of Armored Machinery

    Directory of Open Access Journals (Sweden)

    P. I. Dziachek

    2010-01-01

    Full Text Available An algorithm and a program  have been developed for calculation of temperature mode in a T-72 tank tower while using numerical methods. The paper reveals a satisfactory convergence of calculated and experimental data. The results pertaining to investigation of influence of various factors on the processes of temperature mode formation in the internal surfaces of the studied object are presented in the paper. The obtained results make it possible to develop technical solutions in respect of protection against condensation of water vapor on the internal surfaces of armored machinery.

  8. EMBEDDED FIBER OPTIC SENSORS FOR INTEGRAL ARMOR

    Science.gov (United States)

    This report describes the work performed with Production Products Manufacturing & Sales (PPMS), Inc., under the "Liquid Molded Composite Armor Smart Structures Using Embedded Sensors" Small Business Innovative Research (SBlR) Program sponsored by the U.S. Army Research Laboratory...

  9. Doublet III vacuum vessel neutral beam armor

    International Nuclear Information System (INIS)

    The evolution of the Doublet III neutral beam armor is followed from the initial design of a radiation cooled metallic tile to the present actively cooled graphite design. Results of the thermal and stress analyses that dictated the present design are reviewed

  10. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the proposed Phase II research effort is to develop heterogeneous (HG) blankets for improved sound reduction in aircraft structures. Phase I...

  11. Simulation of armor penetration by tungsten rods: ALEGRA validation report

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, D.E.; Hertel, E.S. Jr.; Trucano, T.G.

    1997-11-01

    Results from simulations of the impact and penetration of tungsten alloy rods into thick rolled armor plates are presented. The calculations were performed with the CTH and ALEGRA computer codes using the DOE massively parallel TFLOPS computer co-developed by Sandia National Laboratory and Intel Corporation. Comparisons with experimental results are presented. Agreement of the two codes with each other and with the empirical results for penetration channel depth and radius is very close. Other shock physics and penetration features are also compared to simulation results.

  12. TBM testing in ITER: Requirements for the development of predictive tools to describe corrosion-related phenomena in HCLL blankets towards DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Krauss, W., E-mail: wolfgang.krauss@kit.edu [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Konys, J. [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Li-Puma, A. [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer To collect corrosion data relevant for TBM operation in updated facilities. Black-Right-Pointing-Pointer To develop predictive tools (corrosion, transport, precipitation, impurity effects). Black-Right-Pointing-Pointer To perform validation of predictive tools. Black-Right-Pointing-Pointer To develop/qualify components for TBM tests in ITER. - Abstract: Compatibility testing of RAFM-steels in Pb-15.7Li environment has shown that liquid metal corrosion is always present and dissolution of steel elements in hot areas of non-isothermal systems takes place whereas a transport of the corrosion products and formed precipitates has to be considered in the TBM design. It is clear that for the design of a HCLL breeding blanket system for DEMO and to ensure the safety over a fusion power plant lifetime, a good knowledge of the corrosion behavior including the dominating mechanisms is required. Simulation tools predicting the corrosion behavior of bare and coated Eurofer in Pb-15.7Li must be implemented and validated in a real fusion environment where numerous physical phenomena are additionally present, compared to the state of the art corrosion knowledge, such as neutron flux, H, T permeation, MHD effects, temperature field with steep gradients. The state of the art will be shown and discussed using some of the main fundamental corrosion data selected from own testing campaigns and published literature regarding corrosion behavior of TBMs. On this basis a test matrix for TBM testing in ITER is presented in the paper and the deficits in present knowledge are outlined deviating future development needs in corrosion.

  13. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    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

  14. Blanket comparison and selection study. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    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. (MOW)

  15. Enhanced stab resistance of armor composites with functionalized silica nanoparticles

    Science.gov (United States)

    Mahfuz, Hassan; Clements, Floria; Rangari, Vijaya; Dhanak, Vinod; Beamson, Graham

    2009-03-01

    Traditionally shear thickening fluid (STF) reinforced with Kevlar has been used to develop flexible armor. At the core of the STF-Kevlar composites is a mixture of polyethylene glycol (PEG) and silica particles. This mixture is often known as STF and is consisted of approximately 45 wt % PEG and 55 wt % silica. During rheological tests, STF shows instantaneous spike in viscosity above a critical shear rate. Fabrication of STF-Kevlar composites requires preparation of STF, dilution with ethanol, and then impregnation with Kevlar. In the current approach, nanoscale silica particles were dispersed directly into a mixture of PEG and ethanol through a sonic cavitation process. Two types of silica nanoparticles were used in the investigation: 30 nm crystalline silica and 7 nm amorphous silica. The admixture was then reinforced with Kevlar fabric to produce flexible armor composites. In the next step, silica particles are functionalized with a silane coupling agent to enhance bonding between silica and PEG. The performance of the resulting armor composites improved significantly. As evidenced by National Institute of Justice spike tests, the energy required for zero-layer penetration (i.e., no penetration) jumped twofold: from 12 to 25 J cm2/g. The source of this improvement has been traced to the formation of siloxane (Si-O-Si) bonds between silica and PEG and superior coating of Kevlar filaments with particles. Fourier transform infrared, x-ray photoemission spectroscopy, and scanning electron microscopy studies were performed to examine chemical bonds, elemental composition, and particle dispersion responsible for such improvement. In summary, our experiments have demonstrated that functionalization of silica particles followed by direct dispersion into PEG resulted in superior Kevlar composites having much higher spike resistance.

  16. Ceramic/polymer functionally graded material (FGM) lightweight armor system

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, J.J.; McClellan, K.J.

    1998-12-31

    This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). Functionally graded material is an enabling technology for lightweight body armor improvements. The objective was to demonstrate the ability to produce functionally graded ceramic-polymer and ceramic-metal lightweight armor materials. This objective involved two aspects. The first and key aspect was the development of graded-porosity boron-carbide ceramic microstructures. The second aspect was the development of techniques for liquid infiltration of lightweight metals and polymers into the graded-porosity ceramic. The authors were successful in synthesizing boron-carbide ceramic microstructures with graded porosity. These graded-porosity boron-carbide hot-pressed pieces were then successfully liquid-infiltrated in vacuum with molten aluminum at 1,300 C, and with liquid polymers at room temperature. Thus, they were able to demonstrate the feasibility of producing boron carbide-aluminum and boron carbide-polymer functionally graded materials.

  17. Safety and personnel access aspects of low activation fusion blankets

    International Nuclear Information System (INIS)

    The use of silicon carbide and carbon materials for structural applications in fusion reactor first wall and blanket regions has been proposed and a continuing effort spent on the development of the ceramics technology. The advantages identified are an extremely low induced radioactivity inventory, a high temperature operating capability, abundant raw material resource availability, and minimized plasma impurity effects. One of the unique features of the applications of these materials to fusion reactor blanket designs is that no alloying element is needed in order to assure the specified mechanical properties such as occurs in metal alloys. The major source of long term radioactivity in these materials is impurities. The impurity elements and their concentrations carried over to the blanket structure during fabrication can be minimized by proper fabrication procedures and techniques. The safety and personnel access aspects of such fusion blankets in conjunction with the impurity element concentration are the main subjects of this paper

  18. Dynamic Behavior of High Strength Armor Steels

    OpenAIRE

    Nahme, H.; Lach, E.

    1997-01-01

    The dynamic properties of the armor steels Mars 190, Mars 240 and Mars 300 have been determined for strain rates 10-3s-1 104s-1). The dynamic properties Hugoniot-elastic-limit, spall strength, shock velocity-particle verlocity-relation and stress-strain-relation will be presented. Tests at intermediate strain rates 102 s-1 < dε/dt < n 103 have been performed usin...

  19. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    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)

  20. Numerical simulation of armored vehicles subjected to undercarriage landmine blasts

    Science.gov (United States)

    Erdik, A.; Kilic, S. A.; Kilic, N.; Bedir, S.

    2015-05-01

    Landmine threats play a crucial role in the design of armored personnel carriers. Therefore, a reliable blast simulation methodology is valuable to the vehicle design development process. The first part of this study presents a parametric approach for the quantification of the important factors such as the incident overpressure, the reflected overpressure, the incident impulse, and the reflected impulse for the blast simulations that employ the Arbitrary Lagrangian-Eulerian formulation. The effects of mesh resolution, mesh topology, and fluid-structure interaction (FSI) parameters are discussed. The simulation results are compared with the calculations of the more established CONventional WEaPons (CONWEP ) approach based on the available experimental data. The initial findings show that the spherical topology provides advantages over the Cartesian mesh domains. Furthermore, the FSI parameters play an important role when coarse Lagrangian finite elements are coupled with fine Eulerian elements at the interface. The optimum mesh topology and the mesh resolution of the parametric study are then used in the landmine blast simulation. The second part of the study presents the experimental blast response of an armored vehicle subjected to a landmine explosion under the front left wheel in accordance with the NATO AEP-55 Standard. The results of the simulations show good agreement with the experimental measurements.

  1. Axial blanket for 16NGF Angra 1 fuel type

    Energy Technology Data Exchange (ETDEWEB)

    Sadde, Luciano Martins; Faria, Eduardo Fernandes [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil)]. E-mails: sadde@inb.gov.br; faria@inb.gov.br; Sang-Keun You [Korea Nuclear Fuel Co. Ltd. (KNFC), Taejon (Korea, Republic of)]. E-mail: skyou@knfc.co.kr

    2007-07-01

    Angra-1, Kori-2 and Krsko are nuclear power plants with the same design. However, the fuel assemblies have some differences in design due to the countries strategies and the differences in the fabrication process. The 16NGF (16x16 Next Generation Fuel) was developed by INB, KNFC and Westinghouse in order to be used in these three nuclear power plants and the 'Axial Blanket' is one of the new features for the 16NGF design. The main purpose of the Axial Blanket Optimization study is to determine which axial blanket enrichment and length would provide the better fuel cycle cost benefit. All of the calculations were performed using Gadolinium as Burnable Absorber and solid pellets type for Axial Blanket. The results indicate 1.8 w/o U235 enrichment and 8 inches length as the best option of Axial Blanket from the fuel cycle cost benefit standpoint. The economy is about 1.8%. The difference in the reload cost in the range between 1.5 and 2.6 w/o U235 enrichment and for the 6 and 8 inches length is not so significant. Due that, from the Fq limit standpoint and also for longer cycle length requirements, a higher axial blanket enrichment (2.6 w/o) and shorter length (6 inches) is recommended. (author)

  2. Development and trial manufacturing of 1/2-scale partial mock-up of blanket box structure for fusion experimental reactor

    Science.gov (United States)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

    Conceptual design of breeding blanket has been discussed during the CDA (Conceptual Design Activities) of ITER (International Thermonuclear Experimental Reactor). Structural concept of breeding blanket is based on box structure integrated with first wall and shield, which consists of three coolant manifolds for first wall, breeding and shield regions. The first wall must have cooling channels to remove surface heat flux and nuclear heating. The box structure includes plates to form the manifolds and stiffening ribs to withstand enormous electromagnetic load, coolant pressure and blanket internal (purge gas) pressure. A 1/2-scale partial model of the blanket box structure for the outboard side module near midplane is manufactured to estimate the fabrication technology, i.e. diffusion bonding by HIP (Hot Isostatic Pressing) and EBW (Electron Beam Welding) procedure. Fabrication accuracy is a key issue to manufacture first wall panel because bending deformation during HIP may not be small for a large size structure. Data on bending deformation during HIP was obtained by preliminary manufacturing of HIP elements. For the shield structure, it is necessary to reduce the welding strain and residual stress of the weldment to establish the fabrication procedure. Optimal shape of the parts forming the manifolds, welding locations and welding sequence have been investigated. In addition, preliminary EBW tests have been performed in order to select the EBW conditions, and fundamental data on built-up shield have been obtained. Especially, welding deformation by joining the first wall panel to the shield has been measured, and total deformation to build-up shield by EBW has been found to be smaller than 2 mm. Consequently, the feasibility of fabrication technologies has been successfully demonstrated for a 1m-scaled box structure including the first wall with cooling channels by means of HIP, EBW and TIG (Tungsten Inert Gas arc)-welding.

  3. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    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/m2 and a particle heat flux of 1 MW/m2. 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

  4. Armored geckos: A histological investigation of osteoderm development in Tarentola (Phyllodactylidae) and Gekko (Gekkonidae) with comments on their regeneration and inferred function.

    Science.gov (United States)

    Vickaryous, M K; Meldrum, G; Russell, A P

    2015-11-01

    Osteoderms are bone-rich organs found in the dermis of many scleroglossan lizards sensu lato, but are only known for two genera of gekkotans (geckos): Tarentola and Gekko. Here, we investigate their sequence of appearance, mode of development, structural diversity and ability to regenerate following tail loss. Osteoderms were present in all species of Tarentola sampled (Tarentola annularis, T. mauritanica, T. americana, T. crombei, T. chazaliae) as well as Gekko gecko, but not G. smithii. Gekkotan osteoderms first appear within the integument dorsal to the frontal bone or within the supraocular scales. They then manifest as mineralized structures in other positions across the head. In Tarentola and G. gecko, discontinuous clusters subsequently form dorsal to the pelvis/base of the tail, and then dorsal to the pectoral apparatus. Gekkotan osteoderm formation begins once the dermis is fully formed. Early bone deposition appears to involve populations of fibroblast-like cells, which are gradually replaced by more rounded osteoblasts. In T. annularis and T. mauritanica, an additional skeletal tissue is deposited across the superficial surface of the osteoderm. This tissue is vitreous, avascular, cell-poor, lacks intrinsic collagen, and is herein identified as osteodermine. We also report that following tail loss, both T. annularis and T. mauritanica are capable of regenerating osteoderms, including osteodermine, in the regenerated part of the tail. We propose that osteoderms serve roles in defense against combative prey and intraspecific aggression, along with anti-predation functions.

  5. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    enhance the production of {sup 238}Pu. However, using recycled uranium as blanket fuel may degrade the neutron economy (as two neutron-captures are required by produce {sup 238}Pu). Another option is to fuel the blanket with thorium and uranium (natural or depleted) oxide. The intent here is to facilitate a controlled production of denatured uranium with about 5% or 20% {sup 233}U in total uranium discharged from the breeder blanket by innovative fuel design. The outcome of this option is to produce 5%-enriched uranium (in {sup 233}U) which can be used as fuel for LWRs, or 20%-enriched uranium (in {sup 233}U) to be used as driver fuel in fast reactor cores. The challenge for this option is to deal with the presence of high radiation from {sup 232}U and its decay daughters which could complicate its use as fresh fuel for LWRs and fast reactors. This paper addresses the pros and cons of these options of degrading the breeder blanket plutonium, and discusses their effectiveness for proliferation resistance. References 1. M. Saito, H. Sagara, and Y. Peryoga, 'Development of Innovative Nuclear Technology to Produce Protected Plutonium with High Proliferation Resistance - Grand Design of Project', Trans. ANS Winter Meeting, Washington DC, Nov. 2004. 2. H. Sagara, M. Saito, Y. Peryoga, A. Ezoubtchenko, and A. Takivayev, 'Denaturing of Plutonium by Transmutation of Minor-Actinides for Enhancement of Proliferation Resistance', Journal of Nuclear Science and Technology, Vol. 42, No.2, pp. 161-168, February 2005. 3. M. Saito, 'Multi-component Self-consistent Nuclear Energy System: Protected Plutonium Production (P3)', Int. J. Nuclear Energy Science and Technology, Vol.1, No. 2/3, 2005. (authors)

  6. Ballistic behavior of high hardness perforated armor plates against 7.62 mm armor piercing projectile

    International Nuclear Information System (INIS)

    Highlights: • High hardness perforated plates can be used effectively in ballistic protection. • Perforated plate has potential of decreasing areal mass efficiency dramatically. • The defeating mechanism of multilayer perforated plates includes three principles. • Deviation from trajectory, core fracture and nose erosion are defeating mechanisms. • With the simulations and tests, the bullet defeating mechanism has been explained. - Abstract: In this paper, some of the important defeating mechanisms of the high hardness perforated plates against 7.62 × 54 armor piercing ammunition were investigated. The experimental and numerical results identified three defeating mechanisms effective on perforated armor plates which are the asymmetric forces deviates the bullet from its incident trajectory, the bullet core fracture and the bullet core nose erosion. The initial tests were performed on the monolithic armor plates of 9 and 20 mm thickness to verify the fidelity of the simulation and material model parameters. The stochastic nature of the ballistic tests on perforated armor plates was analyzed based on the bullet impact zone with respect to holes. Various scenarios including without and with bullet failure models were further investigated to determine the mechanisms of the bullet failure. The agreement between numerical and experimental results had significantly increased with including the bullet failure criterion and the bullet nose erosion threshold into the simulation. As shown in results, good agreement between Ls-Dyna simulations and experimental data was achieved and the defeating mechanism of perforated plates was clearly demonstrated

  7. The ITER Blanket System Design Challenge

    International Nuclear Information System (INIS)

    Full text: The blanket system is one of the most technically challenging components of the ITER machine, having to accommodate high heat fluxes from the plasma, large electromagnetic loads during off-normal events and demanding interfaces with many key components (in particular the vacuum vessel and in-vessel coils) and the plasma. Plasma scenarios impose demanding requirements on the blanket in terms of heat fluxes on various areas of the first wall during different phases of operation (inboard and outboard midplane for start-up/shut-down scenarios and the top region close to the secondary X-point during flat top) as well as large electro-magnetic (EM) loads and transient energy deposition during off-normal plasma events (such as disruptions and vertical displacement events (VDE)). The high heat fluxes resulting in some areas have necessitated the use of “enhanced heat flux” panels capable of accommodating an incident heat flux of up to 5 MW/m2 in steady state. The other regions utilize “normal heat flux” panels, which have been developed and tested for a heat flux of the order of 1 — 2 MW/m2. The FW shaping design requires a compromise between the conflicting requirements for accommodation of steady state and transient loads (energy deposition during off-normal events). A shaped surface increases the heat loads which are due to plasma particles following the field lines compared to a perfectly toroidal surface. The blanket provides a major contribution to the shielding of the vacuum vessel and coils. A challenging criterion is the need to limit the integrated heating in the toroidal field coil (TFC) to ∼ 14 kW. This is particularly severe on the inboard leg where approximately 80% of the total nuclear heat on the TFC is deposited. Several design modifications were considered and analyzed to help achieve this, including increasing the inboard blanket radial thickness and reducing the assembly gaps. This paper summarizes the latest progress in the

  8. Statistical Analysis of Human Reliability of Armored Equipment

    Institute of Scientific and Technical Information of China (English)

    LIU Wei-ping; CAO Wei-guo; REN Jing

    2007-01-01

    Human errors of seven types of armored equipment, which occur during the course of field test, are statistically analyzed. The human error-to-armored equipment failure ratio is obtained. The causes of human errors are analyzed. The distribution law of human errors is acquired. The ratio of human errors and human reliability index are also calculated.

  9. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 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.

  10. Crucial issues on liquid metal blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Malang, S. (Kernforschungszentrum Karlsruhe (Germany)); Leroy, P. (CEA, CEN Saclay, 91 - Gif-sur-Yvette (France)); Casini, G.P. (CEC, Joint Research Centre (JRC), Ispra (Italy)); Mattas, R.F. (Argonne National Lab., IL (United States)); Strebkov, Yu. (Research and Development Inst. of Power Engineering, Moscow (USSR))

    1991-12-01

    Typical design concepts of liquid metal breeder blankets for power reactors are explained and characterized. The major problems of these concepts are described for both water-cooled blankets and self-cooled blankets. Three crucial issues of liquid metal breeder blankets are investigated. They are in the fields of magnetohydrodynamics, tritium control and safety. The influence of the magnetic field on liquid metal flow is of special interest for self-cooled blankets. The main problems in this field and the status of the related R and D work are described. Tritium permeation losses to the cooling water is a crucial issue for water-cooled blankets. Methods for its reduction are discussed. An inherent problem of all liquid breeder blankets is the potential release of activated products in the case of chemical reactions between the breeder material and water or reactive gases. The most important issues in this field are described. (orig.).

  11. Investigation of heat treatment conditions of structural material for blanket fabrication process

    International Nuclear Information System (INIS)

    This paper presents recent results of thermal hysteresis effects on ceramic breeder blanket structural material. Reduced activation ferritic/martensitic (RAF) steel is the leading candidates for the first wall structural materials of breeding blankets. RAF steel demonstrates superior resistance to high dose neutron irradiation, because the steel has tempered martensite structure which contains the number of sink site for radiation defects. This microstructure obtained by two-step heat treatment, first is normalizing at temperature above 1200 K and the second is tempering at temperature below 1100 K. Recent study revealed the thermal hysteresis has significant impacts on the post-irradiation mechanical properties. The breeding blanket has complicated structure, which consists of tungsten armor and thin first wall with cooling pipe. The blanket fabrication requires some high temperature joining processes. Especially hot isostatic pressing (HIP) is examined as a near-net-shape fabrication process for this structure. The process consists of heating above 1300 K and isostatic pressing at the pressure above 150 MPa followed by tempering. Moreover ceramics pebbles are packed into blanket module and the module is to be seamed by welding followed by post weld heat treatment in the final assemble process. Therefore the final microstructural features of RAFs strongly depend on the blanket fabrication process. The objective of this work is to evaluate the effects of thermal hysteresis corresponding to blanket fabrication process on RAFs microstructure in order to establish appropriate blanket fabrication process. Japanese RAFs F82H (Fe-0.1C-8Cr-2W-0.2V-0.05Ta) was investigated by metallurgical method after isochronal heat treatment up to 1473 K simulating high temperature bonding process. Although F82H showed significant grain growth after conventional solid HIP conditions (1313 K x 2 hr.), this coarse grained microstructure was refined by the post HIP normalizing at

  12. 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.

  13. Direct Lit Electrolysis In A Metallic Lithium Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Colon-Mercado, H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Babineau, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Elvington, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Teprovich, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Vaquer, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-13

    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 fission/fusion reactors is critical in order to maintained low concentrations.  This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Because of the high affinity of tritium for the blanket, extraction is complicated at the required low levels. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering the hydrogen and deuterium thru an electrolysis step at high temperatures. 

  14. Dynamic behavior of high strength armor steels

    International Nuclear Information System (INIS)

    The dynamic properties of the armor steels Mars 190, Mars 240 and Mars 300 have been determined for strain rates 10-3 s-16 s-1. The planar plate impact technique in combination with a VISAR has been used for high strain rate testing (dε/dt>104 s-1). The dynamic properties Hugoniot-elastic-limit, spall strength, shock velocity-particle velocity-relation and stress-strain-relation will be presented. Tests at intermediate strain rates 102 s-13 have been performed using Taylor tests and a Split-Hopkinson-Pressure-Bar (SHPB) setup providing information about strain rate sensitivity, strain hardening and dynamic strength. In addition to the determination of the dynamic properties, the influence of the loading process on the microstructure and the fracture mechanisms have been determined using optical and scanning electron microscopy (SEM). (orig.)

  15. Design of armor for protection against blast and impact

    Science.gov (United States)

    Rahimzadeh, Tanaz; Arruda, Ellen M.; Thouless, M. D.

    2015-12-01

    The features of blast and impact that can damage a delicate target supported by a structure include both the peak pressure and the impulse delivered to the structure. This study examines how layers of elastic and visco-elastic materials may be assembled to mitigate these features. The impedance mismatch between two elastic layers is known to reduce the pressure, but dissipation is required to mitigate the transmitted impulse in light-weight armor. A novel design concept called impact or blast tuning is introduced in which a multi-layered armor is used to tune the stress waves resulting from an impact or blast to specific frequencies that match the damping frequencies of visco-elastic layers. The material and geometrical parameters controlling the viscous dissipation of the energy within the armor are identified for a simplified one-dimensional system, to provide insight into how the optimal design of multi-use armor might be based on this concept.

  16. 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)

    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 Li4SiO4 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.)

  17. Manufacture of blanket shield modules for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzetto, P. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: Patrick.Lorenzetto@tech.efda.org; Boireau, B. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Boudot, C. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Bucci, P. [CEA, DTEN/S3ME/LMIC, 17 rue des Martyrs, F-38054 Grenoble (France); Furmanek, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Ioki, K. [ITER IT, Boltzmannstr. 2, D-85748 Garching (Germany); Liimatainen, J. [Metso Powdermet, P.O. Box 306, FIN-33101 Tampere (Finland); Peacock, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Sherlock, P. [NNC Ltd., Booths Hall, Knutsford, Cheshire WA16 8QZ (United Kingdom); Taehtinen, S. [VTT Industrial Systems, P.O. Box 1704, Espoo, FIN-02044 VTT (Finland)

    2005-11-15

    A research and development programme for the ITER blanket shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups (small scale and medium scale) and full-scale prototypes of shield blocks (SB) and first wall (FW) panels. The manufacturing feasibility of FW panels has been demonstrated for two copper alloy candidates. Two designs have been developed for the manufacture of the SB, one for a conventional fabrication route and one for a fabrication route based on the hot isostatic press technology. This paper presents the fabrication routes developed in Europe for the manufacture of the ITER Shield modules.

  18. Nondestructive characterization of UHMWPE armor materials

    Energy Technology Data Exchange (ETDEWEB)

    Chiou, Chien-Ping; Margetan, Frank J.; Barnard, Daniel J.; Hsu, David K.; Jensen, Terrence; Eisenmann, David [Center for Nondestructive Evaluation, Iowa State University, Ames, IA 50011 (United States)

    2012-05-17

    Ultra-high molecular weight polyethylene (UHMWPE) is a material increasingly used for fabricating helmet and body armor. In this work, plate specimens consolidated from thin fiber sheets in series 3124 and 3130 were examined with ultrasound, X-ray and terahertz radiation. Ultrasonic through-transmission scans using both air-coupled and immersion modes revealed that the 3130 series material generally had much lower attenuation than the 3124 series, and that certain 3124 plates had extremely high attenuation. Due to the relatively low inspection frequencies used, pulse-echo immersion ultrasonic testing could not detect distinct flaw echoes from the interior. To characterize the nature of the defective condition that was responsible for the high ultrasonic attenuation, terahertz radiation in the time-domain spectroscopy mode were used to image the flaws. Terahertz scan images obtained on the high attenuation samples clearly showed a distribution of a large number of defects, possibly small planar delaminations, throughout the volume of the interior. Their precise nature and morphology are to be verified by optical microscopy of the sectioned surface.

  19. Numerical simulation of armor capability of AI2O3 and SiC armor tiles

    Science.gov (United States)

    Rashid, T.; Aleem, M. A.; Akbar, S.; Rauf, A.; Shuaib, M.

    2016-08-01

    Alumina and Silicon Carbide armor plates have been tested numerically against 7.62x51 (mm x mm) armor piercing (AP) projectiles. A 2-D problem with axial symmetry has been designedand the simulations were carried out using commercial software ANSYS AUTODYN. Experiments were modeled for Alumina (99.5%), Alumina (99.7%) and SiC with a range of tile thicknesses (5, 10, 15 and 20 mm). The projectile was chosen as 7.62 x 51AP bullet (initial velocity 810 m/sec)with two different core materials Steel 4340 and WC, however, casing material was copper for both cores. SiC showed better defense against AP bullet as compared to Al2O3. The residual velocity and momentum of the bullet were found to decrease with increasing tile thickness. SiC tiles with thickness 15mm and 20 mm successfully sustained penetration against steel 4340 and WC core bullets, respectively. However none of the Alumina targets succeeded in stopping the bullet.

  20. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    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/m2. 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

  1. Simulation of sludge blanket height in clarifiers

    Institute of Scientific and Technical Information of China (English)

    ZHOU Zhen; WU Zhi-chao; WANG Zhi-wei; GU Guo-wei

    2009-01-01

    Sludge blanket height (SBH) is an important parameter in the clarifier design,operation and control.Based on an overview and classification of SBH algorithms,a modifed SBH algorithm is proposed by incorporating a threshold concentration limit into a relative concentration sharp change algorithm to eliminate the disturbance of compression interfaces on the correct simulation of SBH.Pilot-scale test data are adopted to compare reliability of three SBH algorithms reported in literature and the modified SBH algorithm developed in this paper.Calculated results demonstrate that the three SBH algorithms give results with large deviation (>50%) from measured SBH,especially under low solid flux conditions.The modified algorithm is computationally efficient and reliable in matching the measured data.It is incorporated into a onedimensional clarifier model for stable simulation of pilot-scale experimental clarifier data and into dynamic simulation of a full-scale wastewater treatment plant (WWTP) clarifier data.

  2. Neutronic and thermomechanical analysis of the water-cooled lithium-lead blanket design for a DEMONET reactor

    International Nuclear Information System (INIS)

    Within the framework of the European DEMO blanket study programme, CEA and the JRC of Ispra are jointly developing a water-cooled lithium-lead blanket concept. The new DEMONET reactor configuration released in Spring 1990 and currently specified in the EC programme is the basis of neutronic and thermomechanical studies for the proposed box-shaped blanket concept. Considering the high blanket coverage, it is now possible to reach tritium self-sufficiency without making use of beryllium (neutronic calculations indicate a global tritium breeding ratio of the order of 1.16). (orig.)

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

    International Nuclear Information System (INIS)

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m2. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ''beryllium-joint'' concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket

  4. Attachment system for helium-cooled blanket of RF DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leshukov, A. E-mail: leshu@entek.ru; Blinov, Y.; Kovalenko, V.; Shatalov, G.; Strebkov, Y.; Strizhov, A

    2002-11-01

    The development of DEMO thermonuclear reactor is a part of Russian national program on the fusion process mastering. The DEMO-S (stationary thermonuclear reactor) should be the logic continuation of the ITER-type projects (pulse thermonuclear reactors) and the prototype for commercial power plants. DEMO reactor layout suggests to use the segmented blanket with mounting/dismounting procedure through the vacuum vessel vertical ports. Taking into account this layout the blanket attachment system has been developed and the present paper is devoted to this subject. The considered attachment system includes the lower and upper toroidal support assemblies which connect all the blanket segments in the enclosed ring. In it's turn the lower support assemblies attached to the vacuum vessel through the system of hinged support pillars. The heights of support pillars for inboard and outboard blankets are selected so that to indemnify the blanket massif thermal expansions in vertical and radial directions. The support pillars have been calculated on strength taking into account the electromagnetic loads from the plasma disruptions and blanket mass. The selection of high-strength chromium steel as a structural material for the support pillars could be considered as the results of strength analysis. The conclusions on the possibility to apply this attachment system for fusion reactor blanket and the critical issues are contained in this paper too.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J.

    1980-01-01

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

  7. Electromagnetic Armor Based on the Principles of Reconnection Electromagnetic Launcher

    Institute of Scientific and Technical Information of China (English)

    Yu Haiqing; Li Weitao; Gao Haizhen

    2015-01-01

    Active electromagnetic armor is the latest concept of defense system, and it is highly probable for this technology to be applied to many domains such as future tanks, armored vehicles and the armor of warships. The studies on active electromagnetic defense technology in the world have a history of decades. But, there still exist bottlenecks and the interceptor is one of them. This paper studies the problem of interceptors based on the principles of reconnection electromagnetic launcher and presents a kind of simulation model of it. A small scale laboratory model is made, and related experimental researches and theoretical analyses have been carried out when the projectiles are different kinds of material and shape. Satisfactory results are reached and agree with the theoretical analyses.

  8. Multivariable optimization of fusion reactor blankets

    International Nuclear Information System (INIS)

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% 6Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO2 breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO2 breeding blanket enriched to 34% 6Li

  9. 46 CFR 111.05-7 - Armored and metallic sheathed cable.

    Science.gov (United States)

    2010-10-01

    ... Armored and metallic sheathed cable. When installed, the metallic armor or sheath must meet the installation requirements of Section 25 of IEEE 45-2002 (incorporated by reference; see 46 CFR 110.10-1). ... 46 Shipping 4 2010-10-01 2010-10-01 false Armored and metallic sheathed cable. 111.05-7...

  10. 48 CFR 252.225-7030 - Restriction on Acquisition of Carbon, Alloy, and Armor Steel Plate.

    Science.gov (United States)

    2010-10-01

    ... of Carbon, Alloy, and Armor Steel Plate. 252.225-7030 Section 252.225-7030 Federal Acquisition... Acquisition of Carbon, Alloy, and Armor Steel Plate. As prescribed in 225.7011-3, use the following clause: Restriction on Acquisition of Carbon, Alloy, and Armor Steel Plate (DEC 2006) (a) Carbon, alloy, and...

  11. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    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

  12. Steering Control of Wheeled Armored Vehicle with Brushless DC Motor

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    Considering the steering characters of one type of wheeled armored vehicle, a brushless direct current (DC) motor is adapted as the actuator for steering control. After investigating the known algorithms, one kind of algorithm, which combines the fuzzy logic control with the self-adapting PID control and the startup and pre-brake control, is put forward. Then a test-bed is constructed, and an experiment is conducted. The result of experiment confirms the validity of this algorithm in steering control of wheeled armored vehicle with brushless DC motor.

  13. Research on Welding Processes of Armor Steels%装甲钢焊接技术研究进展

    Institute of Scientific and Technical Information of China (English)

    谭俊; 张勇

    2013-01-01

    装甲钢是一种用于作战装备的保护性合金材料,装甲钢的焊接是装甲钢结构的主要连接方式.从装甲钢的裂纹类别、特点以及其形成原因出发,介绍了装甲钢焊接材料、焊接工艺和提高焊接接头性能的措施.主要包括奥氏体焊条、铁素体焊条及奥氏体/铁素体双向焊条;电弧焊、CO2气体保护焊、熔化极气体保护焊、修复装甲钢裂纹的铝热焊、机器人焊,激光焊等焊接工艺;焊后热处理、超声波冲击法、焊趾砂轮打磨法与钨极电弧焊法.并在装甲钢焊接技术研究现状分析的基础上,提出装甲钢焊接技术今后的改进措施与发展方向.%The armor steel is a kind of protective alloy material which is used in a variety of combat equipment. The welding is a main joining method of armor steel structure. On the base of crack category, characteristics and formation causes of armor steel, the welding materials and welding processes of armor steel, and the measures taken to improve the properties of welded joint were introduced. Those include austenite welding electrode, ferrite and austenite/ferrite dual phase welding electrodes, the processes of arc welding, CO2 gas protection welding and thermit welding of armor steel crack repairing, and the measures taken to improve the properties of the welded joint of heat treatment after welding, ultrasonic impact, toe grinding and tungsten arc welding. The improving measures and development direction of armor steel welding were put forward by analyzing the armor steel welding processes.

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

    International Nuclear Information System (INIS)

    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, Li2O/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., N2) 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 Li2O/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

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

    International Nuclear Information System (INIS)

    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, Li2O/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., N2) 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 Li2O/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

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

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

    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 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/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 concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

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

    International Nuclear Information System (INIS)

    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, Li2O/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., N2) 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 Li2O/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

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

    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 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/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. Exploratory Study of Blanket Liquid Curtain

    Institute of Scientific and Technical Information of China (English)

    HUGang; HUANGJinhua; FENGKaiming

    2003-01-01

    Blankets and other in-vessel components are easily damaged owing to their circumstance of high radiation and high heat. To protect them, first wall design should be considered. Owing to its high heat removal nd self-refreshing capability, liquid metal first wall has been seen as a potential first wall for a fusion reactor in the future. Blanketliquid curtain is actually a special liquid metal wall to protect blanket.

  1. Application Effect’s Research of Vetiver Eco Blanket in Pubugou Reservoir Fluctuating Zone

    OpenAIRE

    Lan Huijuan; Zheng Kaiyuan; Ni Fuquan; Deng Yu; Liu Dengyu; Yang Xinwei; Wang Tao

    2015-01-01

    To solve the ecological disasters in Pubugou Reservoir Fluctuating Zone, ecological blanket governance model is proposed in this paper, which may provide good early environment for plants’ survival in fluctuation zone, and then play the function of greening and sustainable development to ensure the slopes’ stability. Meanwhile, based on the result of vetiver ecological blanket in Hanyuan experimental zone, we find that three kinds of typical Fluctuating Zone slope’s greening effect is good, w...

  2. A high tritium breeding ratio (TBR) blanket concept and requirements for nuclear data relating to TBR

    International Nuclear Information System (INIS)

    Significance of developing a blanket having a sufficiently larger tritium breeding ratio (TBR) than 1.0 is discussed. For this purpose, a high TBR blanket with a front breeder zone just before the multiplier is introduced together with conventional blankets. From discussion of TBR characteristics in these blankets, the necessity of improving on nuclear data, i.e. reducing uncertainties is presented as follows; σs, σnp and σnα of structural and coolant materials, and σn2n of the multiplier at higher energies above several MeV, and σnγ of these materials and σnαT of 6Li at energies from several hundred keV to thermal energy. (author)

  3. 从跳舞毯的风靡看动作技能学习的发展趋势%Look forward the developing tendency of action skilllearning from the fashion of dancing blanket

    Institute of Scientific and Technical Information of China (English)

    苏坚贞; 郑岩平

    2001-01-01

    The fashion of dancing blanket can make people think a lot ofthings,and also can lead to a lot of topics of contemporary social.It is one of the topics that if the action skill learning can be accomplished by wielding computers.This paper tries to start with the conception of action skill,and analyzes the possibilities of wielding computers to learning action skill.Then deduces the developing tendency of action skill learning.%跳舞毯的风靡会使人类联想到许许多多的事情,因而也会引出一些当代社会的热门话题,动作技能的学习是否可以通过计算机进行教学来完成正是其中一例。试图从动作技能的概念入手,分析动作技能学习是否具有进行计算机教学的可能性,进而得出动作技能学习的发展趋势。

  4. Resolution of proliferation issues for a SFR blanket with a specific application

    Energy Technology Data Exchange (ETDEWEB)

    Stauff, N.E. [31 rue baudelaire, voisins le bretonneux, 78960 (France); Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States); Forget, B.; Driscoll, M.J. [Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States)

    2009-06-15

    The Sodium Fast Reactor is seen as the most realistic Gen-IV reactor to be built in the near future. France and the US are still developing their designs; these will require improved safety, competitive economics, and also proliferation resistance. To meet this last requirement, both French and American designers show some concerns with the use of breeding blankets. France and the USA won't need breeding blankets to produce plutonium because they already have large amounts of plutonium bred from their LWR fleet to start a new SFR fleet, thus breeding blankets are mainly of interest for minor actinide burning. On the contrary, India and China express great interest in blankets for their SFR designs, to reach a positive breeding gain. For example, the Indian PFBR, a 500 MWe oxide-fueled SFR has a breeding ratio of 1.05. Blankets are used in a Fast Reactor to increase the breeding ratio of the core, by breeding a significant amount of plutonium. The Plutonium bred within these blankets, if these are loaded with Uranium only, is generally of a very high quality, which makes it easily used in a nuclear explosive device. Our research has shown that the plutonium in breeding blankets can be made less attractive to make a nuclear explosive device than LWR-bred plutonium with a burnup of 50 MWd/Kg. Minor actinide doping and moderator addition were the two options studied, as they increase Pu{sup 238} and Pu{sup 240} production. In the work reported here, the methodology developed for securing a breeding blanket was successfully applied to the Indian PFBR. The full paper will describe a design of the PFBR breeding proliferation resistant plutonium within its blankets. The blankets were rendered secure by adding a zirconium hydride moderator and a small volume of MAs. It was demonstrated that reducing the attractiveness of the blanket plutonium would require no external MA dependency by choosing an adequate fuel cycle. The characteristics and performance of this design

  5. Electrical connectors for blanket modules in ITER

    International Nuclear Information System (INIS)

    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

  6. Multiplier, moderator, and reflector materials for lithium-vanadium fusion blankets.

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Smith, D. L.

    1999-10-07

    The self-cooled lithium-vanadium fusion blanket concept has several attractive operational and environmental features. In this concept, liquid lithium works as the tritium breeder and coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because of its superior performance relative to other alloys for this application. However, this concept has poor attenuation characteristics and energy multiplication for the DT neutrons. An advanced self-cooled lithium-vanadium fusion blanket concept has been developed to eliminate these drawbacks while maintaining all the attractive features of the conventional concept. An electrical insulator coating for the coolant channels, spectral shifter (multiplier, and moderator) and reflector were utilized in the blanket design to enhance the blanket performance. In addition, the blanket was designed to have the capability to operate at high loading conditions of 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading. This paper assesses the spectral shifter and the reflector materials and it defines the technological requirements of this advanced blanket concept.

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

    International Nuclear Information System (INIS)

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

  8. Multiscale impacts of armoring on Salish Sea shorelines: Evidence for cumulative and threshold effects

    Science.gov (United States)

    Dethier, Megan N.; Raymond, Wendel W.; McBride, Aundrea N.; Toft, Jason D.; Cordell, Jeffery R.; Ogston, Andrea S.; Heerhartz, Sarah M.; Berry, Helen D.

    2016-06-01

    Shoreline armoring is widespread in many parts of the protected inland waters of the Pacific Northwest, U.S.A, but impacts on physical and biological features of local nearshore ecosystems have only recently begun to be documented. Armoring marine shorelines can alter natural processes at multiple spatial and temporal scales; some, such as starving the beach of sediments by blocking input from upland bluffs may take decades to become visible, while others such as placement loss of armoring construction are immediate. We quantified a range of geomorphic and biological parameters at paired, nearby armored and unarmored beaches throughout the inland waters of Washington State to test what conditions and parameters are associated with armoring. We gathered identical datasets at a total of 65 pairs of beaches: 6 in South Puget Sound, 23 in Central Puget Sound, and 36 pairs North of Puget Sound proper. At this broad scale, demonstrating differences attributable to armoring is challenging given the high natural variability in measured parameters among beaches and regions. However, we found that armoring was consistently associated with reductions in beach width, riparian vegetation, numbers of accumulated logs, and amounts and types of beach wrack and associated invertebrates. Armoring-related patterns at lower beach elevations (further vertically from armoring) were progressively harder to detect. For some parameters, such as accumulated logs, there was a distinct threshold in armoring elevation that was associated with increased impacts. This large dataset for the first time allowed us to identify cumulative impacts that appear when increasing proportions of shorelines are armored. At large spatial and temporal scales, armoring much of a sediment drift cell may result in reduction of the finer grain-size fractions on beaches, including those used by spawning forage fish. Overall we have shown that local impacts of shoreline armoring can scale-up to have cumulative and

  9. Neutronic implications of lead-lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W.R.

    1982-08-01

    Lead-lithium alloys have been proposed for use in several conceptual blanket designs for both inertial and magnetic confinement fusion reactors. In most cases, Pb/sub 83/Li/sub 17/, a eutectic with a melting point of 235/sup 0/C, is the chosen composition. The primary reasons for using Pb/sub 83/Li/sub 17/ instead of Li as the tritium breeding material are the perceived safety advantages, low tritium solubility, and favorable neutronic characteristics. This paper describes the neutronic characteristics of Pb/sub 83/Li/sub 17/ blankets with emphasis on the enhanced neutron leakage through chamber ports and the degradation in blanket performance parameters that occurs as a result of the enhanced leakage.

  10. Highly deformable bones: unusual deformation mechanisms of seahorse armor.

    Science.gov (United States)

    Porter, Michael M; Novitskaya, Ekaterina; Castro-Ceseña, Ana Bertha; Meyers, Marc A; McKittrick, Joanna

    2013-06-01

    Multifunctional materials and devices found in nature serve as inspiration for advanced synthetic materials, structures and robotics. Here, we elucidate the architecture and unusual deformation mechanisms of seahorse tails that provide prehension as well as protection against predators. The seahorse tail is composed of subdermal bony plates arranged in articulating ring-like segments that overlap for controlled ventral bending and twisting. The bony plates are highly deformable materials designed to slide past one another and buckle when compressed. This complex plate and segment motion, along with the unique hardness distribution and structural hierarchy of each plate, provide seahorses with joint flexibility while shielding them against impact and crushing. Mimicking seahorse armor may lead to novel bio-inspired technologies, such as flexible armor, fracture-resistant structures or prehensile robotics.

  11. Production and characterization of ceramics for armor application

    International Nuclear Information System (INIS)

    The fabrication of devices for ballistic protection as bullet proof vests and helmets and armored vehicles has been evolving over the past years along with the materials and models used for this specific application. The requirements for high efficient light-weight ballistic protection systems which not interfere in the user comfort and mobility has driven the research in this area. In this work we will present the results of characterization of two ceramics based on alumina and silicon carbide. The ceramics were produced in lab scale and the specific mass, scanning electron microscopy (SEM) microstructure, Vickers hardness, flexural resistance at room temperature and X-ray diffraction were evaluated. Ballistic tests performed in the selected materials showed that the ceramics present armor efficiency. (author)

  12. Armoring a droplet: Soft jamming of a dense granular interface

    Science.gov (United States)

    Lagubeau, Guillaume; Rescaglio, Antonella; Melo, Francisco

    2014-09-01

    Droplets and bubbles protected by armors of particles have found vast applications in encapsulation, stabilization of emulsions and foams, or flotation processes. The liquid phase stores capillary energy, while concurrently the solid contacts of the granular network induce friction and energy dissipation, leading to hybrid interfaces of combined properties. By means of nonintrusive tensiometric methods and structural measurements, we distinguish three surface phases of increasing rigidity during the evaporation of armored droplets. The emergence of surface rigidity is reminiscent of jamming of granular matter, but it occurs differently since it is marked by a step by step hardening under surface compression. These results show that the concept of the effective surface tension remains useful only below the first jamming transition. Beyond this point, the surface stresses become anisotropic.

  13. Controls on and effects of armoring and vertical sorting in aeolian dune fields: A numerical simulation study

    Science.gov (United States)

    Gao, Xin; Narteau, Clément; Rozier, Olivier

    2016-03-01

    Unlike ripples, there are only few numerical studies on grain size segregation at the scale of dunes in aeolian environments. Here we use a cellular automaton model to analyze vertical sorting in granular mixtures under steady unidirectional flow conditions. We investigate the feedbacks between dune growth and the segregation mechanisms by varying the size of coarse grains and their proportion within the bed. We systematically observe the development of a horizontal layer of coarse grains at the top of which sorted bed forms may grow by amalgamation. The formation of such an armor layer controls the overall sediment transport and availability. The emergence of dunes and the transition from barchan to transverse dune fields depend only on the grain size distribution of the initial sediment layer. As confirmed by observation, this result indicates that armor layers should be present in most arid deserts, where they are likely to control dune morphodynamics.

  14. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    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

  15. Requirements for helium cooled pebble bed blanket and R and D activities

    Energy Technology Data Exchange (ETDEWEB)

    Carloni, D., E-mail: dario.carloni@kit.edu; Boccaccini, L.V.; Franza, F.; Kecskes, S.

    2014-10-15

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.

  16. Requirements for helium cooled pebble bed blanket and R and D activities

    International Nuclear Information System (INIS)

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  18. A recurrent regulatory change underlying altered expression and Wnt response of the stickleback armor plates gene EDA.

    Science.gov (United States)

    O'Brown, Natasha M; Summers, Brian R; Jones, Felicity C; Brady, Shannon D; Kingsley, David M

    2015-01-01

    Armor plate changes in sticklebacks are a classic example of repeated adaptive evolution. Previous studies identified ectodysplasin (EDA) gene as the major locus controlling recurrent plate loss in freshwater fish, though the causative DNA alterations were not known. Here we show that freshwater EDA alleles have cis-acting regulatory changes that reduce expression in developing plates and spines. An identical T → G base pair change is found in EDA enhancers of divergent low-plated fish. Recreation of the T → G change in a marine enhancer strongly reduces expression in posterior armor plates. Bead implantation and cell culture experiments show that Wnt signaling strongly activates the marine EDA enhancer, and the freshwater T → G change reduces Wnt responsiveness. Thus parallel evolution of low-plated sticklebacks has occurred through a shared DNA regulatory change, which reduces the sensitivity of an EDA enhancer to Wnt signaling, and alters expression in developing armor plates while preserving expression in other tissues. PMID:25629660

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

    International Nuclear Information System (INIS)

    This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li2TiO3 and so on, fabrication technology developments and characterization of the Li2TiO3 and Li4SiO4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li2TiO3 and Li4SiO4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

  20. Fragmentation of armor piercing steel projectiles upon oblique perforation of steel plates

    Science.gov (United States)

    Paris, V.; Weiss, A.; Vizel, A.; Ran, E.; Aizik, F.

    2012-08-01

    In this study, a constitutive strength and failure model for a steel core of a14.5 mm API projectile was developed. Dynamic response of a projectile steel core was described by the Johnson-Cook constitutive model combined with principal tensile stress spall model. In order to obtain the parameters required for numerical description of projectile core material behavior, a series of planar impact experiments was done. The parameters of the Johnson-Cook constitutive model were extracted by matching simulated and experimental velocity profiles of planar impact. A series of oblique ballistic experiments with x-ray monitoring was carried out to study the effect of obliquity angle and armor steel plate thickness on shattering behavior of the 14.5 mm API projectile. According to analysis of x-ray images the fragmentation level increases with both steel plate thickness and angle of inclination. The numerical modeling of the ballistic experiments was done using commercial finite element code, LS-DYNA. Dynamic response of high hardness (HH) armor steel was described using a modified Johnson-Cook strength and failure model. A series of simulations with various values of maximal principal tensile stress was run in order to capture the overall fracture behavior of the projectile's core. Reasonable agreement between simulated and x-ray failure pattern of projectile core has been observed.

  1. Constitutive Modeling of Hot Deformation Behavior of High-Strength Armor Steel

    Science.gov (United States)

    Bobbili, Ravindranadh; Madhu, Vemuri

    2016-05-01

    The hot isothermal compression tests of high-strength armor steel under a wide range of deformation temperatures (1100-1250 °C) and strain rates of (0.001-1/s) were performed. Based on the experimental data, constitutive models were established using the original Johnson-Cook (JC) model, modified JC model, and strain-compensated Arrhenius model, respectively. The modified JC model considers the coupled effects of strain hardening, strain rate hardening, and thermal softening. Moreover, the prediction accuracy of these developed models was determined by estimating the correlation coefficient ( R) and average absolute relative error (AARE). The results demonstrate that the flow behavior of high-strength armor steel is considerably influenced by the strain rate and temperature. The original JC model is inadequate to provide good description on the flow stress at evaluated temperatures. The modified JC model and strain-compensated Arrhenius model significantly enhance the predictability. It is also observed from the microstructure study that at low strain rates (0.001-0.01/s) and high temperatures (1200-1250 °C), a typical dynamic recrystallization (DRX) occurs.

  2. Fragmentation of armor piercing steel projectiles upon oblique perforation of steel plates

    Directory of Open Access Journals (Sweden)

    Aizik F.

    2012-08-01

    Full Text Available In this study, a constitutive strength and failure model for a steel core of a14.5 mm API projectile was developed. Dynamic response of a projectile steel core was described by the Johnson-Cook constitutive model combined with principal tensile stress spall model. In order to obtain the parameters required for numerical description of projectile core material behavior, a series of planar impact experiments was done. The parameters of the Johnson-Cook constitutive model were extracted by matching simulated and experimental velocity profiles of planar impact. A series of oblique ballistic experiments with x-ray monitoring was carried out to study the effect of obliquity angle and armor steel plate thickness on shattering behavior of the 14.5 mm API projectile. According to analysis of x-ray images the fragmentation level increases with both steel plate thickness and angle of inclination. The numerical modeling of the ballistic experiments was done using commercial finite element code, LS-DYNA. Dynamic response of high hardness (HH armor steel was described using a modified Johnson-Cook strength and failure model. A series of simulations with various values of maximal principal tensile stress was run in order to capture the overall fracture behavior of the projectile’s core. Reasonable agreement between simulated and x-ray failure pattern of projectile core has been observed.

  3. NUMERICAL SIMULATION OF TWO-DIMENSIONAL OVERTOPPING AGAINST SEAWALLS ARMORED WITH ARTIFICIAL UNITS IN REGULAR WAVES

    Institute of Scientific and Technical Information of China (English)

    LU Yong-jin; LIU Hua; WU Wei; ZHANG Jiu-shan

    2007-01-01

    A new mathematical model for the overtopping against seawalls armored with artificial units in regular waves was established. The 2-D numerical wave flume, based on the Reynolds Averaged Navier-Stokes (RANS) equations and the standard k-ε turbulence model, was developed to simulate the turbulent flows with the free surface, in which the Volume Of Fluid (VOF) method was used to handle the large deformation of the free surface and the relaxation approach of combined wave generation and absorbing was implemented. In order to consider the effects of energy dissipation due to the armors on a slope seawall, a porous media model was proposed and implemented in the numerical wave flume. A series of physical model experiments were carried out in the same condition of the numerical simulation to determine the drag coefficient in the porous media model in terms of the overtopping discharge. Compared the computational value of overtopping over the seawall with the experimental data, the values of the effective drag coefficient was calibrated for the layers of blocks at different locations along the seawalls.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-22

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

  5. Review: BNL graphite blanket design concepts

    International Nuclear Information System (INIS)

    A review of the Brookhaven National Laboratory (BNL) minimum activity graphite blanket designs is made. Three designs are identified and discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a thick graphite screen (typically 30 cm or greater, depending on type as well as application-experimental power reactor or commercial reactor). 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. This energy is then either radiated to a secondary blanket with coolant tubes, as in types A and B, or is removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the structural material of the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude by the graphite screen, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma, whatever the degree of radiation damage

  6. 47 CFR 22.353 - Blanketing interference.

    Science.gov (United States)

    2010-10-01

    ... Operational and Technical Requirements Technical Requirements § 22.353 Blanketing interference. Licensees of...: ER17NO94.007 where d is the radial distance to the boundary, in kilometers p is the radial effective radiated power, in kilowatts The maximum effective radiated power in the pertinent direction,...

  7. Aerogel Blanket Insulation Materials for Cryogenic Applications

    Science.gov (United States)

    Coffman, B. E.; Fesmire, J. E.; White, S.; Gould, G.; Augustynowicz, S.

    2009-01-01

    Aerogel blanket materials for use in thermal insulation systems are now commercially available and implemented by industry. Prototype aerogel blanket materials were presented at the Cryogenic Engineering Conference in 1997 and by 2004 had progressed to full commercial production by Aspen Aerogels. Today, this new technology material is providing superior energy efficiencies and enabling new design approaches for more cost effective cryogenic systems. Aerogel processing technology and methods are continuing to improve, offering a tailor-able array of product formulations for many different thermal and environmental requirements. Many different varieties and combinations of aerogel blankets have been characterized using insulation test cryostats at the Cryogenics Test Laboratory of NASA Kennedy Space Center. Detailed thermal conductivity data for a select group of materials are presented for engineering use. Heat transfer evaluations for the entire vacuum pressure range, including ambient conditions, are given. Examples of current cryogenic applications of aerogel blanket insulation are also given. KEYWORDS: Cryogenic tanks, thermal insulation, composite materials, aerogel, thermal conductivity, liquid nitrogen boil-off

  8. ITER driver blanket, European Community design

    Energy Technology Data Exchange (ETDEWEB)

    Simbolotti, G. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Zampaglione, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Ferrari, M. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Gallina, M. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Mazzone, G. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Nardi, C. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Petrizzi, L. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Rado, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Violante, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Daenner, W. (NET Team, Max-Planck-Inst. fuer Plasmaphysik, Garching (Germany)); Lorenzetto, P. (NET Team, Max-Planck-Inst. fuer Plasmaphysik, Garching (Germany)); Gierszewski, P. (CFFTP, Mississauga, ON (Canada)); Gratt

    1993-07-01

    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.)

  9. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2014-06-15

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

  10. Breeding blanket development; Tritium release from breeder

    OpenAIRE

    土谷 邦彦; 河村 弘; 長尾 美春

    2006-01-01

    核融合炉ブランケットを設計するためには、微小球を用いたブランケット構造体の中性子照射に関する工学的データが必要不可欠である。工学的データのうち、トリチウム生成放出特性は、最も重要なデータの1つである。このため、トリチウム増殖材料の候補材であるチタン酸リチウム(Li2TiO3)微小球からのトリチウム生成放出試験を行い、トリチウム放出特性に対するスイープガス流量,照射温度,スイープガス中の水素添加量,熱中性子束の変化等の効果について調べた。本試験の結果、(1)Li2TiO3微小球充填体の外壁温度が100circC以上になった時、トリチウム放出が観測された。また、充填体の外壁温度が300sim400circCのとき、トリチウム生成・放出率(R/G)は1に到達した。(2)スイープガス流量を100sim900cm3/min(Li2TiO3微小球充填体の空塔速度:0.53sim4.8cm/s)の範囲で変化させても、定常時におけるLi2TiO3微小球充填体からのトリチウム放出に影響はなかった。(3)スイープガス中の水素添加量はトリチウム放出に影響することがわかった。...

  11. Long-term survivability of riprap for armoring uranium-mill tailings and covers: a literature review

    International Nuclear Information System (INIS)

    Pacific Northwest Laboratory (PNL) is investigating the use of a rock armoring blanket (riprap) to mitigate wind and water erosion of an earthen radon suppression cover applied to uranium mill tailings. Because the radon suppression cover and the tailings must remain intact for up to 1000 years or longer, the riprap must withstand natural weathering forces. This report is a review of information on rock weathering and riprap durability. Chemical and physical weathering processes, rock characteristics related to durability, climatic conditions affecting the degree and rate of weathering, and testing procedures used to measure weathering susceptibilities have been reviewed. Sampling and testing techniques, as well as analyses of physical and chemical weathering susceptibilities, are necessary to evaluate rock durability. Many potential riprap materials may not be able to survive 1000 years of weathering. Available techniques for durability testing cannot adequately predict rock durability for the 1000-year period because they do not consider the issue of time (i.e., how long must riprap remain stable). This report includes an Appendix, which discusses rock weathering, written by Dr. Richard Jahns of Stanford University

  12. Long-term survivability of riprap for armoring uranium-mill tailings and covers: a literature review. [203 references

    Energy Technology Data Exchange (ETDEWEB)

    Lindsey, C.G.; Long, L.W.; Begej, C.W.

    1982-06-01

    Pacific Northwest Laboratory (PNL) is investigating the use of a rock armoring blanket (riprap) to mitigate wind and water erosion of an earthen radon suppression cover applied to uranium mill tailings. Because the radon suppression cover and the tailings must remain intact for up to 1000 years or longer, the riprap must withstand natural weathering forces. This report is a review of information on rock weathering and riprap durability. Chemical and physical weathering processes, rock characteristics related to durability, climatic conditions affecting the degree and rate of weathering, and testing procedures used to measure weathering susceptibilities have been reviewed. Sampling and testing techniques, as well as analyses of physical and chemical weathering susceptibilities, are necessary to evaluate rock durability. Many potential riprap materials may not be able to survive 1000 years of weathering. Available techniques for durability testing cannot adequately predict rock durability for the 1000-year period because they do not consider the issue of time (i.e., how long must riprap remain stable). This report includes an Appendix, which discusses rock weathering, written by Dr. Richard Jahns of Stanford University.

  13. Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

    International Nuclear Information System (INIS)

    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 UO2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li2O 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

  14. Carbide-Free Bainitic Weld Metal: A New Concept in Welding of Armor Steels

    Science.gov (United States)

    Krishna Murthy, N.; Janaki Ram, G. D.; Murty, B. S.; Reddy, G. M.; Rao, T. J. P.

    2014-12-01

    Carbide-free bainite, a fine mixture of bainitic ferrite and austenite, is a relatively recent development in steel microstructures. Apart from being very strong and tough, the microstructure is hydrogen-tolerant. These characteristics make it well-suited for weld metals. In the current work, an armor-grade quenched and tempered steel was welded such that the fusion zone developed a carbide-free bainitic microstructure. These welds showed very high joint efficiency and ballistic performance compared to those produced, as per the current industrial practice, using austenitic stainless steel fillers. Importantly, these welds showed no vulnerability to cold cracking, as verified using oblique Y-groove tests. The concept of carbide-free bainitic weld metal thus promises many useful new developments in welding of high-strength steels.

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

    International Nuclear Information System (INIS)

    The key functions required for tritium breeding blankets for a fusion power reactor are: (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li2O/H2O/Be, Mo-alloy/Li2O/He/Be, Mo-alloy/LiAlO2/He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. (author)

  16. 48 CFR 225.7011 - Restriction on carbon, alloy, and armor steel plate.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 3 2010-10-01 2010-10-01 false Restriction on carbon, alloy, and armor steel plate. 225.7011 Section 225.7011 Federal Acquisition Regulations System DEFENSE... on carbon, alloy, and armor steel plate....

  17. Stability formula for armor units to tsunami on rubble mound breakwater

    International Nuclear Information System (INIS)

    The mechanism of the failure of armor units of the rubble mound breakwater due to tsunami was investigated based on the experiments. Tsunamis with leading wave, negative leading wave and soliton fissions were the target tsunamis in the present work. Two different sizes of the wave flumes were used in the experiments. A shorter wave flume is 76m long, 1.2m deep and 0.90m wide and the longer one 205m long, 6.0m deep and 3.4m wide. The mass of the armor unit used ranges from 14g, 30g, 50g, 112g, 240g to 400g. The stability formula was derived from the balance of the fluid force acting armor units and the resistance force of the armor unit in the armor layer. The two coefficients such as the drag coefficient of the armor unit and the friction coefficient are determined to give the stability condition. The relationship between the mass of armor units and the maximum velocity on the crest of the rubble mound breakwater was obtained to determine the conditions of no damage of the armor layer. The numerical simulation method to estimate the velocity on the crest of the rubble mound breakwater was described in the present work. The numerical simulation results were verified with the experimental results of the time histories of the water depth and the horizontal component of the velocity and a good agreements was confirmed. (author)

  18. Field Emissions from Organic Nanorods Armored with Metal Nanoparticles

    Science.gov (United States)

    Suzuki, Toshiya; Ishikawa, Kenji; Takeda, Keigo; Kondo, Hiroki; Sekine, Makoto; Hori, Masaru

    2013-12-01

    We report the fabrication of organic nanorods with a diameter of approximately 10 nm and a height of 106.8 nm (a high aspect ratio of 10.5) armored by Pt nanoparticles. Our results demonstrate that Pt particles deposited by metalorganic supercritical chemical fluid deposition (MOCFD) covering the entire deposition area play important roles in not only etch resistance, especially in protecting the sidewalls, but also the formation of electroconductive Pt/C composites, which were found to have field emission properties.

  19. Diffusive heat blanketing envelopes of neutron stars

    CERN Document Server

    Beznogov, M V; Yakovlev, D G

    2016-01-01

    We construct new models of outer heat blanketing envelopes of neutron stars composed of binary ion mixtures (H - He, He - C, C - Fe) in and out of diffusive equilibrium. To this aim, we generalize our previous work on diffusion of ions in isothermal gaseous or Coulomb liquid plasmas to handle non-isothermal systems. We calculate the relations between the effective surface temperature Ts and the temperature Tb at the bottom of heat blanketing envelopes (at a density rhob= 1e8 -- 1e10 g/cc) for diffusively equilibrated and non-equilibrated distributions of ion species at different masses DeltaM of lighter ions in the envelope. Our principal result is that the Ts - Tb relations are fairly insensitive to detailed distribution of ion fractions over the envelope (diffusively equilibrated or not) and depend almost solely on DeltaM. The obtained relations are approximated by analytic expressions which are convenient for modeling the evolution of neutron stars.

  20. A Precambrian proximal ejecta blanket from Scotland

    Science.gov (United States)

    Amor, Kenneth; Hesselbo, Stephen P.; Porcelli, Don; Thackrey, Scott; Parnell, John

    2008-04-01

    Ejecta blankets around impact craters are rarely preserved onEarth. Although impact craters are ubiquitous on solid bodiesthroughout the solar system, on Earth they are rapidly effaced,and few records exist of the processes that occur during emplacementof ejecta. The Stac Fada Member of the Precambrian Stoer Groupin Scotland has previously been described as volcanic in origin.However, shocked quartz and biotite provide evidence for high-pressureshock metamorphism, while chromium isotope values and elevatedabundances of platinum group metals and siderophile elementsindicate addition of meteoritic material. Thus, the unit isreinterpreted here as having an impact origin. The ejecta blanketreaches >20 m in thickness and contains abundant dark green,vesicular, devitrified glass fragments. Field observations suggestthat the deposit was emplaced as a single fluidized flow thatformed as a result of an impact into water-saturated sedimentarystrata. The continental geological setting and presence of groundwatermake this deposit an analogue for Martian fluidized ejecta blankets.

  1. Water-cooled lithium-lead blanket

    International Nuclear Information System (INIS)

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The present study examines whether the water-cooled lithium-lead blanket designed for NET can be directly extrapolated to a demonstration (DEMO) reactor. A fundamental requirement of the exercise is that the DEMO design should have a tritium breeding ratio which is higher than that in NET. The water-cooled lithium-lead blanket is discussed with respect to: neutronics design, design parameter survey and thermohydraulics, and engineering design. Results are reported of three-dimensional calculations using the Monte Carlo code MORSE-H to investigate possible neutron leakage between the poloidally disposed breeder tubes, and to determine the global tritium breeding ratio for the final double null machine design. (U.K.)

  2. Stellar model atmospheres with magnetic line blanketing

    CERN Document Server

    Kochukhov, O; Shulyak, D

    2004-01-01

    Model atmospheres of A and B stars are computed taking into account magnetic line blanketing. These calculations are based on the new stellar model atmosphere code LLModels which implements direct treatment of the opacities due to the bound-bound transitions and ensures an accurate and detailed description of the line absorption. The anomalous Zeeman effect was calculated for the field strengths between 1 and 40 kG and a field vector perpendicular to the line of sight. The model structure, high-resolution energy distribution, photometric colors, metallic line spectra and the hydrogen Balmer line profiles are computed for magnetic stars with different metallicities and are discussed with respect to those of non-magnetic reference models. The magnetically enhanced line blanketing changes the atmospheric structure and leads to a redistribution of energy in the stellar spectrum. The most noticeable feature in the optical region is the appearance of the 5200 A depression. However, this effect is prominent only in ...

  3. Magnetic non-destructive evaluation of ruptures of tensile armor in oil risers

    Science.gov (United States)

    Pérez-Benitez, J. A.; Padovese, L. R.

    2012-04-01

    Risers are flexible multilayered pipes formed by an inner flexible metal structure surrounded by polymer layers and spiral wound steel ligaments, also known as armor wires. Since these risers are used to link subsea pipelines to floating oil and gas production installations, and their failure could produce catastrophic consequences, some methods have been proposed to monitor the armor integrity. However, until now there is no practical method that allows the automatic non-destructive detection of individual armor wire rupture. In this work we show a method using magnetic Barkhausen noise that has shown high efficiency in the detection of armor wire rupture. The results are examined under the cyclic and static load conditions of the riser. This work also analyzes the theory behind the singular dependence of the magnetic Barkhausen noise on the applied tension in riser armor wires.

  4. DNA barcodes for two scale insect families, mealybugs (Hemiptera: Pseudococcidae) and armored scales (Hemiptera: Diaspididae).

    Science.gov (United States)

    Park, D-S; Suh, S-J; Hebert, P D N; Oh, H-W; Hong, K-J

    2011-08-01

    Although DNA barcode coverage has grown rapidly for many insect orders, there are some groups, such as scale insects, where sequence recovery has been difficult. However, using a recently developed primer set, we recovered barcode records from 373 specimens, providing coverage for 75 species from 31 genera in two families. Overall success was >90% for mealybugs and >80% for armored scale species. The G·C content was very low in most species, averaging just 16.3%. Sequence divergences (K2P) between congeneric species averaged 10.7%, while intra-specific divergences averaged 0.97%. However, the latter value was inflated by high intra-specific divergence in nine taxa, cases that may indicate species overlooked by current taxonomic treatments. Our study establishes the feasibility of developing a comprehensive barcode library for scale insects and indicates that its construction will both create an effective system for identifying scale insects and reveal taxonomic situations worthy of deeper analysis.

  5. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  6. Tritium recovery in Pb17Li-water cooled blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Malara, C. [Safety Technology Inst., Ispra (Italy); Casini, G. [Systems Engineering & Information Inst., Ispra (Italy); Viola, A. [Univ. of Cagliari (Italy)

    1994-12-31

    The question of tritium recovery in Pb17Li, water cooled blankets is under investigation since several years at JRC Ispra. The method which has been more extensively analyzed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging in a suited process apparatus. The design features of the process systems are related to: (1) the very low tritium solubility in Pb17Li which implies high permeation rates through the containment structures; (2) the need of keeping as low as possible the tritium concentration in the cooling water both for safety and economical reasons. A computerized model of the tritium behavior in the blanket units and in the extraction system has been developed.

  7. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Lalitha Jayaraman

    2010-01-01

    Full Text Available This paper presents the application of an analytical tool to quantify material consistency of offset printing blankets. Printing blankets are essentially viscoelastic rubber composites of several laminas. High levels of material consistency are expected from rubber blankets for quality print and for quick recovery from smash encountered during the printing process. The present study aims at determining objectively the consistency of printing blankets at three specific torque levels of tension under two distinct stages; 1. under normal printing conditions and 2. on recovery after smash. The experiment devised exhibits a variation in tone reproduction properties of each blanket signifying the levels of inconsistency also in thicknessdirection. Correlation technique was employed on ink density variations obtained from the blanket on paper. Both blankets exhibited good consistency over three torque levels under normal printing conditions. However on smash the recovery of blanket and its consistency was a function of manufacturing and torque levels. This study attempts to provide a new metrics for failure analysis of offset printing blankets. It also underscores the need for optimizing the torque for blankets from different manufacturers.

  8. Detection of Breeding Blankets Using Antineutrinos

    Science.gov (United States)

    Cogswell, Bernadette; Huber, Patrick

    2016-03-01

    The Plutonium Management and Disposition Agreement between the United States and Russia makes arrangements for the disposal of 34 metric tons of excess weapon-grade plutonium. Under this agreement Russia plans to dispose of its excess stocks by processing the plutonium into fuel for fast breeder reactors. To meet the disposition requirements this fuel would be burned while the fast reactors are run as burners, i.e., without a natural uranium blanket that can be used to breed plutonium surrounding the core. This talk discusses the potential application of antineutrino monitoring to the verification of the presence or absence of a breeding blanket. It is found that a 36 kg antineutrino detector, exploiting coherent elastic neutrino-nucleus scattering and made of silicon, could determine the presence of a breeding blanket at a liquid sodium cooled fast reactor at the 95% confidence level within 90 days. Such a detector would be a novel non-intrusive verification tool and could present a first application of coherent elastic neutrino-nucleus scattering to a real-world challenge.

  9. TCT hybrid preconceptual blanket design studies

    Energy Technology Data Exchange (ETDEWEB)

    Aase, D.T.; Bampton, M.C.C.; Doherty, T.J.; Leonard, B.R.; McCann, R.A.; Newman, D.F.; Perry, R.T.; Stewart, C.W.

    1978-01-01

    The conceptual design of a tokamak fusion-fission (hybrid) reactor, which produces electric power and fissile material, has been performed in a cooperative effort between Princeton's Plasma Physics Laboratory (PPPL) and Battelle's Pacific Northwest Laboratories (PNL). PPPL, who had overall project lead responsibility, designed the fusion driver system. Its core consists of a tokamak plasma maintained in the two-component torus (TCT) mode by both D and T beams and having a single null poloidal divertor. The blanket concept selected by PPPL consists of a neutron multiplying converter region, containing natural Uranium Molybdenum (U-Mo) slugs followed by a fuel burning blanket region of molten salt containing PuF/sub 3/. PNL analyzed this concept to determine its structural, thermal and hydraulic performance characteristics. An adequate first wall cooling method was determined, utilizing low pressure water in a double wall design. A conceptual layout of the converter region tubes was performed, providing adequate helium cooling and the desired movement of U-Mo slugs. A thermal hydraulic analysis of the power-producing blanket regions indicated that either more helium coolant tubes are needed or the salt must be circulated to obtain adequate heat removal capability.

  10. A study on the enhancement of the reliability in gravure offset roll printing with blanket swelling control

    Science.gov (United States)

    Eul Kim, Ga; Woo, Kyoohee; Kang, Dongwoo; Jang, Yunseok; Choi, Young-Man; Lee, Moon G.; Lee, Taik-Min; Kwon, Sin

    2016-10-01

    In roll-offset printing (patterning) technology with a PDMS blanket as a transfer medium, one of the major reliability issues is the occurrence of swelling, which involves absorption of the ink solvent in the printing blanket with repeated printing. This study developed a method to resolve blanket swelling in gravure offset roll printing and performed experiments for performance verification. The physical phenomena of mass and heat transfer were applied to fabricate a device based on convection drying. The proposed device managed to effectively control blanket swelling through drying by blowing air and additional temperature control. The experiments verified that printing quality (in particular the variation of the width of printed patterns) was maintained over 500 continuous printing.

  11. Preliminary In Vivo Evaluation of a Hybrid Armored Vascular Graft Combining Electrospinning and Additive Manufacturing Techniques

    Science.gov (United States)

    Spadaccio, Cristiano; Nappi, Francesco; De Marco, Federico; Sedati, Pietro; Sutherland, Fraser W.H.; Chello, Massimo; Trombetta, Marcella; Rainer, Alberto

    2016-01-01

    In this study, we tested in vivo effectiveness of a previously developed poly-l-lactide/poly-ε-caprolactone armored vascular graft releasing heparin. This bioprosthesis was designed in order to overcome the main drawbacks of tissue-engineered vascular grafts, mainly concerning poor mechanical properties, thrombogenicity, and endothelialization. The bioprosthesis was successfully implanted in an aortic vascular reconstruction model in rabbits. All grafts implanted were patent at four weeks postoperatively and have been adequately populated by endogenous cells without signs of thrombosis or structural failure and with no need of antiplatelet therapy. The results of this preliminary study might warrant for further larger controlled in vivo studies to further confirm these findings. PMID:26949333

  12. Numerical simulation of high-speed penetration-perforation dynamics in layered armor shields

    CERN Document Server

    Ayzenberg-Stepanenko, Mark

    2012-01-01

    Penetration models and calculating algorithms are presented, describing the dynamics and fracture of composite armor shields penetrated by high-speed small arms. A shield considered consists of hard (metal or ceramic) facing and multilayered fabric backing. A simple formula is proved for the projectile residual velocity after perforation of a thin facing. A new plastic-flow jet model is proposed for calculating penetration dynamics in the case of a thick facing of ceramic or metal-ceramic FGM materials. By bringing together the developed models into a calculating algorithm, a computer tool is designed enabling simulations of penetration processes in the above-mentioned shields and analysis of optimization problems. Some results of computer simulation are presented. It is revealed in particular that strength proof of pliable backing can be better as compared with more rigid backing. Comparison of calculations and test data shows sufficient applicability of the models and the tool.

  13. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    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

  14. Fast-Breeder-Blanket Project: FBBF. Final report

    International Nuclear Information System (INIS)

    This report is the final report for DOE contract DE-AC02-76ET37237 with the Purdue Fast Breeder Blanket Project. The Project was initiated to investigate the uncertainties in Fast Breeder Reactor blanket calculations. Absolute measurements of key neutron reaction rates, neutron spectra, and gamma-ray energy depositions were made in simulated FBF blankets in the Fast Breeder Blanket Facility (FBBF), a Cf-252 driven subcritical facility. Calculation of the spectra and integral reaction rates were made using methods, computer codes, and cross section data typical of those currently used in the design of FBR's. Comparisons of calculated to experimental integral neutron reaction rates give good agreement at the inner portions of the blanket by diverge to C/E ratios of about 0.65 at the outer edge of the blanket for reactions sensitive to the neutron density

  15. A review of fusion breeder blanket technology, part 1

    International Nuclear Information System (INIS)

    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

  16. Natural Circulation in the Blanket Heat Removal System During a Loss-of-Pumping Accident (LOFA) Based on Initial Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    A transient natural convection model of the APT blanket primary heat removal (HR) system was developed to demonstrate that the blanket could be cooled for a sufficient period of time for long term cooling to be established following a loss-of-flow accident (LOFA). The particular case of interest in this report is a complete loss-of-pumping accident. For the accident scenario in which pumps are lost in both the target and blanket HR systems, natural convection provides effective cooling of the blanket for approximately 68 hours, and, if only the blanket HR systems are involved, natural convection is effective for approximately 210 hours. The heat sink for both of these accident scenarios is the assumed stagnant fluid and metal on the secondary sides of the heat exchangers.

  17. Finite Element Modeling of the Behavior of Armor Materials Under High Strain Rates and Large Strains

    Science.gov (United States)

    Polyzois, Ioannis

    For years high strength steels and alloys have been widely used by the military for making armor plates. Advances in technology have led to the development of materials with improved resistance to penetration and deformation. Until recently, the behavior of these materials under high strain rates and large strains has been primarily based on laboratory testing using the Split Hopkinson Pressure Bar apparatus. With the advent of sophisticated computer programs, computer modeling and finite element simulations are being developed to predict the deformation behavior of these metals for a variety of conditions similar to those experienced during combat. In the present investigation, a modified direct impact Split Hopkinson Pressure Bar apparatus was modeled using the finite element software ABAQUS 6.8 for the purpose of simulating high strain rate compression of specimens of three armor materials: maraging steel 300, high hardness armor (HHA), and aluminum alloy 5083. These armor materials, provided by the Canadian Department of National Defence, were tested at the University of Manitoba by others. In this study, the empirical Johnson-Cook visco-plastic and damage models were used to simulate the deformation behavior obtained experimentally. A series of stress-time plots at various projectile impact momenta were produced and verified by comparison with experimental data. The impact momentum parameter was chosen rather than projectile velocity to normalize the initial conditions for each simulation. Phenomena such as the formation of adiabatic shear bands caused by deformation at high strains and strain rates were investigated through simulations. It was found that the Johnson-Cook model can accurately simulate the behavior of body-centered cubic (BCC) metals such as steels. The maximum shear stress was calculated for each simulation at various impact momenta. The finite element model showed that shear failure first occurred in the center of the cylindrical specimen and

  18. Analysis of deficiencies in fast reactor blanket physics predictions

    International Nuclear Information System (INIS)

    This analysis addresses a deviation between experimental measurements and fast reactor blanket physics predictions. A review of worldwide results reveals that reaction rates in the blanket are underpredicted with the discrepancy increasing with penetration into the blanket. The analysis of this discrepancy involves two parts: quantifying possible error reductions using the most advanced methods and investigating deficiencies in current methodology. The source of these discrepancies was investigated by application of ''state-of-the-art'' group constant generation and flux prediction methodology to flux calculations for the Purdue University Fast Breeder Blanket Facility (FBBF). Refined group constant generation methods yielded a significant reduction in the blanket deviations; however, only about half of the discrepancy can be accounted for in this manner. Transport theory calculations were used to predict the blanket neutron transmission problem. The surprising result is that transport theory predictions utilizing diffusion theory group constants did not improve the blanket results. Transport theory predictions exhibited blanket underpredictions similar to the diffusion theory results. The residual blanket discrepancies not explained using advanced methods require a refinement of the theory. For this purpose an analysis of deficiencies in current methodology was performed

  19. The effects of exercise and body armor on cognitive function in healthy volunteers.

    Science.gov (United States)

    Roberts, Aaron P J; Cole, Jon C

    2013-05-01

    Police officers routinely wear body armor to protect themselves against the threat posed by firearms and edged weapons, yet little is known of the cognitive effects of doing so. Two studies investigated the effects of exercise and body armor on working memory function in healthy volunteers. In study 1, male undergraduates were assigned to one of four groups: (i) brief exercise, (ii) brief exercise wearing body armor, (iii) extended exercise, and (iv) extended exercise wearing body armor. In study 2, university gym members were assigned to one of two groups: (i) wearing body armor and (ii) not wearing body armor. In both studies, heart rate and oral temperature were measured before, immediately after, and 5 minutes after exercise. The phonemic verbal fluency task and digits backward test were administered at the same time points. In both studies, a mixed analysis of variance revealed statistically significant changes to the cognitive functioning of participants. A change in cognitive strategy was observed, reflected by a decrease in executive function (switches) and an increase in nonexecutive function (cluster size). These data suggest that the cognitive effects of exercise and body armor may have profound implications for police officers' ability to make tactical decisions. PMID:23756004

  20. Neural network modeling to evaluate the dynamic flow stress of high strength armor steels under high strain rate compression

    Directory of Open Access Journals (Sweden)

    Ravindranadh Bobbili

    2014-12-01

    Full Text Available An artificial neural network (ANN constitutive model is developed for high strength armor steel tempered at 500 °C, 600 °C and 650 °C based on high strain rate data generated from split Hopkinson pressure bar (SHPB experiments. A new neural network configuration consisting of both training and validation is effectively employed to predict flow stress. Tempering temperature, strain rate and strain are considered as inputs, whereas flow stress is taken as output of the neural network. A comparative study on Johnson–Cook (J–C model and neural network model is performed. It was observed that the developed neural network model could predict flow stress under various strain rates and tempering temperatures. The experimental stress–strain data obtained from high strain rate compression tests using SHPB, over a range of tempering temperatures (500–650 °C, strains (0.05–0.2 and strain rates (1000–5500/s are employed to formulate J–C model to predict the high strain rate deformation behavior of high strength armor steels. The J-C model and the back-propagation ANN model were developed to predict the high strain rate deformation behavior of high strength armor steel and their predictability is evaluated in terms of correlation coefficient (R and average absolute relative error (AARE. R and AARE for the J–C model are found to be 0.7461 and 27.624%, respectively, while R and AARE for the ANN model are 0.9995 and 2.58%, respectively. It was observed that the predictions by ANN model are in consistence with the experimental data for all tempering temperatures.

  1. Neural network modeling to evaluate the dynamic flow stress of high strength armor steels under high strain rate compression

    Institute of Scientific and Technical Information of China (English)

    Ravindranadh BOBBILI; V. MADHU; A.K. GOGIA

    2014-01-01

    An artificial neural network (ANN) constitutive model is developed for high strength armor steel tempered at 500 ?C, 600 ?C and 650 ?C based on high strain rate data generated from split Hopkinson pressure bar (SHPB) experiments. A new neural network configuration consisting of both training and validation is effectively employed to predict flow stress. Tempering temperature, strain rate and strain are considered as inputs, whereas flow stress is taken as output of the neural network. A comparative study on JohnsoneCook (JeC) model and neural network model is performed. It was observed that the developed neural network model could predict flow stress under various strain rates and tempering temperatures. The experimental stressestrain data obtained from high strain rate compression tests using SHPB, over a range of tempering temperatures (500e650 ?C), strains (0.05e0.2) and strain rates (1000e5500/s) are employed to formulate JeC model to predict the high strain rate deformation behavior of high strength armor steels. The J-C model and the back-propagation ANN model were developed to predict the high strain rate deformation behavior of high strength armor steel and their predictability is evaluated in terms of correlation coefficient (R) and average absolute relative error (AARE). R and AARE for the JeC model are found to be 0.7461 and 27.624%, respectively, while R and AARE for the ANN model are 0.9995 and 2.58%, respectively. It was observed that the predictions by ANN model are in consistence with the experimental data for all tempering temperatures.

  2. Overview of requirements and design integration for the ITER EU Test Blanket Systems instrumentation

    International Nuclear Information System (INIS)

    The ITER project aims at building a fusion device with the general goal of demonstrating the scientific and technical feasibility of fusion power. The testing of Tritium Breeder Blanket concepts is one of the ITER missions and has been recognized as an essential milestone in the development of a future fusion reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The strategy for the development of the instrumentation of the HCLL and HCPB Test Blanket Systems, which include the TBMs and their Ancillary Systems, is briefly recalled in this paper, along with the overview of the requirements coming from the harsh operational environment and the main challenges related to the integration with the complex design of the TBS components. (authors)

  3. Imperfection analysis of flexible pipe armor wires in compression and bending

    DEFF Research Database (Denmark)

    Østergaard, Niels Højen; Lyckegaard, Anders; Andreasen, Jens H.

    2012-01-01

    The work presented in this paper is motivated by a specific failure mode known as lateral wire buckling occurring in the tensile armor layers of flexible pipes. The tensile armor is usually constituted by two layers of initially helically wound steel wires with opposite lay directions. During pipe...... laying in ultra deep waters, a flexible pipe experiences repeated bending cycles and longitudinal compression. These loading conditions are known to impose a danger to the structural integrity of the armoring layers, if the compressive load on the pipe exceeds the total maximum compressive load carrying...

  4. [To the hygienic standardization of the stroke of shock elements through the armored clothes].

    Science.gov (United States)

    Logatkin, S M

    2008-01-01

    The author considers major criteria for assessing the armored clothes by the armored dynamic force in the impenetrability of shocking elements. The used material simulators and the methods of studies are critically analyzed. The analysis has led to the conclusion that deformity depth of human soft tissue stimulators (clay, plasticine, gelatin blocks), used to evaluate the quality of armored clothes) may not be an indicator for the hygienic standardization of the stroke of shock elements through a barrier if the time of interaction is not taken into account. PMID:18368708

  5. Analysis of Tile-Reinforced Composite Armor. Part 1; Advanced Modeling and Strength Analyses

    Science.gov (United States)

    Davila, C. G.; Chen, Tzi-Kang; Baker, D. J.

    1998-01-01

    The results of an analytical and experimental study of the structural response and strength of tile-reinforced components of the Composite Armored Vehicle are presented. The analyses are based on specialized finite element techniques that properly account for the effects of the interaction between the armor tiles, the surrounding elastomers, and the glass-epoxy sublaminates. To validate the analytical predictions, tests were conducted with panels subjected to three-point bending loads. The sequence of progressive failure events for the laminates is described. This paper describes the results of Part 1 of a study of the response and strength of tile-reinforced composite armor.

  6. Characteristics of behind armor blunt trauma produced by bullets with different structural materials: an experimental study

    OpenAIRE

    Ling-qing WANG; Xi-nan LAI; Zhang, Bo; Zheng-lin SU; Huang, Yi-Feng; Wang, Li-Li

    2013-01-01

    Objective To investigate the effect of structural materials of bullets on behind armor blunt trauma (BABT). Methods Ten healthy male Landraces were randomly divided into two groups (5 each): 56 type 7.62-mm rifle bullet group and SS109 5.56-mm rifle bullet group. The kinetic energy of two types of bullets was adjusted to the same level (about 1880J) by the way of grow downwards gunpowder. Then the animals as protected with both grade NIJ Ⅲ ceramic hard armor and grade Ⅱ police soft body armor...

  7. Armor stability of hardly (or partly) reshaping berm breakwaters

    DEFF Research Database (Denmark)

    Moghim, M. N.; Andersen, Thomas Lykke

    2015-01-01

    the structure toe. The main goal of the present paper is to provide an estimation technique based on this physical principle to predict the deformation of the front slope in terms of the eroded area. The proposed method is verified by comparison with model test data. It is found that by using the maximum wave......This paper deals with stability of hardly (or partly) reshaping berm breakwaters. A simple physical argument is used to derive a new stability formula based on the assumption that the maximum wave force causing damage of armor layer is proportional to the maximum wave momentum flux near...... momentum flux approach the damage to the front slope (eroded area) can be very well predicted. Moreover, a simple method to estimate the eroded area based on measured or calculated berm recession (Rec) and depth of intersection of reshaped and initial profile (hf) is presented. The performance...

  8. Laser-induced micro-jetting from armored droplets

    KAUST Repository

    Marston, J. O.

    2015-06-23

    We present findings from an experimental study of laser-induced cavitation within a liquid drop coated with a granular material, commonly referred to as “armored droplets” or “liquid marbles.” The cavitation event follows the formation of plasma after a nanosecond laser pulse. Using ultra-high-speed imaging up to 320,610 fps, we investigate the extremely rapid dynamics following the cavitation, which manifests itself in the form of a plethora of micro-jets emanating simultaneously from the spaces between particles on the surface of the drop. These fine jets break up into droplets with a relatively narrow diameter range, on the order of 10 μm. © 2015, Springer-Verlag Berlin Heidelberg.

  9. River bed armoring in a local scour under no-supply conditions; experimental investigation and numerical model validation

    Science.gov (United States)

    Török, Gergerly; Baranya, Sandor; Rüther, Nils

    2016-04-01

    The aim of this study is to present a novel method for numerical modeling of morphological changes. The essence of the method doesn't mean the development of a new sediment transport formula, but the combined application of the existing, conditionally validated sediment transport models. Many bedload transport formulas can be found in the literature, which were developed based on different field and laboratory measurements. Thus, the most reliability of the models usually can be expected only for the given morphological and hydrological conditions connected to the base measurements. However, commonly in the analysed cases the morphological and hydrological features are more variable both in time and in space. Therefore, the hypothesis of this study is that, complex hydromorphological processes can't be modeled by one sediment transport formula. The authors present a solution based on laboratory experiments. Spatio-temporal developments of bed armoring, local scouring and local sediment deposition under no supply condition was monitored and analysed. The sediment transport model of Wilcock and Crowe (2003) was expected to calculate properly the local scouring and bed armoring processes, while the motion and aggradation of the finer materials were supposed to estimate reliably by the van Rijn formula (1984). The main challenge of the combining method is to find an appropriate criterion to decide which transport formula is activated in the given space and in the given time step. The result of the investigation showed that the most reliable criteria is based on the d50 value. As soon as the d50 grain size goes below a certain value, van Rijn is activated, otherwise the Wilcock and Crowe formula calculates the sorting and armoring processes. The results show that the combining method clearly improve the reliability of the morphological calculation. The benefit of the Wilcock and Crowe model is that it estimates quite well the sediment transport in mixed or armored

  10. Options and methods for instrumentation of Test Blanket Systems for experiment control and scientific mission

    International Nuclear Information System (INIS)

    Highlights: • This work defined options and methods to instrument ITER TBSs based on functional categories: safety, interlock and control and scientific exploitation based on the ITER research program. • Presented the general architecture of the HCLL and HCPB Test Blanket System Instrumentation and Control. • Defined safety and interlock sensors count and technology selection based on preliminary safety analysis. • Discussed the development status of scientific instrumentation, with focus on integration with design and fulfillment of TBM research program. - Abstract: Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required for their operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives

  11. Shock imprint and rolling direction influence upon the breaking tenacity for 2P armor steel

    Science.gov (United States)

    Zichil, V.; Coseru, A.; Schnakovszky, C.; Herghelegiu, E.; Radu, C.

    2016-08-01

    The state of art in present literature shows that the breaking tenacity of a material is influenced by the integrity of the structure. Since armors used in aviation and to protect military vehicles are frequently impact loaded, through the contact between armor sheet and projectiles, or other foreign bodies, the authors have proposed to study the dependence between the breaking tenacity of 2P armor steel depending on the direction of the rolling of the armor plate, of the geometry (spherical imprint, pyramidal and linear imprint) and the depth of the deformation that results after impact. Tests were conducted upon CT (ASTM E- 399) specimen type, using the critical factor of stress intensity during the state of planar strain.

  12. MIT LMFBR blanket research project. Final summary report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    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

  14. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    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 238U 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

  15. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li2TiO3, tritium release behavior of Li2TiO3 and Li2ZrO3 including tritium diffusion, modeling of tritium release from Li2ZrO3 in ITER condition, helium release behavior from Li2O, results of tritium release irradiation tests of Li4SiO4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  16. Nacre-like ceramic/polymer laminated composite for use in body-armor applications

    Directory of Open Access Journals (Sweden)

    Mica Grujicic

    2016-01-01

    Full Text Available Nacre is a biological material constituting the innermost layer of the shells of gastropods and bivalves. It consists of polygonal tablets of aragonite, tessellated to form individual layers and having the adjacent layers as well as the tablets within a layer bonded by a biopolymer. Due to its highly complex hierarchical microstructure, nacre possesses an outstanding combination of mechanical properties, the properties which are far superior to the ones that are predicted using the techniques such as the rule of mixture. In the present work, an attempt is made to model a nacre-like composite armor consisting of boron carbide (B4C tablets and polyurea tablet/tablet interfaces. The armor is next investigated with respect to impact by a solid right-circular-cylindrical rigid projectile, using a transient non-linear dynamics finite element analysis. The ballistic-impact response and the penetration resistance of the armor is then compared with that of the B4C monolithic armor having an identical areal density. Furthermore, the effect of various nacre microstructural features (e.g. surface profiling, micron-scale asperities, mineral bridges between the overlapping tablets lying in adjacent layers, and B4C nano-crystallinity on the ballistic-penetration resistance of the composite-armor is investigated in order to identify an optimal nacre-like composite-armor architecture having the largest penetration resistance. The results obtained clearly show that a nacre-like armor possesses a superior penetration resistance relative to its monolithic counterpart, and that the nacre microstructural features considered play a critical role in the armor penetration resistance.

  17. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, B. William [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chiu, Ing L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-26

    Future fusion power plants based on the deuterium-tritium (DT) fuel cycle will be required to breed the T fuel via neutron reactions with lithium, which will be incorporated in a breeding blanket that surrounds the fusion source. Recent work by LLNL proposed the used of liquid Li as the breeder in an inertial fusion energy (IFE) power plant. Subsequently, an LDRD was initiated to develop alternatives ternary alloy liquid metal breeders that have reduced chemical reactivity with water and air compared to pure Li. Part of the work plan was to experimentally investigate the phase diagrams of ternary alloys. Of particular interest was measurement of the melt temperature, which must be low enough to be compatible with the temperature limits of the steel used in the construction of the chamber and heat transfer system.

  18. Dynamics analysis of planar armored cable motion in deep-sea ROV system

    Institute of Scientific and Technical Information of China (English)

    全伟才; 张竺英; 张艾群

    2014-01-01

    The armored cable used in a deep-sea remotely operated vehicle (ROV) may undergo large displacement motion when subjected to dynamic actions of ship heave motion and ocean current. A novel geometrically exact finite element model for two-dimensional dynamic analysis of armored cable is presented. This model accounts for the geometric nonlinearities of large displacement of the armored cable, and effects of axial load and bending stiffness. The governing equations are derived by consistent linearization and finite element discretization of the total weak form of the armored cable system, and solved by the Newmark time integration method. To make the solution procedure avoid falling into the local extreme points, a simple adaptive stepping strategy is proposed. The presented model is validated via actual measured data. Results for dynamic configurations, motion and tension of both ends of the armored cable, and resonance-zone are presented for two numerical cases, including the dynamic analysis under the case of only ship heave motion and the case of joint action of ship heave motion and ocean current. The dynamics analysis can provide important reference for the design or product selection of the armored cable in a deep-sea ROV system so as to improve the safety of its marine operation under the sea state of 4 or above.

  19. Research on Protection Capability of Flying-Whip Multifunctional Explosive Reactive Armor

    Institute of Scientific and Technical Information of China (English)

    张虎生; 张宝(金平); 张庆明; 刘长林

    2004-01-01

    The experimental research on protection capability of the flying-whip multifunctional explosive reactive armor (ERA) was performed, in which the comparison experiment was made on the damage effect of the flying-whip's geometrical figuration, material property and driven velocity on the long-rod armor-piercing-projectile. The moving velocity of the flying-whip driven by different explosives and the pressure attenuation law of shock wave travelling in the back plate were measured respectively with the electric probe method and the manganin piezoresistive gauge technique. The following conclusions based on a great quantity of experimental data were drawn: compared with the sandwich ERA the flying-whip multifunctional ERA has very good protection function against the long-rod armor-piercing-projectile, the shaped charge warhead and the anti-armor tandem warhead. In addition, the composite plate made of the armor-steel and rubber plate can lessen the vibration and shock of the main armor caused by the explosion of the charge.

  20. Initial meetings of the re-established Test Blanket Working Group

    International Nuclear Information System (INIS)

    The ITER Test Blanket Working Group (TBWG) was first established in 1995. Its activities covered successively the final part of the ITER EDA and the extension period, the main results being a preliminary assessment of the breeding blanket testing capabilities of ITER and a proposal of a coherent test blanket programme, reported in 2001, that optimized the sharing of the three available testing ports between the three Parties present in 2001 (EU, JA and RF) taking into account the different coolant characteristics. The TBWG was re-established by the ITER Interim Project Leader in September 2003, with the support of the Participant Team Leaders. It is now comprised of four members from the ITER International Team and up to three members from each of the six ITER Participant Teams. The International Team delegation is led by Dr. V. Chuyanov, who has also been appointed as TBWG Co-Chair, while the six Participant Team delegations are led by Prof. M. Abdou (US), Dr. M. Akiba (JA), Dr. A. Cardella (EU), Dr. B.G. Hong (KO), Dr. C. Pan (CN) and Dr.Y. Strebkov (RF). The revised TBWG charter defines the four missions of the activities: i) provide the Design Description Document (DDD) of the Test Blanket Module (TBM) systems proposed by the participants, including the description of the interfaces with the main ITER machine, ii) promote cooperation among participants on the associated R and D programmes, iii) verify the integration of TBM testing in ITER site safety and environmental evaluations, and finally, iv) develop and propose coordinated TBM test programmes taking into account ITER operation planning. TBMs have to be representative of the breeding blanket for DEMO (the next reactor after ITER), capable of ensuring tritium-breeding self-sufficiency and of accommodating high-grade coolants for electricity production

  1. Spacecraft thermal blanket cleaning: Vacuum bake of gaseous flow purging

    Science.gov (United States)

    Scialdone, John J.

    1990-01-01

    The mass losses and the outgassing rates per unit area of three thermal blankets consisting of various combinations of Mylar and Kapton, with interposed Dacron nets, were measured with a microbalance using two methods. The blankets at 25 deg C were either outgassed in vacuum for 20 hours, or were purged with a dry nitrogen flow of 3 cu. ft. per hour at 25 deg C for 20 hours. The two methods were compared for their effectiveness in cleaning the blankets for their use in space applications. The measurements were carried out using blanket strips and rolled-up blanket samples fitting the microbalance cylindrical plenum. Also, temperature scanning tests were carried out to indicate the optimum temperature for purging and vacuum cleaning. The data indicate that the purging for 20 hours with the above N2 flow can accomplish the same level of cleaning provided by the vacuum with the blankets at 25 deg C for 20 hours, In both cases, the rate of outgassing after 20 hours is reduced by 3 orders of magnitude, and the weight losses are in the range of 10E-4 gr/sq cm. Equivalent mass loss time constants, regained mass in air as a function of time, and other parameters were obtained for those blankets.

  2. DEMO blanket testing in ITER. Influence on reaching DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Shatalov, G. E-mail: geshat@nfi.kiae.ru

    2001-10-01

    ITER goal was specified as one step between now and the DEMO fusion reactor. One of the major issues is the tritium breeding blankets test relevant to future reactors. The major objectives of blanket modules (TBM) experiments in ITER are reduced in comparison with proposed test objectives in ITER-FDR. Thus, results of DEMO blanket designs testing in ITER will provide limited (but still useful) information that will need strong support from non-fusion facilities testing. The role of non-fusion tests is increased now to provide additional data required for DEMO blanket construction and qualification. A strategy of testing steps to DEMO blanket qualifications has to include parallel testing in ITER and in non-fusion devices. Experiments in fission reactors are able to provide essential data on materials radiation properties; tritium release, inventory and permeation; and thermomechanical behavior of the blanket breeder/multiplier. However, the volume in fission reactors is rather small and neutron spectra differ from the fusion reactor one. Nonetheless in the near future one depends primarily on fission reactor irradiation. The powerful accelerator based neutron source IFMIF could also provide useful information on radiation material properties. Plasma based neutron sources of different fusion devices could be the best choice for testing DEMO materials and blanket mock-ups. Timetable and costs of these devices are not clear now.

  3. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    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

  4. Lightweight solar array blanket tooling, laser welding and cover process technology

    Science.gov (United States)

    Dillard, P. A.

    1983-01-01

    A two phase technology investigation was performed to demonstrate effective methods for integrating 50 micrometer thin solar cells into ultralightweight module designs. During the first phase, innovative tooling was developed which allows lightweight blankets to be fabricated in a manufacturing environment with acceptable yields. During the second phase, the tooling was improved and the feasibility of laser processing of lightweight arrays was confirmed. The development of the cell/interconnect registration tool and interconnect bonding by laser welding is described.

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

    International Nuclear Information System (INIS)

    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

  6. Material and design considerations for the carbon armored ITER divertor

    International Nuclear Information System (INIS)

    The properties of materials for the carbon armored ITER divertor were evaluated from literature and manufacturers' documentation. Most of these data, however, have been not known or not published yet. We have evaluated an optimum data set of the candidate materials of the ITER divertor, which were needed for finite element analyses (FEM). The materials evaluated are as follows; MFC-1, CX2002U, SEP-N112, P-130, IG-430U for the carbon based materials, and Oxygen Free Copper (OFCu), Dispersion Strengthened Copper (DSCu), TZM, W5Re and W-Cu as a heat sink material. It should be noted that W-Cu is first proposed for a heat sink application of the ITER divertor plate. The finite element analyses were performed for the residual stress induced by brazing, thermal response and thermal stresses under a uniform heat flux of 15 MW/m2 to the plasma facing surface. The stress free temperature of 750degC is assumed for the residual stress by brazing. Ten different geometries of the divertor were considered in the analyses including possible material combinations. The FEM results show that the material combinations of MFC-1 and W-30Cu or DSUc in the flat-plate geometry satisfy the presently accepted ITER requirements. The combinations of CX2002U and TZM or W5Re is considered a good choice in terms of residual and thermal stresses, whereas the surface temperature exceeds the ITER requirements. (author) 106 refs

  7. Armored kinorhynch-like scalidophoran animals from the early Cambrian.

    Science.gov (United States)

    Zhang, Huaqiao; Xiao, Shuhai; Liu, Yunhuan; Yuan, Xunlai; Wan, Bin; Muscente, A D; Shao, Tiequan; Gong, Hao; Cao, Guohua

    2015-01-01

    Morphology-based phylogenetic analyses support the monophyly of the Scalidophora (Kinorhyncha, Loricifera, Priapulida) and Nematoida (Nematoda, Nematomorpha), together constituting the monophyletic Cycloneuralia that is the sister group of the Panarthropoda. Kinorhynchs are unique among living cycloneuralians in having a segmented body with repeated cuticular plates, longitudinal muscles, dorsoventral muscles, and ganglia. Molecular clock estimates suggest that kinorhynchs may have diverged in the Ediacaran Period. Remarkably, no kinorhynch fossils have been discovered, in sharp contrast to priapulids and loriciferans that are represented by numerous Cambrian fossils. Here we describe several early Cambrian (~535 million years old) kinorhynch-like fossils, including the new species Eokinorhynchus rarus and two unnamed but related forms. E. rarus has characteristic scalidophoran features, including an introvert with pentaradially arranged hollow scalids. Its trunk bears at least 20 annuli each consisting of numerous small rectangular plates, and is armored with five pairs of large and bilaterally placed sclerites. Its trunk annuli are reminiscent of the epidermis segments of kinorhynchs. A phylogenetic analysis resolves E. rarus as a stem-group kinorhynch. Thus, the fossil record confirms that all three scalidophoran phyla diverged no later than the Cambrian Period. PMID:26610151

  8. Fabrication, testing and modeling of a new flexible armor inspired from natural fish scales and osteoderms

    International Nuclear Information System (INIS)

    Crocodiles, armadillo, turtles, fish and many other animal species have evolved flexible armored skins in the form of hard scales or osteoderms, which can be described as hard plates of finite size embedded in softer tissues. The individual hard segments provide protection from predators, while the relative motion of these segments provides the flexibility required for efficient locomotion. In this work, we duplicated these broad concepts in a bio-inspired segmented armor. Hexagonal segments of well-defined size and shape were carved within a thin glass plate using laser engraving. The engraved plate was then placed on a soft substrate which simulated soft tissues, and then punctured with a sharp needle mounted on a miniature loading stage. The resistance of our segmented armor was significantly higher when smaller hexagons were used, and our bio-inspired segmented glass displayed an increase in puncture resistance of up to 70% compared to a continuous plate of glass of the same thickness. Detailed structural analyses aided by finite elements revealed that this extraordinary improvement is due to the reduced span of individual segments, which decreases flexural stresses and delays fracture. This effect can however only be achieved if the plates are at least 1000 stiffer than the underlying substrate, which is the case for natural armor systems. Our bio-inspired system also displayed many of the attributes of natural armors: flexible, robust with ‘multi-hit’ capabilities. This new segmented glass therefore suggests interesting bio-inspired strategies and mechanisms which could be systematically exploited in high-performance flexible armors. This study also provides new insights and a better understanding of the mechanics of natural armors such as scales and osteoderms. (paper)

  9. Fracture mechanical analysis of tungsten armor failure of a water-cooled divertor target

    International Nuclear Information System (INIS)

    Highlights: • The FEM-based VCE method and XFEM were employed for computing KI (or J-integral) and predicting progressive cracking, respectively. • The most probable pattern of crack formation is radial cracking in the tungsten armor block. • The most probable site of cracking is the upper interfacial region of the tungsten armor block adjacent to the top position of the copper interlayer. • The initiation of a major crack becomes likely, only when the strength of tungsten armor block is significantly reduced from its original strength. - Abstract: The inherent brittleness of tungsten at low temperature and the embrittlement by neutron irradiation are its most critical weaknesses for fusion applications. In the current design of the ITER and DEMO divertor, the high heat flux loads during the operation impose a strong constraint on the structure–mechanical performance of the divertor. Thus, the combination of brittleness and the thermally induced stress fields due to the high heat flux loads raises a serious reliability issue in terms of the structural integrity of tungsten armor. In this study, quantitative estimates of the vulnerability of the tungsten monoblock armor cracking under stationary high heat flux loads are presented. A comparative fracture mechanical investigation has been carried out by means of two different types of computational approaches, namely, the extended finite element method (XFEM) and the finite element method (FEM)-based virtual crack tip extension (VCE) method. The fracture analysis indicates that the most probable pattern of crack formation is radial cracking in the tungsten armor starting from the interface to tube and the most probable site of cracking is the upper interfacial region of the tungsten armor adjacent to the top position of the copper interlayer. The strength threshold for crack initiation and the high heat flux load threshold for crack propagation are evaluated based on XFEM simulations and computations of

  10. Effect of channel wall conductance on the performance characteristics of self-cooled liquid metal fusion reactor blankets

    International Nuclear Information System (INIS)

    One of the critical issues in self-cooled liquid metal tritium breeding blankets in magnetically confined fusion reactors is strong MHD effects particularly when the channel walls are not electrically insulated from the flowing liquid metals. Another critical issue is the cooling of the first wall which is subjected to intense heat load from the fusion plasma. In this work we investigate the effect of channel wall conductance on the friction factor and Nusselt number. It is shown by solving the indication and linear momentum equations that even for relatively small channel wall conductance ratios, the friction factor increases by an order of magnitude for the typical Hartmann numbers encountered in fusion reactor blankets. Furthermore, by solving the temperature equation, it is shown that channel wall conductance has negligible effect on Nusselt number in spite of high velocity jets developing near the side walls. Taking into account these limitations, it is shown however, that the self-cooled liquid metal blankets remain a feasible proposition for both first wall heat extraction and bulk heat removal from the blanket. The most important thermal-hydraulic performance parameter -the heat removal rate to pumping power ratio- can still be kept quite high by suitably choosing the design variables of the liquid metal cooling system. The results are presented and compared for the three prime candidates for self-cooled liquid metal breeding blankets, i.e., lithium, lead-lithium, and tin-lithium alloys. (author)

  11. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    Science.gov (United States)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-10-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 °C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 °C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI).

  12. 18 CFR 284.402 - Blanket marketing certificates.

    Science.gov (United States)

    2010-04-01

    ... effective for an affiliated marketer with respect to transactions involving affiliated pipelines when an affiliated pipeline receives its blanket certificate pursuant to § 284.284. (2) Should a marketer...

  13. Lightweight IMM Multi-Junction Photovoltaic Flexible Blanket Assembly Project

    Data.gov (United States)

    National Aeronautics and Space Administration — DSS's recently completed successful NASA SBIR Phase 1 program has established a TRL 3/4 classification for an innovative IMM PV Integrated Modular Blanket Assembly...

  14. ITER [International Thermonuclear Experimental Reactor] shield and blanket work package report

    International Nuclear Information System (INIS)

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs

  15. Mechanical design and analysis for a EPR first wall/blanket/shield system

    International Nuclear Information System (INIS)

    Continuing studies are in progress at ANL to expand upon the design of a first wall/blanket/shield FW/B/S system and power conversion for a tokamak type Experimental Power Reactor (EPR). The FW/B/S system has evolved from an earlier design for a low beta, circular cross section plasma (major radius = 6 m) to one for a higher beta elongated plasma with a 4.7 m major radius. Basic mechanical design and layout features of the old and new EPR designs showing some of the more important design developments are given. These developments are aimed at simplifying the design, reducing the costs and in addition, improving the plant thermal efficiency and overall maintainability. In the area of the reactor blanket, significant thermal hydraulic and stress analysis have been performed to substantiate the integrity of the chosen concept. This paper deals with the discussion of these improved features

  16. Mechanical design and analysis for a EPR first wall/blanket/shield system

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1977-01-01

    Continuing studies are in progress at ANL to expand upon the design of a first wall/blanket/shield FW/B/S system and power conversion for a tokamak type Experimental Power Reactor (EPR). The FW/B/S system has evolved from an earlier design for a low beta, circular cross section plasma (major radius = 6 m) to one for a higher beta elongated plasma with a 4.7 m major radius. Basic mechanical design and layout features of the old and new EPR designs showing some of the more important design developments are depicted. These developments are aimed at simplifying the design, reducing the costs and, in addition, improving the plant thermal efficiency and overall maintainability. In the area of the reactor blanket, significant thermal hydraulic and stress analysis have been performed to substantiate the integrity of the chosen concept. This paper deals with the discussion of these improved features.

  17. Friction Stir Weld Failure Mechanisms in Aluminum-Armor Structures Under Ballistic Impact Loading Conditions

    Science.gov (United States)

    Grujicic, M.; Pandurangan, B.; Arakere, A.; Yen, C.-F.; Cheeseman, B. A.

    2013-01-01

    A critical assessment is carried out of the microstructural changes in respect of the associated reductions in material mechanical properties and of the attendant ballistic-impact failure mechanisms in prototypical friction stir welding (FSW) joints found in armor structures made of high-performance aluminum alloys (including solution-strengthened and age-hardenable aluminum alloy grades). It is argued that due to the large width of FSW joints found in thick aluminum-armor weldments, the overall ballistic performance of the armor is controlled by the ballistic limits of its weld zones (e.g., heat-affected zone, the thermomechanically affected zone, the nugget, etc.). Thus, in order to assess the overall ballistic survivability of an armor weldment, one must predict/identify welding-induced changes in the material microstructure and properties, and the operative failure mechanisms in different regions of the weld. Toward this end, a procedure is proposed in the present study which combines the results of the FSW process modeling, basic physical-metallurgy principles concerning microstructure/property relations, and the fracture mechanics concepts related to the key blast/ballistic-impact failure modes. The utility of this procedure is demonstrated using the case of a solid-solution strengthened and cold-worked aluminum alloy armor FSW-weld test structure.

  18. Modeling, Validation, and Control of Electronically Actuated Pitman Arm Steering for Armored Vehicle

    Directory of Open Access Journals (Sweden)

    Vimal Rau Aparow

    2016-01-01

    Full Text Available In this study, 2 DOF mathematical models of Pitman arm steering system are derived using Newton’s law of motion and modeled in MATLAB/SIMULINK software. The developed steering model is included with a DC motor model which is directly attached to the steering column. The Pitman arm steering model is then validated with actual Pitman arm steering test rig using various lateral inputs such as double lane change, step steer, and slalom test. Meanwhile, a position tracking control method has been used in order to evaluate the effectiveness of the validated model to be implemented in active safety system of a heavy vehicle. The similar method has been used to test the actual Pitman arm steering mechanism using hardware-in-the-loop simulation (HILS technique. Additional friction compensation is added in the HILS technique in order to minimize the frictional effects that occur in the mechanical configuration of the DC motor and Pitman arm steering. The performance of the electronically actuated Pitman arm steering system can be used to develop a firing-on-the-move actuator (FOMA for an armored vehicle. The FOMA can be used as an active safety system to reject unwanted yaw motion due to the firing force.

  19. Blankets for tritium catalyzed deuterium (TCD) fusion reactors

    International Nuclear Information System (INIS)

    The TCD fusion fuel cycle - where the 3He from the D(D,n)3He reaction is transmuted, by neutron capture in the blanket, into tritium which is fed back to the plasma - was recently recognized as being potentially more promising than the Catalyzed Deuterium (Cat-D) fuel cycle for tokamak power reactors. It is the purpose of the present work to assess the feasibility of, and to identify promising directions for designing blankets for TCD fusion reactors

  20. Impact of prescribed burning on blanket peat hydrology

    OpenAIRE

    Holden, J; Palmer, SM; Johnston, K; Wearing, C.; Irvine, B; Brown, LE

    2015-01-01

    Fire is known to impact soil properties and hydrological flowpaths. However, the impact of prescribed vegetation burning on blanket peatland hydrology is poorly understood. We studied ten blanket peat headwater catchments. Five were subject to prescribed burning, while five were unburnt controls. Within the burnt catchments we studied plots where the last burn occurred ∼2 (B2), 4 (B4), 7 (B7) or greater than 10 years (B10+) prior to the start of measurements. These were compared with plots at...

  1. Analysis on tritium controlling of the dual-cooled lithium lead blanket for fusion power reactor FDS-II

    International Nuclear Information System (INIS)

    A tritium flow model of the entire FDS-II blanket system was developed and the preliminary analysis on tritium permeation and extraction for FDS-II blanket system were done by using Tritium Analysis Software (TAS). The factors which affected tritium extraction and permeation were calculated and evaluated, such as tritium permeation reduction factor in blanket, proportion of LiPb flow in tritium extraction system and helium leakage rate, etc. The results of the presented analysis shows that further R and D efforts are still required to guarantee the tritium self-sufficient and safety, for example high quality tritium permeation barriers, efficiency of tritium extraction from LiPb and fabrication technology of the LiPb heat exchanger, etc.

  2. Analysis on tritium controlling of the dual-cooled lithium lead blanket for fusion power reactor FDS-II

    Energy Technology Data Exchange (ETDEWEB)

    Song Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)], E-mail: ysong@ipp.ac.cn; Huang Qunying [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Wang Yongliang [College of Physical Science and Technology, Sichuan University, Chengdu, Sichuan 610064 (China); Ni Muyi [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2009-06-15

    A tritium flow model of the entire FDS-II blanket system was developed and the preliminary analysis on tritium permeation and extraction for FDS-II blanket system were done by using Tritium Analysis Software (TAS). The factors which affected tritium extraction and permeation were calculated and evaluated, such as tritium permeation reduction factor in blanket, proportion of LiPb flow in tritium extraction system and helium leakage rate, etc. The results of the presented analysis shows that further R and D efforts are still required to guarantee the tritium self-sufficient and safety, for example high quality tritium permeation barriers, efficiency of tritium extraction from LiPb and fabrication technology of the LiPb heat exchanger, etc.

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

    International Nuclear Information System (INIS)

    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.)

  4. Modeling and Analysis of Tritium Transport in Multi-Region Lead-Lithium Liquid Metal Blankets

    OpenAIRE

    Zhang, Hongjie

    2014-01-01

    It is critical to be able to predict tritium transport in lead-lithium liquid metal (LM) blankets with great accuracy to provide information for fusion reactor safety and economy analyses. However, tritium transport processes are complex and affected by multiple physics such as magnetohydrodynamic (MHD) flow, yet there is no single computer code capable of simulating these phenomena inclusively. Thus the objectives of this research are: 1) to develop mathematical models and computational code...

  5. Biophysical Assessment and Predicted Thermophysiologic Effects of Body Armor.

    Directory of Open Access Journals (Sweden)

    Adam W Potter

    Full Text Available Military personnel are often required to wear ballistic protection in order to defend against enemies. However, this added protection increases mass carried and imposes additional thermal burden on the individual. Body armor (BA is known to reduce combat casualties, but the effects of BA mass and insulation on the physical performance of soldiers are less well documented. Until recently, the emphasis has been increasing personal protection, with little consideration of the adverse impacts on human performance.The purpose of this work was to use sweating thermal manikin and mathematical modeling techniques to quantify the tradeoff between increased BA protection, the accompanying mass, and thermal effects on human performance.Using a sweating thermal manikin, total insulation (IT, clo and vapor permeability indexes (im were measured for a baseline clothing ensemble with and without one of seven increasingly protective U.S. Army BA configurations. Using mathematical modeling, predictions were made of thermal impact on humans wearing each configuration while working in hot/dry (desert, hot/humid (jungle, and temperate environmental conditions.In nearly still air (0.4 m/s, IT ranged from 1.57 to 1.63 clo and im from 0.35 to 0.42 for the seven BA conditions, compared to IT and im values of 1.37 clo and 0.45 respectively, for the baseline condition (no BA.Biophysical assessments and predictive modeling show a quantifiable relationship exists among increased protection and increased thermal burden and decreased work capacity. This approach enables quantitative analysis of the tradeoffs between ballistic protection, thermal-work strain, and physical work performance.

  6. The integrated-blanket-coil concept applied to the poloidal field and blanket systems of a tokamak reactor

    International Nuclear Information System (INIS)

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component. This concept, designated the ''integrated-blanket-coil'' (IBC) concept, is applied to the poloidal field and blanket systems of a tokamak reactor. An examination of resistive power losses in the IBC suggests that these losses can be limited to 10% of the fusion thermal power. By assuming a sandwich construction for the IBC walls, magnetohydrodynamic (MHD)-induced pressure drops and associated pressure stresses are shown to be modest and well below design limits. For the stainless steel reference case examined, the MHD-induced pressure drop was estimated to be about 1/3 MPa and the associated primary membrane stress was estimated to be about 47 MPa. The preliminary analyses indicate that the IBC concept offers promise as a means for making fusion reactors more compact by combining blanket and coil functions in a single component

  7. Safety Evaluation of the EVOLVE Blanket Concept

    International Nuclear Information System (INIS)

    This article summarizes the results of the safety evaluation of the Evaporation of Lithium and Vapor Extraction (EVOLVE) W-alloy first wall (FW) and blanket concept. We have analyzed the EVOLVE design response during a confinement bypass accident. A confinement bypass accident was chosen because, based on previous safety studies, this accident can produce environmental releases by breaching the primary radioactive confinement boundary of EVOLVE, which is the EVOLVE vacuum vessel (VV). As a consequence of a bypass accident, air from a room adjoining the reactor enters the plasma chamber by way of a failed VV port. This air reacts with the high temperature metals inside of the VV to release energy in the case of a lithium spill, or to mobilize radioactive material by oxidation, and then transport this material to the environment by natural convection airflow through the failed VV port. We use the MELCOR code to analyze the response of EVOLVE during this accident. Based on these results, the EVOLVE concept can meet the no-evacuation dose goal set by the DOE Fusion Safety Standard if the EVOLVE confinement building ventilation system is closed within two hours of the onset of this accident

  8. Current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phase of the Li-LiH, Li-LiD, and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li3N, Li2O, and Li2C2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g., Li--Al and Li--Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li--M alloys can be estimated from lithium activity data for these alloys

  9. Flow characteristics of the Cascade granular blanket

    International Nuclear Information System (INIS)

    Analysis of a single granule on a rotating cone shows that for the 350 half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer

  10. APT Blanket System Loss-of-Helium-Gas Accident Based on Initial Conceptual Design - Helium Supply Rupture into Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The model results are used to determine if beam power shutdown is necessary (or not) as a result of the LOHGA accident to maintain the blanket system well below any of the thermal-hydraulic constraints imposed on the design. The results also provide boundary conditions to the detailed bin model to study the detailed temperature response of the hot blanket module structure. The results for these two cases are documented in the report.

  11. 基于技术检测的装甲装备武器系统执行任务能力评估%Ability to Implement the Mandate Assessment of Armored Equipment Weapons System Based on Technical Detection

    Institute of Scientific and Technical Information of China (English)

    张辽宁; 屈洋; 张政; 邢旺

    2016-01-01

    Technology state affects the ability to implement the mandate of armored equipment weapons system greatly, yet there are few studies on its evaluation. Aimed at the problem, the paper put forward a method of evaluating the ability to implement the mandate of armored equipment based on the technical detection of equipment, and a case evaluation and analysis of a certain armored equipment weapon system is carried out. The method is operable and can provide data reference for the development of technical detection equipment and technology of armored equipment weapons system.%技术状态对装甲装备武器系统的执行任务能力有着重要影响,目前对其评估的研究还很少。针对这一问题,提出一种利用技术检测对装甲装备武器系统执行任务能力进行评估的方法,并结合某装甲装备武器系统进行了实例分析。分析结果表明:该方法具有可操作性,能够为装甲装备武器系统检测设备和技术的发展提供数据参考。

  12. Design requirements for SiC/SiC composites structural material in fusion power reactor blankets

    International Nuclear Information System (INIS)

    This paper recalls the main features of the TAURO blanket, a self-cooled Pb-17Li concept using SiC/SiC composites as structural material, developed for FPR. The objective of this design activity is to compare the characteristics of present-day industrial SiC-SiC composites with those required for a fusion power reactor blanket (FPR) and to evaluate the main needs of further R and D. The performed analyses indicated that the TAURO blanket would need the availability of SiC/SiC composites approximately 10 mm thick with a thermal conductivity through the thickness of approximately 15 Wm-1K-1 at 1000 C and a low electrical conductivity. A preliminary MHD analysis has indicated that the electrical conductivity should not be greater than 500 Ω-1m-1. Irradiation effects should be included in these figures. Under these conditions, the calculated pressure drop due to the high Pb-17Li velocity (approximately 1 m s-1) is much lower then 0.1 MPa. The characteristics and data base of the recently developed 3D-SiC/SiC composite, Cerasep trademark N3-1, are reported and discussed in relation to the identified blanket design requirements. The progress on joining techniques is briefly reported. For the time being, the best results have been obtained using Si-based brazing systems initially developed for SiC ceramics and whose major issue is the higher porosity of the SiC/SiC composites. (orig.)

  13. Shock tube design for high intensity blast waves for laboratory testing of armor and combat materiel

    Directory of Open Access Journals (Sweden)

    Elijah Courtney

    2014-06-01

    Full Text Available Shock tubes create simulated blast waves which can be directed and measured to study blast wave effects under laboratory conditions. It is desirable to increase available peak pressure from ∼1 MPa to ∼5 MPa to simulate closer blast sources and facilitate development and testing of personal and vehicle armors. Three methods are experimentally investigated to increase peak simulated blast pressure produced by an oxy-acetylene driven shock tube while maintaining suitability for laboratory studies. The first method is the addition of a Shchelkin spiral priming section which supports a deflagration to detonation transition. This approach increases the average peak pressure from 1.17 MPa to 5.33 MPa while maintaining a relevant pressure-time curve (near Friedlander waveform. The second method is a bottleneck between the driving and driven sections. Coupling a 79 mm diameter driving section to a 53 mm driven section increases the peak pressure from 1.17 MPa to 2.25 MPa. A 103 mm driving section is used to increase peak pressure to 2.64 MPa. The third method, adding solid fuel to the driving section with the oxy-acetylene, results in a peak pressure increasing to 1.70 MPa.

  14. Shock Tube Design for High Intensity Blast Waves for Laboratory Testing of Armor and Combat Materiel

    CERN Document Server

    Courtney, Elijah; Courtney, Michael

    2015-01-01

    Shock tubes create simulated blast waves which can be directed and measured to study blast wave effects under laboratory conditions. It is desirable to increase available peak pressure from ~1 MPa to ~5 MPa to simulate closer blast sources and facilitate development and testing of personal and vehicle armors. Three methods were investigated to increase peak simulated blast pressure produced by an oxy-acetylene driven shock tube while maintaining suitability for laboratory studies. The first method is the addition of a Shchelkin spiral priming section which works by increasing the turbulent flow of the deflagration wave, thus increasing its speed and pressure. This approach increased the average peak pressure from 1.17 MPa to 5.33 MPa while maintaining a relevant pressure-time curve (Friedlander waveform). The second method is a bottleneck between the driving and driven sections. Coupling a 79 mm diameter driving section to a 53 mm driven section increased the peak pressure from 1.17 MPa to 2.25 MPa. Using a 1...

  15. Shock tube design for high intensity blast waves for laboratory testing of armor and combat materiel

    Institute of Scientific and Technical Information of China (English)

    Elijah COURTNEY; Amy COURTNEY; Michael COURTNEY

    2014-01-01

    Shock tubes create simulated blast waves which can be directed and measured to study blast wave effects under laboratory conditions. It is desirable to increase available peak pressure from w1 MPa to w5 MPa to simulate closer blast sources and facilitate development and testing of personal and vehicle armors. Three methods are experimentally investigated to increase peak simulated blast pressure produced by an oxy-acetylene driven shock tube while maintaining suitability for laboratory studies. The first method is the addition of a Shchelkin spiral prim-ing section which supports a deflagration to detonation transition. This approach increases the average peak pressure from 1.17 MPa to 5.33 MPa while maintaining a relevant pressure-time curve (near Friedlander waveform). The second method is a bottleneck between the driving and driven sections. Coupling a 79 mm diameter driving section to a 53 mm driven section increases the peak pressure from 1.17 MPa to 2.25 MPa. A 103 mm driving section is used to increase peak pressure to 2.64 MPa. The third method, adding solid fuel to the driving section with the oxy-acetylene, results in a peak pressure increasing to 1.70 MPa.

  16. Combat body armor and injuries to the head, face, and neck region: a systematic review.

    Science.gov (United States)

    Tong, Darryl; Beirne, Ross

    2013-04-01

    There has been a reported increase in combat-related head, face, and neck (HFN) injuries among service personnel wearing combat body armor (CBA) that have deployed to Iraq and Afghanistan. Modern ceramic plate CBA has decreased the incidence of fatal-penetrating injuries to the torso but offers no protection to the limbs and face which remain exposed to gunshot wounds and fragments from explosive devices. The aim of this review was to systematically summarize the literature reporting on HFN injuries sustained by combat personnel wearing CBA and to highlight recommendations for increased protection to the facial region. Three major contributing factors were identified with this proportional increase in HFN injuries, namely the increased survivability of soldiers because of CBA, fragments injuries from explosive devices, and the lack of protection to the face and limbs. There appears to be no evidence to suggest that by virtue of wearing CBA the likelihood of sustaining an HFN injury increases as such, but a higher incidence of fragment injuries to the HFN region may be due to the more common use of improvised explosive devicess and other explosive devices. Further development of lightweight protection for the face is needed.

  17. Combat body armor and injuries to the head, face, and neck region: a systematic review.

    Science.gov (United States)

    Tong, Darryl; Beirne, Ross

    2013-04-01

    There has been a reported increase in combat-related head, face, and neck (HFN) injuries among service personnel wearing combat body armor (CBA) that have deployed to Iraq and Afghanistan. Modern ceramic plate CBA has decreased the incidence of fatal-penetrating injuries to the torso but offers no protection to the limbs and face which remain exposed to gunshot wounds and fragments from explosive devices. The aim of this review was to systematically summarize the literature reporting on HFN injuries sustained by combat personnel wearing CBA and to highlight recommendations for increased protection to the facial region. Three major contributing factors were identified with this proportional increase in HFN injuries, namely the increased survivability of soldiers because of CBA, fragments injuries from explosive devices, and the lack of protection to the face and limbs. There appears to be no evidence to suggest that by virtue of wearing CBA the likelihood of sustaining an HFN injury increases as such, but a higher incidence of fragment injuries to the HFN region may be due to the more common use of improvised explosive devicess and other explosive devices. Further development of lightweight protection for the face is needed. PMID:23707828

  18. Flow and Turbulence at Rubble-Mound Breakwater Armor Layers under Solitary Wave

    DEFF Research Database (Denmark)

    Jensen, Bjarne; Christensen, Erik Damgaard; Sumer, B. Mutlu;

    2015-01-01

    to the breakwater, and (3) a porous breakwater where the porous core was added below the breakwater front. One breakwater slope of 1:1.5 was applied. In this paper the focus is on the details of a single sequence of wave approach, run-up, and rundown. To isolate this sequence the experiments were performed applying......This paper presents the results of an experimental investigation of the flow and turbulence at the armor layer of rubble-mound breakwaters during wave action. The study focused on the details of the flow and turbulence in the armor layer and on the effect of the porous core on flow and stability...... a solitary wave. The individual sources of turbulence generation were distinguished using Laser Doppler anemometry measurements, and the effect of the armor layer and porous core was described in terms of a reduced impact of the rundown process, production of lee wake turbulence, and less transport...

  19. Effect of Multi-Shot X-Ray Exposures in IFE Armor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Latkowski, J F; Abbott, R P; Schmitt, R C; Bell, B K

    2004-12-10

    As part of the High Average Power Laser (HAPL) program the performance of tungsten as an armor material is being studied. While the armor would be exposed to neutrons, x-rays and ions within an inertial fusion energy (IFE) power plant, the thermomechanical effects are believed to dominate. Using a pulsed x-ray source, long-term exposures of tungsten have been completed at fluences that are of interest for the IFE application. Modeling is used in conjunction with experiments on the XAPPER x-ray damage facility in an effort to recreate the effects that would be expected in an operating IFE power plant. X-ray exposures have been completed for a variety of x-ray fluences and number of shots. Analysis of the samples suggests that surface roughening has a threshold that is very close to the fluences that reproduce the peak temperatures expected in an IFE armor material.

  20. Efficient COD removal and nitrification in an upflow microaerobic sludge blanket reactor for domestic wastewater.

    Science.gov (United States)

    Zheng, Shaokui; Cui, Cancan

    2012-03-01

    The treatment performance of an upflow microaerobic sludge blanket reactor (UMSB) for synthetic domestic wastewater was investigated at two dissolved oxygen (DO) levels, 0.3-0.5 and 0.7-0.9 mg l(-1), focusing on nitrification performance. The higher DO level induced complete nitrification of ammonia nitrogen (NH(3)-N), achieving chemical oxygen demand and NH(3)-N removals of 97 and 92%, respectively. There were consistently significantly higher nitrate nitrogen (NO(3)-N) and nitrite nitrogen (NO(2)-N) levels in the effluent, with ~66% of newly-produced oxidised nitrogen as NO(2)-N. Despite the high nitrification efficiency, only about 23% of the removed NH(3)-N amount from the influent was ultimately transformed into oxidised nitrogen due to the simultaneous nitrification-denitrification. Sludge blanket development and granulation occurred simultaneously in the UMSB. PMID:22105554

  1. Four loss-of-flow accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    This report presents the thermal-hydraulic analysis of four Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the alternative SEAFP reactor design. The LOFAs considered result from a loss of electrical power for the recirculation pump in the primary cooling circuit. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analyses, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall and blanket. For the LOFA without plasma shutdown, significant loss of heat removal due to dryout occurs at the midplane of the outboard first wall cooling pipes about 41 s after pump trip. For the three LOFA cases with emergency plasma shutdown that have been studied, the temperature increase in the Be-coating at the midplane of the outboard first wall is limited to about 30 K. (orig.)

  2. Considerations on techniques for improving tritium confinement in helium-cooled ceramic breeder blankets

    International Nuclear Information System (INIS)

    Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic BIT blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. Our results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (orig.)

  3. Manufacturing process scale-up of optical grade transparent spinel ceramic at ArmorLine Corporation

    Science.gov (United States)

    Spilman, Joseph; Voyles, John; Nick, Joseph; Shaffer, Lawrence

    2013-06-01

    While transparent Spinel ceramic's mechanical and optical characteristics are ideal for many Ultraviolet (UV), visible, Short-Wave Infrared (SWIR), Mid-Wave Infrared (MWIR), and multispectral sensor window applications, commercial adoption of the material has been hampered because the material has historically been available in relatively small sizes (one square foot per window or less), low volumes, unreliable supply, and with unreliable quality. Recent efforts, most notably by Technology Assessment and Transfer (TA and T), have scaled-up manufacturing processes and demonstrated the capability to produce larger windows on the order of two square feet, but with limited output not suitable for production type programs. ArmorLine Corporation licensed the hot-pressed Spinel manufacturing know-how of TA and T in 2009 with the goal of building the world's first dedicated full-scale Spinel production facility, enabling the supply of a reliable and sufficient volume of large Transparent Armor and Optical Grade Spinel plates. With over $20 million of private investment by J.F. Lehman and Company, ArmorLine has installed and commissioned the largest vacuum hot press in the world, the largest high-temperature/high-pressure hot isostatic press in the world, and supporting manufacturing processes within 75,000 square feet of manufacturing space. ArmorLine's equipment is capable of producing window blanks as large as 50" x 30" and the facility is capable of producing substantial volumes of material with its Lean configuration and 24/7 operation. Initial production capability was achieved in 2012. ArmorLine will discuss the challenges that were encountered during scale-up of the manufacturing processes, ArmorLine Optical Grade Spinel optical performance, and provide an overview of the facility and its capabilities.

  4. Ballistic Resistance of Armored Passenger Vehicles: Test Protocols and Quality Methods

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey M. Lacy; Robert E. Polk

    2005-07-01

    This guide establishes a test methodology for determining the overall ballistic resistance of the passenger compartment of assembled nontactical armored passenger vehicles (APVs). Because ballistic testing of every piece of every component of an armored vehicle is impractical, if not impossible, this guide describes a testing scheme based on statistical sampling of exposed component surface areas. Results from the test of the sampled points are combined to form a test score that reflects the probability of ballistic penetration into the passenger compartment of the vehicle.

  5. Ceramic helium-cooled blanket test module

    Energy Technology Data Exchange (ETDEWEB)

    Leshukov, A. E-mail: leshu@entek.ru; Kovalenko, V.; Shatalov, G.; Goroshkin, G.; Obukhov, A

    2000-11-01

    The design of RF DEMO-relevant ceramic helium cooled blanket test module (CHC BTM) for testing in international thermonuclear experimental reactor (ITER) is under consideration. The RF concept of DEMO BTM is based upon the breeder inside tube (BIT)-concept. This concept suggests the use of solid breeding ceramic material, helium as coolant and tritium purge-gas, ferrite-martensite steel as structural material, and beryllium as neutron multiplier. The parameters of the primary circuit coolant are the following, pressure -8 MPa, inlet/outlet temperature -300/550 deg. C, respectively. Helium (0.1 MPa pressure) is used for tritium removal from ceramic breeder. The ITER water coolant is the secondary circuit coolant of DEMO BTM cooling system. Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is used as tritium breeding material (pebbles-bed of diameter 0.5-1 mm spheres). It is planned to use the beryllium as neutron multiplier (spheres diameter 1 mm pebbles-bed or the porous beryllium). The 3-D neutronic calculations on Monte Carlo method, in accordance with FENDL-1 library of the nuclear data, have been performed for CHC BTM. To validate the CHC BTM concept, the thermal hydraulic analysis has been performed for the design elements and cooling system equipment. The preliminary stress analysis for BTM design elements has been carried out on the ASME-code and RF strength regulations. The four types of LOFA and LOCA accidents have been investigated. The parameters of cooling, coolant purification and tritium extraction systems have been determined.

  6. Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

    Energy Technology Data Exchange (ETDEWEB)

    Malara, C. [Institute Regional des Materiaux Avances, Ispra (Italy); Casini, G. [Systems Engineering and Informatics Institute, JRC Ispra, Ispra (Vatican City State, Holy See) (Italy); Viola, A. [Department of Chemical Engineering, University of Cagliari, Cagliari (Italy)

    1995-03-01

    The question of the tritium recovery in water-cooled Pb17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pb17Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit. (orig.).

  7. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  8. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji [ed.

    1998-03-01

    This report is the Proceedings of `the Sixth International Workshop on Ceramic Breeder Blanket Interactions` which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: (1) fabrication and characterization of ceramic breeders, (2) properties data for ceramic breeders, (3) tritium release characteristics, (4) modeling of tritium behavior, (5) irradiation effects on performance behavior, (6) blanket design and R and D requirements, (7) hydrogen behavior in materials, and (8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li{sub 2}TiO{sub 3}, tritium release behavior of Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} including tritium diffusion, modeling of tritium release from Li{sub 2}ZrO{sub 3} in ITER condition, helium release behavior from Li{sub 2}O, results of tritium release irradiation tests of Li{sub 4}SiO{sub 4} pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  9. Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

    Science.gov (United States)

    Li, Jia; Zhang, Xiaokang; Gao, Fangfang; Pu, Yong

    2016-02-01

    China Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2 to ensure tritium self-sufficiency. A concept design for a water cooled ceramics breeding blanket (WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR. Based on this concept, a one-dimensional (1D) radial built breeding blanket was first designed, and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build. A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models, addressing neutron wall loading (NWL), tritium breeding ratio (TBR), fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components. The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  10. [Air conditioning units and warm air blankets in the operating room].

    Science.gov (United States)

    Kerwat, Klaus; Piechowiak, Karolin; Wulf, Hinnerk

    2013-01-01

    Nowadays almost all operating rooms are equipped with air conditioning (AC units). Their main purpose is climatization, like ventilation, moisturizing, cooling and also the warming of the room in large buildings. In operating rooms they have an additional function in the prevention of infections, especially the avoidance of postoperative wound infections. This is achieved by special filtration systems and by the creation of specific air currents. Since hypothermia is known to be an unambiguous factor for the development of postoperative wound infections, patients are often actively warmed intraoperatively using warm air blankets (forced-air warming units). In such cases it is frequently discussed whether such warm air blankets affect the performance of AC units by changing the air currents or whether, in contrast, have exactly the opposite effect. However, it has been demonstrated in numerous studies that warm air blankets do not have any relevant effect on the functioning of AC units. Also there are no indications that their use increases the rate of postoperative wound infections. By preventing the patient from experiencing hypothermia, the rate of postoperative wound infections can even be decreased thereby.

  11. Impact of blanket tritium against the tritium plant of fusion reactor

    International Nuclear Information System (INIS)

    The breeder blanket and the blanket tritium recovery system are tested using test blanket modules during ITER campaign. And then, these are integrated with the tritium plant for the first time at a prototype reactor after ITER. In this work, impact to the tritium plant by integration of the solid breeder blanket was discussed. The method of tritium extraction from the blanket and the choice of the process for breeder blanket interface should be discussed not only from the viewpoint of tritium release but also from the viewpoint of the load of processing. (author)

  12. A Residual Mass Ballistic Testing Method to Compare Armor Materials or Components (Residual Mass Ballistic Testing Method)

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin Langhorst; Thomas M Lillo; Henry S Chu

    2014-05-01

    A statistics based ballistic test method is presented for use when comparing multiple groups of test articles of unknown relative ballistic perforation resistance. The method is intended to be more efficient than many traditional methods for research and development testing. To establish the validity of the method, it is employed in this study to compare test groups of known relative ballistic performance. Multiple groups of test articles were perforated using consistent projectiles and impact conditions. Test groups were made of rolled homogeneous armor (RHA) plates and differed in thickness. After perforation, each residual projectile was captured behind the target and its mass was measured. The residual masses measured for each test group were analyzed to provide ballistic performance rankings with associated confidence levels. When compared to traditional V50 methods, the residual mass (RM) method was found to require fewer test events and be more tolerant of variations in impact conditions.

  13. Surface property variations in Venusian fluidized ejecta blanket craters

    Science.gov (United States)

    Johnson, Jeffrey R.; Baker, Victor R.

    1994-01-01

    A comprehensive study of Magellan Cycles 1 and 2 radar data from Venus reveals surface roughness and dielectric variations associated with fluidized ejecta blanket (FEB) craters that help illuminate styles of flow ejecta emplacement. This study develops new procedures of digital unit mapping and polygon-filling algorithms using Magellan synthetic aperture radar (SAR), altimetry, and radiometry data. These techniques allow the extraction of radiophysical information for FEB crater materials, nearby plains, and lava flows. Backscatter curve slopes of the FEBs studied here are consistent with surface textures that are transitional between a'a and pahoehoe-like. Average surface property values of ejecta units are relatively similar for a given crater, but are discernibly different from other craters. Individual crater ejecta reflectivity and emissivity values are relatively similar to those for the surrounding plains, which may suggest a link between plains material and ejecta dielectric properties. Increasing FEB roughness downflow are interpreted to be associated with more lava-like flows, while decreasing roughness are more similar to trends typical of gravity (pyroclastic-like or debris-like) flows. Most commonly, FEB crater flow materials exhibit transitions from proximal, lava/melt-like flow styles to distal, gravity flow-like styles. Some FEBs show more complicated behavior, however, or appear to be more dominated by dielectric differences downflow, as inferred from correlations between the data sets. Such transitions may result from changes in local topography or from overlapping of flow lobes during FEB emplacement.

  14. 27 CFR 478.148 - Armor piercing ammunition intended for sporting or industrial purposes.

    Science.gov (United States)

    2010-04-01

    ... intended for sporting or industrial purposes. 478.148 Section 478.148 Alcohol, Tobacco Products, and... ammunition intended for sporting or industrial purposes. The Director may exempt certain armor piercing... for any such ammunition which is primarily intended for sporting purposes or intended for...

  15. 7 CFR 1755.406 - Shield or armor ground resistance measurements.

    Science.gov (United States)

    2010-01-01

    ...) The insulation resistance test set should have an output voltage not to exceed 500 volts dc and may be hand cranked or battery operated. (3) The dc bridge type megohmmeter, which may be ac powered, should... to 10 megohms. The voltage that is applied to the shield or armor during the test should not be...

  16. Desenvolvimento e operação de reator anaeróbio de manta de lodo (UASB no tratamento dos efluentes da suinocultura em escala laboratorial Development and operation of an upflow anaerobic sludge blanket reactor (UASB treating liquid effluent from swine manure in laboratory scale

    Directory of Open Access Journals (Sweden)

    Cláudio Milton Montenegro Campos

    2006-02-01

    Full Text Available A atividade suinícola vem, desde meados da década de 70, sendo uma das mais poluidoras atividades agroindustriais no Estado de Minas Gerais. Sendo assim, objetivou-se desenvolver um Reator Anaeróbio de Manta de Lodo (UASB-Upflow Anaerobic Sludge Blanket visando tratar os dejetos produzidos com máxima eficiência dentro de um tempo e com custo reduzidos. Para tanto um experimento em escala laboratorial foi projetado e monitorado no Laboratório de Análise de Água do Departamento de Engenharia da Universidade Federal de Lavras (LAADEG, sendo composto por um Tanque de Acidificação e Equalização (TAE, um Reator Anaeróbio de Manta de Lodo e uma Lagoa Aerada Facultativa (LAF, o qual foi alimentado com fluxo contínuo. As análises físico-químicas realizadas foram: DQO, DBO5, Sólidos Totais (Fixos e Voláteis, Temperatura, pH, Nitrogênio, Fósforo, Acidez e Alcalinidade Total. O sistema proporcionou eficiência de remoção média de 93% de DQO, 84% de DBO5 e 85% de Sólidos Totais Voláteis, demonstrando adequada adaptação aos diversos tempos de detenção hidráulica adotados (55, 40, 30, 25, 18 e 15 horas. Os parâmetros adotados na partida do reator UASB foram: COV: 1,11kgDQO.m-3.d-1, COB: 0,019 kgDBO5.kgSVT-1.d-1 e TDH: 55h.The swine production, since 70th , is one of the most pollutant agro-industrial activities in the Minas Gerais State, Brazil. The objective of this research was to develop an Upflow Anaerobic Sludge Blanket Reactor (UASB, aiming at treating the effluent generated within a maximum efficiency and minimum time and cost. Therefore, a lab-scale reactor was built up and monitored in the laboratory of Engineering Department at the Federal University of Lavras (UFLA. The system consisted of an Acidification and Equalization Tank (AET, an Upflow Anaerobic Sludge Blanket reactor (UASB, and an Aerated Facultative Pond (AFP. The system was fed continuously. The physical-chemical analyses carried out were: COD, BOD5, Total

  17. External Quality Assessment for the Detection of Measles Virus by Reverse Transcription-PCR Using Armored RNA.

    Directory of Open Access Journals (Sweden)

    Dong Zhang

    Full Text Available In recent years, nucleic acid tests for detection of measles virus RNA have been widely applied in laboratories belonging to the measles surveillance system of China. An external quality assessment program was established by the National Center for Clinical Laboratories to evaluate the performance of nucleic acid tests for measles virus. The external quality assessment panel, which consisted of 10 specimens, was prepared using armored RNAs, complex of noninfectious MS2 bacteriophage coat proteins encapsulated RNA of measles virus, as measles virus surrogate controls. Conserved sequences amplified from a circulating measles virus strain or from a vaccine strain were encapsulated into these armored RNAs. Forty-one participating laboratories from 15 provinces, municipalities, or autonomous regions that currently conduct molecular detection of measles virus enrolled in the external quality assessment program, including 40 measles surveillance system laboratories and one diagnostic reagent manufacturer. Forty laboratories used commercial reverse transcription-quantitative PCR kits, with only one laboratory applying a conventional PCR method developed in-house. The results indicated that most of the participants (38/41, 92.7% were able to accurately detect the panel with 100% sensitivity and 100% specificity. Although a wide range of commercially available kits for nucleic acid extraction and reverse transcription polymerase chain reaction were used by the participants, only two false-negative results and one false-positive result were generated; these were generated by three separate laboratories. Both false-negative results were obtained with tests performed on specimens with the lowest concentration (1.2 × 104 genomic equivalents/mL. In addition, all 18 participants from Beijing achieved 100% sensitivity and 100% specificity. Overall, we conclude that the majority of the laboratories evaluated have reliable diagnostic capacities for the detection

  18. First wall and blanket concepts for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-07-01

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

  19. Modeling Space-Time Dependent Helium Bubble Evolution in Tungsten Armor under IFE Conditions

    International Nuclear Information System (INIS)

    The High Average Power Laser (HAPL) program is a coordinated effort to develop Laser Inertial Fusion Energy. The implosion of the D-T target produces a spectrum of neutrons, X-rays, and charged particles, which arrive at the first wall (FW) at different times within about 2.5 μs at a frequency of 5 to 10 Hz. Helium is one of several high-energy charged particle constituents impinging on the candidate tungsten armored low activation ferritic steel First Wall. The spread of the implanted debris and burn helium energies results in a unique space-time dependent implantation profile that spans about 10 μm in tungsten. Co-implantation of X-rays and other ions results in spatially dependent damage profiles and rapid space-time dependent temperature spikes and gradients. The rate of helium transport and helium bubble formation will vary significantly throughout the implanted region. Furthermore, helium will also be transported via the migration of helium bubbles and non-equilibrium helium-vacancy clusters. The HEROS code was developed at UCLA to model the spatial and time-dependent helium bubble nucleation, growth, coalescence, and migration under transient damage rates and transient temperature gradients. The HEROS code is based on kinetic rate theory, which includes clustering of helium and vacancies, helium mobility, helium-vacancy cluster stability, cavity nucleation and growth and other microstructural features such as interstitial loop evolution, grain boundaries, and precipitates. The HEROS code is based on space-time discretization of reaction-diffusion type equations to account for migration of mobile species between neighboring bins as single atoms, clusters, or bubbles. HAPL chamber FW implantation conditions are used to model helium bubble evolution in the implanted tungsten. Helium recycling rate predictions are compared with experimental results of helium ion implantation experiments. (author)

  20. Analysis on tritium management in FLiBe blanket for LHD-type helical reactor FFHR2

    International Nuclear Information System (INIS)

    In FFHR2 (LHD-type helical reactor) design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extracting tritium from breeder and controlling the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The results of the analysis showed that it was reasonable to select W alloy as heat exchanger (HX) material, the proportion of FLiBe flow in tritium recover system (TRS) was 0.2, the efficiency of TRS was 0.85 and tritium permeation reduction factor (TPRF) was 20 in blanket etc.. In addition, further R and D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. (author)

  1. Analysis on tritium management in FLiBe blanket for force-free helical reactor FFHR2

    International Nuclear Information System (INIS)

    In FFHR2 design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extract tritium from breeder and control the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The factors which affected tritium extraction and permeation were calculated and evaluated, such as the heat exchanger material, tritium permeation reduction factor (TPRF) in blanket, proportion of FLiBe flow in tritium recover system (TRS) and efficiency of TRS etc. The results of the analysis showed that further R and D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. (author)

  2. Tritium self-sufficiency of HCPB blanket modules for DEMO considering time-varying neutron flux spectra and material compositions

    Energy Technology Data Exchange (ETDEWEB)

    Aures, A., E-mail: Alexander.Aures@ccfe.ac.uk; Packer, L.W.; Zheng, S.

    2013-10-15

    Highlights: • Simulations on the tritium breeding performance of HCPB blanket modules were done. • MCNP5 and FISPACT were used for coupled transport and activation calculations. • Material transmutation affects the neutron flux spectra within the blanket modules. • The consequences of time-dependent spectra on TBR and tritium self-sufficiency were investigated. -- Abstract: Significant transmutation of solid-type breeding blanket materials affects the time and spatial variation of neutron energy within such materials. This has an impact on simulation assumptions required to accurately assess tritium surplus quantities for conceptual power plant devices. This paper details an investigation, via simulation, of the consequences for the tritium breeding ratio and the tritium self-sufficiency of a DEMO concept with homogeneous Helium-Cooled Pebble Bed blanket modules containing Li{sub 4}SiO{sub 4} ceramic breeder material. For this purpose, a code was developed to couple MCNP5 and FISPACT to supply material compositions from activation calculations to the neutron transport calculation in an iterative loop covering several time steps. Simulation results are presented for a simple 1D spherical device model and a DEMO tokamak model.

  3. Maximizing fluence rate and field uniformity of light blanket for intraoperative PDT

    OpenAIRE

    LIANG, XING; Kundu, Palak; Finlay, Jarod; Goodwin, Michael; Zhu, Timothy C.

    2012-01-01

    A light blanket is designed with a system of cylindrically diffusing optical fibers, which are spirally oriented. This 25×30 cm rectangular light blanket is capable of providing uniform illumination during intraoperative photodynamic therapy. The flexibility of the blanket proves to be extremely beneficial when conforming to the anatomical structures of the patient being treated. Previous tests of light distribution from the blanket have shown significant loss of intensity with the length of ...

  4. APT Blanket Safety Analysis: Counter Current Flow Limitation for Cavity Spaces

    International Nuclear Information System (INIS)

    The thermal-hydraulic modeling aspects for the APT blanket system have been broken up into two basic modeling components: (1) the blanket system and (2) the cavity flood system. In most cases these systems are modeled separately. This separate study for the coolability of the blanket modules can also be used to establish/evaluate a functional design requirement on gap size between the blanket modules

  5. 75 FR 60095 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-09-29

    ... LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY..., by Sempra LNG Marketing, LLC (Sempra), requesting blanket authorization to export up to a total of... Order No. 2795, which granted Cheniere Marketing, LLC (Cheniere) blanket authorization to...

  6. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    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

  7. 75 FR 11557 - Woven Electric Blankets From China

    Science.gov (United States)

    2010-03-11

    ... permitted by section 201.8 of the Commission's rules, as amended, 67 FR 68036 (November 8, 2002). Even where... specified in II (C) of the Commission's Handbook on Electronic Filing Procedures, 67 FR 68168, 68173... COMMISSION Woven Electric Blankets From China AGENCY: United States International Trade Commission....

  8. 18 CFR 33.1 - Applicability, definitions, and blanket authorizations.

    Science.gov (United States)

    2010-04-01

    ... the outstanding voting securities; or (iii) Any security of a subsidiary company within the holding... company subsidiary in connection with such acquisition. (4) A holding company granted blanket... subsidiaries, or associate companies within the holding company system has captive customers in the...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-04

    fabrication capacity per unit of core power. Nevertheless, these high-performance cores were designed to set upper bounds on the S&B core performance by using larger height and pressure drop than those of typical SFR design. A study was subsequently undertaken to quantify the tradeoff between S&B core design variables and the core performance. This study concludes that a viable S&B core can be designed without significant deviation from SFR core design practices. For example, the S&B core with 120cm active height will be comparable in volume, HM mass and specific power with the S-PRISM core and could fit within the S-PRISM reactor vessel. 43% of this core power will be generated by the once-through thorium blanket; the required capacity for reprocessing and remote fuel fabrication per unit of electricity generated will be approximately one fifth of that for a comparable ABR. The sodium void worth of this 120cm tall S&B core is significantly less positive than that of the reference ABR and the Doppler coefficient is only slightly smaller even though the seed uses a fertile-free fuel. The seed in the high transmutation core requires inert matrix fuel (TRU-40Zr) that has been successfully irradiated by the Fuel Cycle Research & Development program. An additional sensitivity analysis was later conducted to remove the bias introduced by the discrepancy between radiation damage constraints -- 200 DPA applied for S&B cores and fast fluence of 4x1023 n(>0.1MeV)/cm2 applied for ABR core design. Although the performance characteristics of the S&B cores are sensitive to the radiation damage constraint applied, the S&B cores offer very significant performance improvements relative to the conventional ABR core design when using identical constraint.

  10. Probing the Mechanical Strength of an Armored Bubble and Its Implication to Particle-Stabilized Foams

    Science.gov (United States)

    Taccoen, Nicolas; Lequeux, François; Gunes, Deniz Z.; Baroud, Charles N.

    2016-01-01

    Bubbles are dynamic objects that grow and rise or shrink and disappear, often on the scale of seconds. This conflicts with their uses in foams where they serve to modify the properties of the material in which they are embedded. Coating the bubble surface with solid particles has been demonstrated to strongly enhance the foam stability, although the mechanisms for such stabilization remain mysterious. In this paper, we reduce the problem of foam stability to the study of the behavior of a single spherical bubble coated with a monolayer of solid particles. The behavior of this armored bubble is monitored while the ambient pressure around it is varied, in order to simulate the dissolution stress resulting from the surrounding foam. We find that above a critical stress, localized dislocations appear on the armor and lead to a global loss of the mechanical stability. Once these dislocations appear, the armor is unable to prevent the dissolution of the gas into the surrounding liquid, which translates into a continued reduction of the bubble volume, even for a fixed overpressure. The observed route to the armor failure therefore begins from localized dislocations that lead to large-scale deformations of the shell until the bubble completely dissolves. The critical value of the ambient pressure that leads to the failure depends on the bubble radius, with a scaling of Δ Pcollapse∝R-1 , but does not depend on the particle diameter. These results disagree with the generally used elastic models to describe particle-covered interfaces. Instead, the experimental measurements are accounted for by an original theoretical description that equilibrates the energy gained from the gas dissolution with the capillary energy cost of displacing the individual particles. The model recovers the short-wavelength instability, the scaling of the collapse pressure with bubble radius, and the insensitivity to particle diameter. Finally, we use this new microscopic understanding to predict

  11. Intelligent Layout Method of the Powerhouse for Tank & Armored Vehicles Based on 3-Dimensional Rectangular Packing Theory

    Institute of Scientific and Technical Information of China (English)

    WANG Yan-long; MAO Ming; LU Yi-ping; BIE Jie-min

    2005-01-01

    Probes into a new and effective method in arranging the powerhouses of tank & armored vehicles. Theory and method of 3-dimensional rectangular packing are adapted to arrange effectively almost all the systems and components in the powerhouse of the vehicle, thus the study can be regarded as an attempt for the theory's engineering applications in the field of tank & armored vehicle design. It is proved that most parts of the solutions attained are reasonable, and some of the solutions are innovative.

  12. Integrated-blanket-coil (IBC) concept applied to the poloidal field and blanket systems of a tokamak reactor

    International Nuclear Information System (INIS)

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component. This concept, designated the integrated-blanket-coil (IBC) concept, is applied to the poloidal field and blanket systems of a Tokamak reactor. An examination of resistive power losses in the IBC suggests that these losses can be limited to less than or equal to 10% of the fusion thermal power. By assuming a sandwich construction for the IBC walls, MHD-induced pressure drops and associated pressure stresses are shown to be modest and well below design limits. For the stainless steel reference case examined in this paper, the MHD-induced pressure drop was estimated to be approx. 1/3 MPa and the associated primary membrane stress was estimated to be approx. 47 MPa. The preliminary analyses presented in this paper indicate that the IBC concept offers promise as a means for making fusion reactors more compact by combining blanket and coils functions in a single component

  13. Persistence of 10-year old Exxon Valdez oil on Gulf of Alaska beaches: The importance of boulder-armoring

    Energy Technology Data Exchange (ETDEWEB)

    Irvine, G.V. [United States Geological Survey , Anchorage, AK (United States); Mann, D.H. [University of Alaska, Fairbanks, AK (United States). Institute of Arctic Biology; Short, J.W. [National Oceanic and Atmospheric Administration, Juneau, AK (United States). Auke Bay Fisheries Laboratory

    2006-09-15

    Oil stranded as a result of the 1989 Exxon Valdez spill has persisted for >10 years at study sites on Gulf of Alaska shores distant from the spill's origin. These sites were contaminated by 'oil mousse', which persists in these settings due to armoring of underlying sediments and their included oil beneath boulder. The boulder-armored beaches that we resampled in 1999 showed continued contamination by subsurface oil, despite their exposure to moderate to high wave energies. Significant declines in surface oil cover occurred at all study sites. In contrast, mousse has persisted under boulders in amounts similar to what was present in 1994 and probably in 1989. Especially striking is the general lack of weathering of this subsurface oil over the last decade. Oil at five of the six armored-beach sites 10 years after the spill is compositionally similar to 11-day old Exxon Valdez oil. Analysis of movements in the boulder-armor that covers the study beaches reveals that only minor shifts have occurred since 1994, suggesting that over the last five, and probably over the last 10 years, boulder-armors have remained largely unmoved at the study sites. These findings emphasize the importance of particular geomorphic parameters in determining stranded oil persistence. Surface armoring, combined with stranding of oil mousse, results in the unexpectedly lengthy persistence of only lightly to moderately weathered oil within otherwise high-energy wave environments. (author)

  14. Thermal-hydraulic analysis of a cylindrical blanket module using ATHENA code

    International Nuclear Information System (INIS)

    ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer) is a new computer code for thermal-hydraulic analyses of many energy systems. Multiple-loop and multiple-fluid capabilities have been emphasized during the code development. A pilot version of ATHENA has incorporated a fusion kinetic package to model the effect of first wall temperature variation on the reactor conditions. The capability has been demonstrated by analyzing the performance under various conditions of a cylindrical fusion blanket module. The results have shown the viability of using ATHENA for fusion reactor design and safety analyses

  15. Neutronics analysis on helium-cooled blanket of a fusion-driven spent fuel burner

    International Nuclear Information System (INIS)

    Neutronics design and analysis of helium-cooled spent fuel burning blanket for a fusion driven sub-critical system are performed to ensure the system be able to meet the requirements of energy production (>1 GWe), more fuel breeding, more waste transmutation and long period run with deep subcritical (Keff <0.95), tritium sustainable, reasonable power density (<100 MW · m-3), which is based on 1-D burnup calculations with home-developed code VisualBUS and the data library HENDL. (authors)

  16. Production and characterization of ceramics for armor application; Producao e caracterizacao de ceramicas para blindagem balistica

    Energy Technology Data Exchange (ETDEWEB)

    Alves, J.T.; Lopes, C.M.A. [Instituto Tecnologico de Aeronautica (ITA), Sao Jose dos Campos, SP (Brazil). Div. de Engenharia Mecanica-Aeronautica; Assis, J.M.K.; Melo, F.C.L., E-mail: cmoniz@iae.cta.b [Instituto de Aeronautica e Espaco (IAE), Sao Jose dos Campos, SP (Brazil). Div. de Materiais

    2010-07-01

    The fabrication of devices for ballistic protection as bullet proof vests and helmets and armored vehicles has been evolving over the past years along with the materials and models used for this specific application. The requirements for high efficient light-weight ballistic protection systems which not interfere in the user comfort and mobility has driven the research in this area. In this work we will present the results of characterization of two ceramics based on alumina and silicon carbide. The ceramics were produced in lab scale and the specific mass, scanning electron microscopy (SEM) microstructure, Vickers hardness, flexural resistance at room temperature and X-ray diffraction were evaluated. Ballistic tests performed in the selected materials showed that the ceramics present armor efficiency. (author)

  17. Histopathological evaluation of seven Amazon species of freshwater ornamental armored catfish

    Directory of Open Access Journals (Sweden)

    Rodrigo Yudi Fujimoto

    2014-08-01

    Full Text Available Fish commonly known as acaris or plecos are freshwater armored catfish economically important as a food resource and as ornamental fish. Most of these species are captured in the Amazon region. However, despite its economic importance, there is a lack of knowledge about their biological aspects. Thus, this study aimed to characterize and evaluate the histopathological aspects of important organs as gills, liver, integument and kidney of seven species of armored freshwater ornamental catfish fromGuamáRiver,Pará State,Brazil. All organs showed typical characteristics of organs of other teleosts. In some species, gills and liver showed slight histopathological changes: telangiectasis, edema and morphological changes related to the presence of parasites (Monogenea and Digenea in the gills, and changes in the arrangement of hepatocytes rows, and vacuolation of hepatocytes in the liver. Thus, the knowledge of the normal structure of organs and changes found can be used as tools for environmental and health monitoring of animals.

  18. A parametric design of ceramic faced composite armor subject to air weapon threats

    Science.gov (United States)

    Guo, Y. N.; Sun, Q.

    2015-12-01

    By taking into consideration the two categories of military projectile threats to aircraft structures, an optimal layer configuration of ceramic faced composite armor was designed in this paper. Using numerical simulations and the same layer arrangement of ceramic, UHMWPE, and carbon fiber laminates, a parametric finite element model using LS-DYNA code was built. Several thickness combinations were analyzed in order to determine the final lightest configuration that is capable of supporting a high-speed impact load and HEI blast wave load, which implements a high anti-penetration design for aircraft armor. This configuration can be used to improve the anti-impact ability of aircraft structures as well as achieve a structure/function integration design that considers a lighter weight.

  19. Macrocomposite mechanical design, modeling, and behavior of physical models of bioinspired fish armor

    Science.gov (United States)

    Browning, Ashley; Ortiz, Christine; Boyce, Mary C.

    2012-02-01

    The macrocomposite design of flexible biological exoskeletons, consisting of overlapping mineralized armor units embedded in a compliant tissue, is a key determinant of their mechanical function (e.g penetration resistance and biomechanical flexibility). Here, we investigate the role of macrocomposite structure, composition, geometric orientation, and spatial distribution in a flexible model natural armor system present in the majority of teleost fish species. Physical multi-material composite models are fabricated using a combination of 3-D printing and molding methods. Mechanical experiments using digital image correlation enable measurement of both the macroscopic response and underlying deformation mechanisms during various loading scenarios. Finite element-based mechanical models yield detailed insights into the roles and the tradeoffs of the composite structure providing constraint, shear, and bending mechanisms to impart protection and flexibility.

  20. An Advanced Probabilistic Neural Network for the Design of Breakwater Armor Blocks

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    In this study, an advanced probabilistic neural network (APNN) method is proposed to reflect the global probability density function (PDF) by summing up the heterogeneous local PDF which is automatically determined in the individual standard deviation of variables. The APNN is applied to predict the stability number of armor blocks of breakwaters using the experimental data of van der Meer, and the estimated results of the APNN are compared with those of an empirical formula and a previous artificial neural network (ANN) model. The APNN shows better results in predicting the stability number of armor blocks of breakwater and it provided the promising probabilistic viewpoints by using the individual standard deviation in a variable.

  1. Relationship between Ballistic Coefficient and Static Mechanical Properties for Armor Materials

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The relationship between the ballistic coefficient and the static mechanical properties of armor materials was studied. The results show that the ballistic coefficient is determined by the strength, hardness and the toughness of materials. According to the Martel rule, the equation of the relationship between ballistic coefficient and static mechanical properties satisfies the following formula: . From the mixture law of composite, the prerequisite, for which ballistic coefficient has maximum to reinforcement volume fraction, is obtained by the following equation: .

  2. Taxonomy and identification of the armored dinoflagellate genus Heterocapsa (Peridiniales, Dinophyceae)

    OpenAIRE

    Iwataki, Mitsunori

    2008-01-01

    The armored dinoflagellate genus Heterocapsa is composed of relatively small species, including a species responsible for harmful red tides, H. circularisquama. Some Heterocapsa species such as H. rotundata and H. triquetra have been well documented as red tide-forming species, but are not recognized as causing harmful effects. Following sequential shellfish mass mortalities in the coastal waters of western Japan due to H. circularisquama red tides, this species has attracted considerable int...

  3. Simulation on Dual-stream Transmission System of Unmanned Tracked Armored Vehicle Using ADAMS

    Directory of Open Access Journals (Sweden)

    Sun Wei

    2015-01-01

    Full Text Available For the dual-stream transmission system of unmanned tracked armored vehicle, simulation analysis is carried out. Using SolidWorks to establish three-dimensional model of its chassis, the result of the simulation is processed in AdAMS/Solver. The simulation results are showed in lines. Comparative analysis for each simulation lines is conducted, and it verifies the feasibility of the dual-stream transmission system.

  4. An Overview of CNT-Kevlar Nanocomposites in the Application of Body Armor

    Institute of Scientific and Technical Information of China (English)

    冉路

    2014-01-01

    The superlative mechanical properties of carbon nanotubes make them the filler material of choice for composite reinforcement. Carbon nanotubes have been considered to be one of the most researched materials of the 21st century. In this paper, carbon nanotubes reinforced Kevlar composite and its application in personal armor has been analyzed as well as the processing cost being evaluated. The advantages and disadvantages of the nanocomposite for the application were shortlisted. Current and future developmental trends for this application were discussed.

  5. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    International Nuclear Information System (INIS)

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  6. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; In, S. R.; Bae, Y. D. (and others)

    2006-02-15

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  7. Conceptual design of a water cooled breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pu, Yong; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Jia; Peng, ChangHong [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Ma, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Chen, Lei [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by {sup 6}Li(n,α)T reaction. Li{sub 2}TiO{sub 3} pebbles and Be{sub 12}Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li{sub 2}TiO{sub 3} and Be{sub 12}Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be{sub 12}Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option

  8. Conceptual design of a water cooled breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by 6Li(n,α)T reaction. Li2TiO3 pebbles and Be12Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li2TiO3 and Be12Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be12Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into

  9. Dynamic Lift on an Artificial Static Armor Layer During Highly Unsteady Open Channel Flow

    Directory of Open Access Journals (Sweden)

    Stephan Mark Spiller

    2015-09-01

    Full Text Available The dynamic lift acting on a 100 mm × 100 mm section of a static armor layer during unsteady flow is directly measured in a series of physical experiments. The static armor layer is represented by an artificial streambed mold, made from an actual gravel bed. Data from a total of 190 experiments are presented, undertaken in identical conditions. Results show that during rapid discharge increases, the dynamic lift on the streambed repeatedly exhibits three clear peaks. The magnitude of the observed lift depends on the following hydrograph characteristics: (1 the initial flow depth; (2 the ramping duration and therefore the ramping rate; and (3 the total discharge increase. An adjusted unsteadiness parameter combines those three hydrograph characteristics for rapid discharge increases. Direct correlations between the unsteadiness parameter and the measured dynamic lift during unsteady flow are presented. In addition, the armor layer porosity showed a major impact on the observed effects. It is shown that increasing bed porosity leads to decreasing dynamic lift.

  10. Stability of hard plates on soft substrates and application to the design of bioinspired segmented armor

    Science.gov (United States)

    Martini, R.; Barthelat, F.

    2016-07-01

    Flexible natural armors from fish, alligators or armadillo are attracting an increasing amount of attention from their unique and attractive combinations of hardness, flexibility and light weight. In particular, the extreme contrast of stiffness between hard plates and surrounding soft tissues give rise to unusual and attractive mechanisms, which now serve as model for the design of bio-inspired armors. Despite a growing interest in bio-inspired flexible protection, there is little guidelines as to the choice of materials, optimum thickness, size, shape and arrangement for the protective plates. In this work, we focus on a failure mode we recently observed on natural and bio-inspired scaled armors: the unstable tilting of individual scales subjected to off-centered point forces. We first present a series of experiments on this system, followed by a model based on contact mechanics and friction. We condense the result into a single stability diagram which capture the key parameters that govern the onset of plate tilting from a localized force. We found that the stability of individual plates is governed by the location of the point force on the plate, by the friction at the surface of the plate, by the size of the plate and by the stiffness of the substrate. We finally discuss how some of these parameters can be optimized at the design stage to produce bio-inspired protective systems with desired combination of surface hardness, stability and flexural compliance.

  11. Ballistic Studies on TiB2-Ti Functionally Graded Armor Ceramics

    Directory of Open Access Journals (Sweden)

    Neha Gupta

    2012-11-01

    Full Text Available The objective of this paper is to discuss the results of the ballistic testing of the spark plasma sintered TiB2-Ti based functionally graded materials (FGMs with an aim to assess their performance in defeating small-calibre armor piercing projectiles. We studied the efficacy of FGM design and compared its ballistic properties with those of TiB2-based composites as well as other competing ceramic armors. The ballistic properties are critically analyzed in terms of depth of penetration, ballistic efficiency, fractographs of fractured surfaces as well as quantification of the shattered ceramic fragments. It was found that all the investigated ceramic compositions exhibit ballistic efficiency (η of 5.1 -5.9. We also found that by increasing the thickness of FGM from 5mm to 7.8 mm, the ballistic property of the composite degraded. On comparing the results with available armor systems, it has been concluded that TiB2 based composites show better ballistic properties except B4C. SEM analysis of the fragments showed that the FGM fractured by mixed mode of failure.Defence Science Journal, 2012, 62(6, pp.382-389, DOI:http://dx.doi.org/10.14429/dsj.62.2666

  12. Rapid molecular identification of armored scale insects (Hemiptera: Diaspididae) on Mexican 'Hass' avocado.

    Science.gov (United States)

    Rugman-Jones, Paul F; Morse, Joseph G; Stouthamer, Richard

    2009-10-01

    'Hass' avocado, Persea americana Mill., fruit imported into California from Mexico are infested with high levels of armored scale insects (Hemiptera: Diaspididae), constituting several species. The paucity and delicate nature of morphological characters traditionally used to diagnose armored scales often require careful preparation of slide-mounted specimens and expert knowledge of the group, for their accurate identification. Here, we present a simple, quick, and accurate means to identify armored scales on Mexican avocados, based on amplification of the internal transcribed spacer two of ribosomal DNA, by using the polymerase chain reaction (PCR). This region seems to show a high level of intraspecific conformity among scale specimens originating from different localities. A suite of species-specific reverse PCR primers are combined in a single reaction, with a universal forward primer, and produce a PCR product of a unique size, that after standard gel electrophoresis, allows the direct diagnosis of six diaspidid species: Abgrallaspis aguacatae Evans, Watson & Miller; Hemiberlesia lataniae (Signoret); Hemiberlesia sp. near latania; Hemiberlesia rapax (Comstock); Acutaspis albopicta (Cockerell); and Pinnaspis strachani (Cooley). Two additional species, Diaspis miranda (Cockerell) and Diaspis sp. near miranda, also are separated from the others by using this method and are subsequently diagnosed by secondary digestion of the PCR product with the restriction endonuclease smaI.

  13. Target/Blanket Design for the Accelerator Production of Tritium Plant

    International Nuclear Information System (INIS)

    The Accelerator Production of Tritium Target/Blanket (T/B) system is comprised of an assembly of tritium-producing modules supported by safety, heat removal, shielding, and retargeting systems. The T/B assembly produces tritium using a high-energy proton beam, a tungsten/lead spallation neutron source and 3He gas as the tritium-producing feedstock. The supporting heat removal systems remove the heat deposited by the proton beam during both normal and off-normal conditions. The shielding protects workers from ionizing radiation, and the retargeting systems remove and replace components that have reached their end of life. All systems reside within the T/B building, which is located at the end of a linear accelerator. For the nominal production mode, protons are accelerated to an energy of 1030 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons are expanded to a 0.19- x 1.9-m beam spot before striking a centrally located tungsten neutron source. A surrounding lead blanket produces additional neutrons from scattered high-energy particles. A total of 27 neutrons are produced per incident proton. Tritium is produced by neutron capture in 3He gas that is contained in aluminum tubes throughout the blanket. The 3He/tritium mixture is removed on a semi-continuous basis for purification in an adjacent Tritium Separation Facility. Systems and components are designed with safety as a primary consideration to minimize risk to the workers and the public. Materials and component designs were chosen based on the experiences of operating spallation neutron sources that have been designed and built for the neutron science community. An extensive engineering development and demonstration program provides detailed information for the completion of the design

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

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fratoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-11-20

    Pre-conceptual fusion blanket designs require research and development to reflect important proposed changes in the design of essential systems, and the new challenges they impose on related fuel cycle systems. One attractive feature of using liquid lithium as the breeder and coolant is that it has very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns. If the chemical reactivity of lithium could be overcome, the result would have a profound impact on fusion energy and associated safety basis. The overriding goal of this project is to develop a lithium-based alloy that maintains beneficial properties of lithium (e.g. high tritium breeding and solubility) while reducing overall flammability concerns. To minimize the number of alloy combinations that must be explored, only those alloys that meet certain nuclear performance metrics will be considered for subsequent thermodynamic study. The specific scope of this study is to evaluate the neutronics performance of lithium-based alloys in the blanket of an inertial confinement fusion (ICF) engine. The results of this study will inform the development of lithium alloys that would guarantee acceptable neutronics performance while mitigating the chemical reactivity issues of pure lithium.

  15. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    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

  16. Characteristics of behind armor blunt trauma produced by bullets with different structural materials: an experimental study

    Directory of Open Access Journals (Sweden)

    Ling-qing WANG

    2013-07-01

    Full Text Available Objective To investigate the effect of structural materials of bullets on behind armor blunt trauma (BABT. Methods Ten healthy male Landraces were randomly divided into two groups (5 each: 56 type 7.62-mm rifle bullet group and SS109 5.56-mm rifle bullet group. The kinetic energy of two types of bullets was adjusted to the same level (about 1880J by the way of grow downwards gunpowder. Then the animals as protected with both grade NIJ Ⅲ ceramic hard armor and grade Ⅱ police soft body armor, were shot at the left midclavicular line of fourth intercostal space (shooting distance was 25m. The damage to the animals was observed. Other 2 healthy male Landraces were selected, and biomechanical sensor was subcutaneously implanted into the soft tissue in precordium and intracalvarium to detect the pressure at the time point of bullet contact under the protection of armor, and the relationship between pressure and damage was analyzed. Results Respiration, heart rate and systolic arterial pressure of animals in two groups were all elevated after injury, but there was no significant difference between the two groups. No obvious change was found on blood oxygen saturation of both groups. Gross anatomy showed the predominant local injury was cardiac and pulmonary contusions. The area of pulmonary hemorrhage of 7.62mm group was 6.00%±3.18%, significantly higher than that of 5.56mm group (3.59%±2.11%, P<0.05. Histopathological examination revealed acute injuries of lung tissues, myocardial tissue and cerebral cortical neurons. The contents of cardiac troponin T (TnT, creatine kinase (CK and creatine kinase-MB (CK-MB isoenzyme were all increased 3 hours after injury, and the rise was higher in 7.62mm group than in 5.56mm group (P<0.05. Biomechanical testing showed the pressure of precordium and intracalvarium was elevated at the moment of bullet contact, and the rise was higher in 7.62mm group than in 5.56mm group (P<0.05. Conclusions

  17. Analysis of heat conduction in a drum brake system of the wheeled armored personnel carriers

    Science.gov (United States)

    Puncioiu, A. M.; Truta, M.; Vedinas, I.; Marinescu, M.; Vinturis, V.

    2015-11-01

    This paper is an integrated study performed over the Braking System of the Wheeled Armored Personnel Carriers. It mainly aims to analyze the heat transfer process which is present in almost any industrial and natural process. The vehicle drum brake systems can generate extremely high temperatures under high but short duration braking loads or under relatively light but continuous braking. For the proper conduct of the special vehicles mission in rough terrain, we are talking about, on one hand, the importance of the possibility of immobilization and retaining position and, on the other hand, during the braking process, the importance movement stability and reversibility or reversibility, to an encounter with an obstacle. Heat transfer processes influence the performance of the braking system. In the braking phase, kinetic energy transforms into thermal energy resulting in intense heating and high temperature states of analyzed vehicle wheels. In the present work a finite element model for the temperature distribution in a brake drum is developed, by employing commercial finite element software, ANSYS. These structural and thermal FEA models will simulate entire braking event. The heat generated during braking causes distortion which modifies thermoelastic contact pressure distribution drum-shoe interface. In order to capture the effect of heat, a transient thermal analysis is performed in order to predict the temperature distribution transitional brake components. Drum brakes are checked both mechanical and thermal. These tests aim to establish their sustainability in terms of wear and the variation coefficient of friction between the friction surfaces with increasing temperature. Modeling using simulation programs led eventually to the establishment of actual thermal load of the mechanism of brake components. It was drawn the efficiency characteristic by plotting the coefficient of effectiveness relative to the coefficient of friction shoe-drum. Thus induced

  18. MFTF-B Upgrade for blanket-technology testing

    International Nuclear Information System (INIS)

    Based on preliminary studies at Lawrence Livermore National Laboratory (LLNL), we believe the Mirror Fusion Test Facility (MFTF-B) could be upgraded for operation in a hot-ion Kelley mode in a portion of the central cell to provide fusion nuclear engineering data, particularly blanket technology information, by the end of the decade. Cost of this mode of operation would be modest compared with that of the other fusion devices considered in the last few years for such purposes

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

    Science.gov (United States)

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

    1988-01-01

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

  20. Model problem of MHD flow in a lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Cherepanov, V.Y.

    1978-01-01

    A model problem is considered for a feasibility study concerning controlled MHD flow in the blanket of a Tokamak nuclear reactor. The fundamental equations for the steady flow of an incompressible viscous fluid in a uniform transverse magnetic field are solved in rectangular coordinates, in the zero-induction approximation and with negligible induced currents. A numerical solution obtained for a set of appropriate boundary constraints establishes the conditions under which no stagnation zones will be formed.

  1. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

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

  2. The feasibility study I on the blanket fuel options for the ATW/HYPER

    International Nuclear Information System (INIS)

    The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended

  3. The feasibility study I on the blanket fuel options for the ATW/HYPER

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok; Meyer, M.K; Hayes, S.L

    2001-01-01

    The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended.

  4. APT Blanket Safety Analysis: Preliminary Analyses of Downflow Through a Lateral Row 1 Blanket Model Under Near RHR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    To address a concern about a potential maldistribution of coolant flow through an APT blanket module under low flow near RHR conditions, a scoping study of downflow mixed convection in parallel channels was conducted. Buoyancy will adversely effect the flow distribution in module bins with downflow and non-uniform power distributions. This study consists of two parts: a simple analytical model of flow in a two channel network, and a lumped eleven channel FLOWTRAN-TF model of a front lateral Row-1 blanket module bin. Results from both models indicate that the concern about coolant flow in a vertical model being diverted away from high power regions by buoyancy is warranted. The FLOWTRAN-TF model predicted upflow (i.e., a flow reversal) through several of the high power channels, under some low flow conditions. The transition from the regime with downflow in all channels to a regime with upflow in some channels was abrupt.

  5. Blanket concept with liquid Li/sub 17/Pb/sub 83/ for tritium breeding in INTOR-NET

    Energy Technology Data Exchange (ETDEWEB)

    Airola, J.; Biggio, M.; Casini, G.; Farfaletti-Casali, F.; Li Bassi, P.; Ponti, C.; Rieger, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Piana, C. (Milan Univ. (Italy))

    1984-04-01

    A blanket concept with eutectic Li/sub 17/Pb/sub 83/ as liquid breeder, suited for tritium production in an experimental Tokamak power reactor is outlined and discussed. This design has been developed to satisfy the INTOR-Phase-I specifications, in particular: (I) modular arrangement of the blanket units inside the vacuum vessel; (II) no use of the heat deposited for electricity production, (III) a net tritium breeding of a least 60%. In this article the main results of the neutronics and thermohydraulics analysis are reviewed and the problems identified. Methods to keep liquid in the breeder during operation are proposed and discussed. The consequences of a coolant tube rupture in a breeder unit appears to be the most serious problem.

  6. Axial blanket fuel design and demonstration. First semi-annual progress report, January-September 1980

    International Nuclear Information System (INIS)

    The axial blanket fuel design in this program, which is retrofittable in operating pressurized water reactors, involves replacing the top and bottom of the enriched fuel column with low-enriched (less than or equal to 1.0 wt % 235U) fertile uranium. This repositioning of the fissile inventory in the fuel rod leads to decreased axial leakage and increased discharge burnups in the enriched fuel. Various axial blanket fuel designs, with blanket thicknesses from 0 to 10 inches and blanket enrichments from 0.2 to 1.0 wt % 235U, were investigated to determine the relationship between uranium utilization and power peaking. Analyses were preformed to assess the nuclear, mechanical, and thermal-hydraulic effects arising from the use of axial blankets. Four axial blanket lead test assemblies are being fabricated for scheduled irradiation in cycle 5 of Sacramento Municipal Utility District's Rancho Seco pressurized water reactor. Analyses to support licensing cycle 5 are in progress

  7. The impact of new experimental data on the design of Pb-17Li/water breeding blankets

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1989-04-01

    The Pb-17Li/water-cooled blanket is one of the concepts being developed in Europe for testing in NET (Next European Torus). JRC-Ispra is strongly involved in this development. This paper describes the impact of the latest experimental results on the blanket design. The points considered are: breeder operating temperature and thermomechanical design: experiments on corrosion with steel 316L and liquid metal embrittlement tests have provided upper and lower limits for the breeder operating temperature (280-400/sup 0/C); tritium recovery from the breeder and permeation rate to the coolant: Ispra measurements indicate that solubility and diffusivity of hydrogen in Pb-17Li are lower as compared with the previous values used in blanket tritium analyses. The impact of these results on the design of the tritium recovery system is discussed; accident analyses: the experiments in progress at Ispra on the Pb-17Li/water interaction are reviewed and their application to a coolant pipe break accident is shown. (orig.).

  8. Study on Fission Blanket Fuel Cycling of a Fusion-Fission Hybrid Energy Generation System

    International Nuclear Information System (INIS)

    Full text: Direct application of ITER-scale tokamak as a neutron driver in a subcritical fusion-fission hybrid reactor to generate electric power is of great potential in predictable future. This paper reports a primary study on neutronic and fuel cycle behaviors of a fission blanket for a new type of fusion-driven system (FDS), namely a subcritical fusion-fission hybrid reactor for electric power generation aiming at energy generation fueled with natural or depleted uranium. Using COUPLE2 developed at INET of Tsinghua University by coupling the MCNP code with the ORIGEN code to study the neutronic behavior and the refueling scheme, this paper focuses on refueling scheme of the fissionable fuel while keeping some important parameters such as tritium breeding ratio (TBR) and energy gain. Different fission fuels, coolants and their volumetric ratios arranged in the fission blanket satisfy the requirements for power generation. The results show that soft neutron spectrum with optimized fuel to moderator ratio can yield an energy amplifying factor of M> 20 while maintaining the TBR > 1.1 and the CR > 1 (the conversion ratio of fissile materials) in a reasonably long refueling cycle. Using an in-site fuel recycle plant, it will be an attractive way to realize the goal of burning 238U with such a new type of fusion-fission hybrid reactor system to generate electric power. (author)

  9. Application Effect’s Research of Vetiver Eco Blanket in Pubugou Reservoir Fluctuating Zone

    Directory of Open Access Journals (Sweden)

    Lan Huijuan

    2015-01-01

    Full Text Available To solve the ecological disasters in Pubugou Reservoir Fluctuating Zone, ecological blanket governance model is proposed in this paper, which may provide good early environment for plants’ survival in fluctuation zone, and then play the function of greening and sustainable development to ensure the slopes’ stability. Meanwhile, based on the result of vetiver ecological blanket in Hanyuan experimental zone, we find that three kinds of typical Fluctuating Zone slope’s greening effect is good, which includes the dirt piling up slope, the whole lump of rock slope and the gravel piling up slope, and it gets an average coverage of 90.3 % as well as good strength. Due to the different geological conditions, the ecological blankets’ governance effect differs from slope to slope. Using analytic hierarchy process to calculate the weight, we get the dirt piling up slope, the whole lump of rock slope and the gravel piling up slope’s weights are 0.41, 0.17, 0.42, respectively, namely, the dirt piling up slope and the gravel piling up slope have good results overall, followed by the whole lump of rock slope.

  10. Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Highlights: ► HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM) to be tested in ITER. ► Tritium extraction by gas purging, removal and transfer to the Tritium Plant. ► Conceptual design of TES and revision of the previous configuration. ► Main components: adsorption column, ZrCo getter beds and PERMCAT reactor. - Abstract: The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.

  11. Thermal-hydraulic and neutronic considerations for designing a lithium-cooled tokamak blanket

    International Nuclear Information System (INIS)

    A methodology for the design of lithium cooled blankets is developed. The thermal-hydraulics, neutronics and interactions between them are extensively investigated. In thermal hydraulics, two models illustrate the methodology used to obtain the acceptable ranges for a set of design parameters. The methodology can be used to identify the limiting constraints for a particular design. A complete neutronic scheme is set up for the calculations of the volumetric heating rate as a function of the distance from the first wall, the breeding ratio as a function of the amount of structural material in the blanket, and the radiation damage in terms of atom displacements and gas production rate. Different values of the volume percent of Type-316 stainless steel are assigned in four breeding zones to represent a nonuniformly distributed structural material which satisfies various thermal-hydraulic requirements. The role that the radiation damage plays in the overall design methodology is described. The product of the first wall lifetime and neutron loading is limited by the radiation damage which degrades the mechanical properties of the material

  12. Application of the Integrated-Blanket-Coil concept to a compact reversed-field pinch reactor

    International Nuclear Information System (INIS)

    The Integrated-Blanket-Coil (IBC) concept has been examined in the context of a compact reversed-field pinch (RFP) fusion reactor. The IBC approach is novel in that the functions of the blanket (tritium breeding and energy recovery) and the coil (magnetic field production) are fulfilled in a single component. This combination of functions is accomplished by using lithium metal as the coolant, breeding medium, and electrical conductor. Economics and physics modeling indicates that the toroidal field and divertor coil systems are appropriate applications for IBC components. Conceptual designs for the TF-IBC and IBC divertor systems are developed, based on parameters generated by the TITAN RFP Reactor Design Study. Design of the IBC divertor is similar to the TF-IBC, but with the added concern for proper mapping of the field lines. Improved magnetic coupling and additional energy recovery and tritium breeding enhance the attractiveness of the IBC divertor relative to copper coils. Both the TF and divertor IBC systems are capable of operating compatibly with the Oscillating Field Current Drive (OFCD). The conceptual design process indicates that the TF-IBC and IBC divertor are technically feasible. As such, they represent viable alternatives for a compact RFP reactor

  13. Fusion-Driven Sub-Critical Dual-Cooled Waste Transmutation Blanket:Design and Analysis

    Institute of Scientific and Technical Information of China (English)

    Wang Weihua(汪卫华); Wu Yican(吴宜灿); Ke Yan(柯严); Kang Zhicheng(康志诚); Wang Hongyan(王红艳); Huang Qunying(黄群英)

    2003-01-01

    The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB),as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermalhydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account.All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis.

  14. A fail–safe and cost effective fabrication route for blanket First Walls

    Energy Technology Data Exchange (ETDEWEB)

    Commin, L., E-mail: lorelei.commin@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-AWP), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rieth, M.; Dafferner, B.; Zimmermann, H.; Bolich, D.; Baumgärtner, S.; Ziegler, R. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-AWP), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dichiser, S.; Fabry, T.; Fischer, S.; Hildebrand, W.; Palussek, O.; Ritz, H.; Sponda, A. [Karlsruhe Institute of Technology (KIT), Technische Infrastruktur und Dienste (TID), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2013-11-15

    Helium Cooled Lithium Lead and Helium Cooled Pebble Bed concepts have been selected as European Test Blanket Modules (TBM) for ITER. The TBM fabrication will need the assembly of six Reduced Activation Ferritic Martensitic steel sub-components, namely First Wall, Caps, Stiffening Grid, Breeding Units, Back Plates/Manifolds, and Attachment system. The fabrication of the First Wall requires the production of cooling channels inside 30 mm thick bended plates. For this specific component, the main issues consist of the lack of accessibility of some areas to join, the process tolerances, the dimensional stability and the resulting assembly mechanical properties. Several fabrication routes have been already investigated, which involve diffusion welding and fusion welding (electron beam, laser beam, hybrid MIG/laser). In this study, an alternative processing method was developed, based on Hot Isostatic Pressing of inner pipes within two half-shells. This method presents some major advantages over the existing ones, in particular its inherent fail–safe design due to the application of the double containment principle, the solely use of cost effective standard fabrication processes and the resulting component dimensional stability. A four channel mock-up was fabricated and analyzed to validate the fabrication procedure. The joint quality was assessed using microstructural characterization and Charpy tests. The results confirm the predicted perfect weld lines as well as the preservation of the mechanical properties. Therefore, the presented fabrication procedure is very appropriate for the fabrication of First Walls for fusion reactor blankets.

  15. Study on electromagnetic-structural behavior of first wall/blanket structure for tokamak fusion reactor

    International Nuclear Information System (INIS)

    The electromagnetic problems related to the structural design of the first wall/blanket structure, which is a major component of Fusion Reactor, have been studied. The electromagnetic load, which is characteristic and very important of Tokamak type, is necessary for the evaluation of the structural integrity at the last item of the design process. A transient electromagnetic phenomena, which include the measurement of the eddy current obtained by the simulated plasma disruption experiment, the vibration behavior of the beam-plate by the dynamic electromagnetic load and the verification of the numerical codes, have been clarified. A static electromagnetic phenomena have been studied to evaluate the applicability of the ferromagnetic material to the first wall/blanket structure of Tokamak Power Reactor. The numerical code, which can calculate the magnetic field of the finite ferromagnetic body, has been developed and the magnetic field distortions inside and outside the materials has been studied. The deformation by the magnetic torque, which generates inside the ferromagnetic material placed in the magnetic field, has been studied. The effects of the magnetic stiffness and the saturated magnetic field to the deformation has been also clarified. (author)

  16. Fabrication of ITER Semi-Prototype Blanket First Wall for the Final Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byoung Kwon; Jung, Yang Il; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Suk Kwon; Lee, Don Won; Kim, Duck Hoi; Cho, Seung Yon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The ITER semi-prototype was designed to qualify the manufacturing technology for the ITER blanket first wall. According to the design of the semi-prototype, its fabrication is expected to face great difficulty. The blanket first wall consists of three different materials, i.e., beryllium (Be), CuCrZr, and stainless-steel (SS), which are joined into one part. For fabrication of these multi-layered structures, hot isostatic pressing (HIP), which is one of the diffusion bonding methods, has been considered as a promising technology to realize sufficient mechanical integrity of a joint under the anticipated high neutron and stress fields. HIP provides high dimensional accuracy, low residual stress during the joining process, and the joining of three-dimensionally complex structures in comparison with other joining methods. Even though the joining technology for the different materials had been developed in the first stage of the qualification, the joining is still a key issue for the fabrication of the semi-prototype

  17. Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ciampichetti, A., E-mail: andrea.ciampichetti@enea.it [ENEA CR Brasimone, 40032 Camugnano (Italy); European TBM Consortium of Associates (Germany); Nitti, F.S.; Aiello, A. [ENEA CR Brasimone, 40032 Camugnano (Italy); European TBM Consortium of Associates (Germany); Ricapito, I. [Fusion for Energy, 08019 Barcelona (Spain); Liger, K. [CEA, DEN, DTN/STPA/LIPC, Cadarache, 13108 St. Paul-lez-Durance (France); European TBM Consortium of Associates (Germany); Demange, D. [Karlsruhe Institute of Technology, ITEP-TLK, Postfach 36 40, 76021 Karlsruhe (Germany); European TBM Consortium of Associates (Germany); Sedano, L.; Moreno, C. [EURATOM-CIEMAT Association, 28040 Madrid (Spain); European TBM Consortium of Associates (Germany); Succi, M. [SAES Getters Spa, 20020 Lainate (Italy)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM) to be tested in ITER. Black-Right-Pointing-Pointer Tritium extraction by gas purging, removal and transfer to the Tritium Plant. Black-Right-Pointing-Pointer Conceptual design of TES and revision of the previous configuration. Black-Right-Pointing-Pointer Main components: adsorption column, ZrCo getter beds and PERMCAT reactor. - Abstract: The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.

  18. 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.

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

    International Nuclear Information System (INIS)

    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.

  20. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

  3. A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

    International Nuclear Information System (INIS)

    Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

  4. Hazard report update. ECRI Institute revises its recommendation for temperature limits on blanket warmers.

    Science.gov (United States)

    2009-07-01

    ECRI Institute now recommends that temperature settings on blanket warming cabinets be limited to 130 degrees F (54 degrees C). We had previously recommended a limit of 110 degrees F (43 degrees C) because solutions were often being warmed in the same cabinets as blankets, and the lower temperature eliminated the serious burn risk presented by excessively heated solutions. With increasing recognition in the healthcare community that solutions should be kept at lower temperatures than--and therefore heated separately from--blankets, we believe that our recommendation for blankets can be made less stringent. We continue to recommend that solution warming cabinets be limited to 110 degrees F. PMID:20848953

  5. Vacuum UV spectroscopy of armor erosion from plasma gun disruption simulation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rockett, P.D. [Sandia Nat. Labs., Albuquerque, NM (United States). Fusion Tech. Dept.; Hunter, J.A. [Sandia Nat. Labs., Albuquerque, NM (United States). Fusion Tech. Dept.; Bradley, J.T. III [Electrical Engineering and Computer Engineering Department, University of New Mexico, Albuquerque, NM 87131 (United States); Gahl, J.M. [Electrical Engineering and Computer Engineering Department, University of New Mexico, Albuquerque, NM 87131 (United States); Zhitlukhin, A. [Troitsk Institute for Innovation and Technology (TRINITI), Troitsk, Moscow Region (Russian Federation); Arkhipov, K. [Troitsk Institute for Innovation and Technology (TRINITI), Troitsk, Moscow Region (Russian Federation); Bakhtin, V. [Troitsk Institute for Innovation and Technology (TRINITI), Troitsk, Moscow Region (Russian Federation); Toporkov, D. [Troitsk Institute for Innovation and Technology (TRINITI), Troitsk, Moscow Region (Russian Federation); Ovchinnokov, I. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kuznetsov, V.E. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Titov, V.A. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation)

    1995-03-01

    Extensive simulations of tokamak disruptions have provided a picture of material erosion that is limited by the transfer of energy from the incident plasma to the armor solid surface through a dense vapor shield. Two transmission grating vacuum ultraviolet (VUV) spectrographs were designed and utilized to study the plasma-material interface in plasma gun simulation experiments. Target materials included POCO graphite, ATJ graphite, boron nitride and plasma-sprayed tungsten. Detailed spectra were recorded with a spatial resolution of ca. 0.7mm resolution on VIKA at Efremov and on 2MK-200 at Troitsk. Time-resolved data with 40-200ns resolution were then recorded along with the same spatial resolution on 2MK-200. The VIKA plasma gun directly illuminated a target with a high-intensity plasma pulse of 2-100MJm{sup -2} with low-energy ions of ca. 100eV. The 2MK-200 plasma gun illuminated the target via a magnetic cusp that permitted only deuterium to pass with energies of ca. 1keV, but which produced a fairly low intensity of 2MJm{sup -2}. Power densities on target ranged from 10{sup 7} to 10{sup 8}Wcm{sup -2}. Emitted spectra were recorded from 15 to 450A over a distance from 0 to 7cm above the armor target surface. The data from both plasma gun facilities demonstrated that the hottest plasma region was sitting several millimeters above the armor tile surface. This apparently constituted the absorption region, which confirmed past computer simulations. Spectra indicated both the species and ionization level that were being ablated from the target, demonstrating impurity content, and showing plasma ablation velocity. Graphite samples clearly showed CV lines as well as impurity lines from O V and O VI. The BN tiles produced textbook examples of BIV and BV, and extensive NIV, V and VI lines. These are being compared with radiation-hydrodynamic calculations. (orig.).

  6. Gas Metal Arc Welding Process Modeling and Prediction of Weld Microstructure in MIL A46100 Armor-Grade Martensitic Steel

    Science.gov (United States)

    Grujicic, M.; Arakere, A.; Ramaswami, S.; Snipes, J. S.; Yavari, R.; Yen, C.-F.; Cheeseman, B. A.; Montgomery, J. S.

    2013-06-01

    A conventional gas metal arc welding (GMAW) butt-joining process has been modeled using a two-way fully coupled, transient, thermal-mechanical finite-element procedure. To achieve two-way thermal-mechanical coupling, the work of plastic deformation resulting from potentially high thermal stresses is allowed to be dissipated in the form of heat, and the mechanical material model of the workpiece and the weld is made temperature dependent. Heat losses from the deposited filler-metal are accounted for by considering conduction to the adjoining workpieces as well as natural convection and radiation to the surroundings. The newly constructed GMAW process model is then applied, in conjunction with the basic material physical-metallurgy, to a prototypical high-hardness armor martensitic steel (MIL A46100). The main outcome of this procedure is the prediction of the spatial distribution of various crystalline phases within the weld and the heat-affected zone regions, as a function of the GMAW process parameters. The newly developed GMAW process model is validated by comparing its predictions with available open-literature experimental and computational data.

  7. Biological armors under impact—effect of keratin coating, and synthetic bio-inspired analogues

    International Nuclear Information System (INIS)

    A number of biological armors, such as turtle shells, consist of a strong exoskeleton covered with a thin keratin coating. The mechanical role upon impact of this keratin coating has surprisingly not been investigated thus far. Low-velocity impact tests on the turtle shell reveal a unique toughening phenomenon attributed to the thin covering keratin layer, the presence of which noticeably improves the fracture energy and shell integrity. Synthetic substrate/coating analogues were subsequently prepared and exhibit an impact behavior similar to the biological ones. The results of the present study may improve our understanding, and even future designs, of impact-tolerant structures. (paper)

  8. The impact of blanket design on activation and thermal safety

    International Nuclear Information System (INIS)

    Activation and thermal safety analyses for experimental and power reactors are presented. The effects of a strong neutron absorber, B4C, on activation and temperature response of experimental reactors to Loss-of-Cooling Accidents are investigated. Operational neutron fluxes, radioactivities of elements and thermal transients are calculated using the codes ONEDANT, REAC and THIOD, respectively. The inclusion of a small amount of B4C in the steel blanket of an experimental reactor reduces its activation and the post LOCA temperature escalation significantly. Neither the inclusion of excessive amounts of B4C nor enriched 10B in the first walls of an experimental reactor bring much advantage. The employment of a 2 cm graphite tile liner before the first wall helps to limit the post LOCA escalation of first wall temperature. The effect of replacing a 20 cm thick section of a steel shield of a fusion power reactor with B4C is also analyzed. The first wall temperature peak is reduced by 100 degree C in the modified blanket. The natural convection effect on thermal safety of a liquid lithium cooled blanket are investigated. Natural convection has no impact at all, unless the magnetic field can be reduced. If magnets can be shut off rapidly after the accident, then the temperature escalation of the first wall will be limited. Upflow of the coolant is better than the initial downflow design from a thermal safety point of view. Activities of three structural materials, OTR stainless steel, SS-316 and VCrTi are compared. Although VCrTi has higher activity for a period of two hours after the accident, it has one to two orders of magnitude less activity than those of the steels in the mid- and long-terms. 29 refs., 42 figs., 9 tabs

  9. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li3N, Li2O, and Li2C2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

  10. Impact of prescribed burning on blanket peat hydrology

    Science.gov (United States)

    Holden, Joseph; Palmer, Sheila M.; Johnston, Kerrylyn; Wearing, Catherine; Irvine, Brian; Brown, Lee E.

    2015-08-01

    Fire is known to impact soil properties and hydrological flow paths. However, the impact of prescribed vegetation burning on blanket peatland hydrology is poorly understood. We studied 10 blanket peat headwater catchments. Five were subject to prescribed burning, while five were unburnt controls. Within the burnt catchments, we studied plots where the last burn occurred ˜2 (B2), 4 (B4), 7 (B7), or greater than 10 years (B10+) prior to the start of measurements. These were compared with plots at similar topographic wetness index locations in the control catchments. Plots subject to prescribed vegetation burning had significantly deeper water tables (difference in means = 5.3 cm) and greater water table variability than unburnt plots. Water table depths were significantly different between burn age classes (B2 > B4 > B7 > B10+) while B10+ water tables were not significantly different to the unburnt controls. Overland flow was less common on burnt peat than on unburnt peat, recorded in 9% and 17% of all runoff trap visits, respectively. Storm lag times and hydrograph recession limb periods were significantly greater (by ˜1 and 13 h on average, respectively) in the burnt catchments overall, but for the largest 20% of storms sampled, there was no significant difference in storm lag times between burnt and unburnt catchments. For the largest 20% of storms, the hydrograph intensity of burnt catchments was significantly greater than those of unburnt catchments (means of 4.2 × 10-5 and 3.4 × 10-5 s-1, respectively), thereby indicating a nonlinear streamflow response to prescribed burning. Together, these results from plots to whole river catchments indicate that prescribed vegetation burning has important effects on blanket peatland hydrology at a range of spatial scales.

  11. Recovery of tritium from a liquid lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Talbot, J.B.

    1981-01-01

    The sorption of tritium on yttrium from liquid lithium and the subsequent release of tritium from yttrium by thermal regeneration of the metal sorbent were investigated to study such a tritium-recovery process for a fusion reactor blanket of liquid lithium. Recent static sorption experiments have shown the effects of lithium temperature and possible impurities on the sorption of tritium. Diffusivity data, obtained from previous tritium recovery experiments, were evaluated to show the importance of the yttrium surface condition in controlling the release of tritium.

  12. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

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

  13. Nitriding treatment of reduced activation ferritic steel as functional layer for liquid breeder blanket

    International Nuclear Information System (INIS)

    The development of functional layers such as a tritium permeation barrier and an anti-corrosion layer is the essential technology for the development of a molten salt type self cooled fusion blanket. In the present study, the characteristics of a nitriding treatment on a reduced activation ferritic steel, JLF-1 (Fe-9Cr-2W-0.1C) as the functional layer were investigated. The steel surface was nitrided by an ion nitriding treatment or a radical nitriding treatment. The nitridation characteristic of the steel surface was made clear based on the thermodynamic stability. The thermal diffusivity, the hydrogen permeability and the chemical stability in the molten salt Flinak were investigated. The results indicated that the nitriding treatment can improve the compatibility in the Flinak without the decrease of the thermal diffusivity, though there was little improvement as the hydrogen permeation barrier. (author)

  14. Verification test results of a cutting technique for the ITER blanket cooling pipes

    International Nuclear Information System (INIS)

    For replacement of the first wall (FW) of the international thermonuclear experimental reactor (ITER), cutting and welding tools for the cooling pipes must be able to access a pipe from the surface side of the FW and cut/weld the pipe from the inside the cooling pipe (inner diameter: 42.72 mm, thickness: 2.77 mm). The cutting tool for the pipe end is required to cut a flat plate circularly from the surface side of the FW (cutting diameter: approximately 44 mm, plate thickness: 5 mm). To determine the specifications for both the tools and the blanket hydraulic connections, the ITER Organization (IO) and the Japan Domestic Agency (JADA) conducted research and development activities regarding the FW replacement. This paper describes the current status of the development of cutting tools for the cooling pipe connection.

  15. Cratering Experiments on the Self Armoring of Coarse-Grained Granular Targets

    CERN Document Server

    Güttler, Carsten; Nakamura, Akiko M

    2012-01-01

    Recently published crater statistics on the small asteroids 25143 Itokawa and 433 Eros show a significant depletion of craters below approx. 100 m in diameter. Possible mechanisms that were brought up to explain this lack of craters were seismic crater erasure and self armoring of a coarse, boulder covered asteroid surface. While seismic shaking has been studied in this context, the concept of armoring lacks a deeper inspection and an experimental ground truth. We therefore present cratering experiments of glass bead projectiles impacting into granular glass bead targets, where the grain sizes of projectile and target are in a similar range. The impact velocities are in the range of 200 to 300 m/s. We find that craters become fainter and irregular shaped as soon as the target grains are larger than the projectile sizes and that granular craters rarely form when the size ratio between projectile and target grain is around 1:10 or smaller. In that case, we observe a formation of a strength determined crater in ...

  16. Codevelopment of conceptual understanding and critical attitude: toward a systemic analysis of the survival blanket

    Science.gov (United States)

    Viennot, Laurence; Décamp, Nicolas

    2016-01-01

    One key objective of physics teaching is the promotion of conceptual understanding. Additionally, the critical faculty is universally seen as a central quality to be developed in students. In recent years, however, teaching objectives have placed stronger emphasis on skills than on concepts, and there is a risk that conceptual structuring may be disregarded. The question therefore arises as to whether it is possible for students to develop a critical stance without a conceptual basis, leading in turn to the issue of possible links between the development of conceptual understanding and critical attitude. In an in-depth study to address these questions, the participants were seven prospective physics and chemistry teachers. The methodology included a ‘teaching interview’, designed to observe participants’ responses to limited explanations of a given phenomenon and their ensuing intellectual satisfaction or frustration. The explanatory task related to the physics of how a survival blanket works, requiring a full and appropriate system analysis of the blanket. The analysis identified five recurrent lines of reasoning and linked these to judgments of adequacy of explanation, based on metacognitive/affective (MCA) factors, intellectual (dis)satisfaction and critical stance. Recurrent themes and MCA factors were used to map the intellectual dynamics that emerged during the interview process. Participants’ critical attitude was observed to develop in strong interaction with their comprehension of the topic. The results suggest that most students need to reach a certain level of conceptual mastery before they can begin to question an oversimplified explanation, although one student’s replies show that a different intellectual dynamics is also possible. The paper ends with a discussion of the implications of these findings for future research and for decisions concerning teaching objectives and the design of learning environments.

  17. Beryllium usage in fusion blankets and beryllium data needs

    International Nuclear Information System (INIS)

    Increasing numbers of designers are choosing beryllium for fusion reactor blankets because it, among all nonfissile materials, produces the highest number (2.5 neutron in an infinite media) of neutrons per 14-MeV incident neutron. In amounts of about 20 cm of equivalent solid density, it can be used to produce fissile material, to breed all the tritium consumed in ITER from outboard blankets only, and in designs to produce Co-60. The problem is that predictions of neutron multiplication in beryllium are off by some 10 to 20% and appear to be on the high side, which means that better multiplication measurements and numerical methods are needed. The n,2n reactions result in two helium atoms, which cause radiation damage in the form of hardening at low temperatures (300/degree/C). The usual way beryllium parts are made is by hot pressing the powder. A lower cost method is to cold press and then sinter. There is no radiation damage data on this form of beryllium. The issues of corrosion, safety relative to the release of the tritium built-up inside beryllium, and recycle of used beryllium are also discussed. 10 figs

  18. Fission power flattening in hybrid blankets using mixed fuel

    International Nuclear Information System (INIS)

    In a source-driven fissionable blanket, a flat fission power density (FPD) is achieved by using a mixed fuel (ThO2 and natural UO2) with the thoriumuranium ratio changing from front to back in the ten fuel rows along the radial direction. A straightforward graphic method is used. The temporal behavior of the FPD has been observed for an operation period of 6 months and for a plant load factor of 75% by applying a fusion driver neutron flux of 1014 14-MeV neutrons(cm2 . s) at the first wall, corresponding to --2.25 MWm2. To keep the power density flat, it is necessary to replace the fuel in rows 1, 2, and 3, close to the first wall. The time intervals for this operation increase, counting from initial start-up, typically, 2 months, 6 months, etc. One result of this study is that plutonium produced in such a hybrid blanket contains very low amounts of even isotopic components even over very long operation times of --3 yr. Hence, if fusion reactors are introduced into the energy market, special regulations are needed for international safeguarding

  19. Damage Identification of Continuum Structures Using Blanketing Effect

    International Nuclear Information System (INIS)

    A damage identification method for the continuum structures with making use of the blanketing effect is proposed in this paper. The conventional damage indicator methods such as flexibility matrix method are difficult to deal with the damage identification of complex continuum structure such as multi-span beam. The presented method takes the change rates according to certain diagonal elements of the damaged structure flexibility matrix as identification indicator function. The 'blanketing effect' of the indicator function with respect to the damage factors makes it easy to identify the damage of multi-span beam and plate structures. Some useful formulas are derived and the sensitivity matrix of the identification indicator function is studied. The numerical simulation is performed and the identification results of several damage status are compared. Both beam and plate structures are used for numerical simulation. The results show that the presented indicator function method is very brevity and effective. It is especially suitable for the damage identification of the multi-span beam and plate structures such as bridges and floor

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

    Science.gov (United States)

    Zikanov, Oleg; Liu, Li

    2015-11-01

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

  1. Annual report of the CTR Blanket Engineering research facility in 1994

    International Nuclear Information System (INIS)

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor(CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1994. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  2. Conceptual design of an electricity generating tritium breeding blanket sector for INTOR/NET

    International Nuclear Information System (INIS)

    A study is made of a fusion reactor power blanket and its associated equipment with the objective of producing a conceptual design for a blanket sector of INTOR, or one of its national variants (e.g. NET), from which electricity could be generated simultaneously with the breeding of tritium. (author)

  3. 77 FR 76013 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-12-26

    ... LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported Liquefied.... 2885, which granted Sempra LNG Marketing authority to export a cumulative total of ] 250 Bcf of... Marketing requests blanket authorization to export LNG from the Cameron Terminal that has been...

  4. 75 FR 19954 - Cheniere Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-04-16

    ... Cheniere Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY... Cheniere Marketing, LLC (CMI), requesting blanket authorization to export liquefied natural gas (LNG) that... exported from the Sabine Pass LNG terminal owned by CMI's affiliate, Sabine Pass LNG, L.P., in...

  5. 78 FR 4400 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2013-01-22

    ... USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported Liquefied... and order (Order No. 2923) that granted Eni USA Gas Marketing authority to export a cumulative total... Application, Eni USA Gas Marketing requests blanket authorization to export LNG from the Cameron Terminal...

  6. Wash resistance and repellent properties of Africa University mosquito blankets against mosquitoes

    Directory of Open Access Journals (Sweden)

    N. Lukwa

    2013-04-01

    Full Text Available The effect of permethrin-treated Africa University (AU mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes and repellence (ability to prevent ≥80% of mosquito bites properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up to 20 washes, declining to 90% after 25 washes. Untreated AU blankets did not cause any mortality on mosquitoes. However, mosquito repellence was 96%, 94%, 97.9%, 87%, 85% and 80.7% for treated AU blankets washed 0, 5, 10, 15, 20 and 25 times, respectively. Mosquito repellence was consistently above 80% from 0-25 washes. In conclusion, AU blankets washed 25 times were effective in repelling and killing An. gambiae sl mosquitoes under laboratory conditions.

  7. Resonance self-shielding in the blanket of a hybrid reactor

    International Nuclear Information System (INIS)

    Three sets of energy group cross sections were obtained using various approximations for resonance self shielding. The three models used in obtaining the cross sections were: (a) infinitely dilute model, (b) homogeneous-medium resonance self shielding, and (c) heterogeneous-medium resonance self shielding. The effects on the blanket performance of fusion--fission hybrid reactors, and in particular, on the performance of the current reference Westinghouse Demonstration Tokamak Hybrid Reactor blanket, were compared and analyzed for a variety of fuel-coolant combinations. It has been concluded that (1) the infinitely dilute cross sections can be used to produce preliminary crude estimates for beginning-of-life (BOL) only, (2) the resonance absorber finite dilution should be considered for BOL, poorly moderated blankets and well moderated blankets with low fissile material content situations, and (3) the spacial details should be considered in high fissile content, well moderated blanket situations

  8. Blanket design and performance for the LOTUS fusion-fission hybrid test facility

    International Nuclear Information System (INIS)

    This report summarizes the results of studies performed during 1982 to design an optimized blanket for the initial series of experiments to be conducted in the LOTUS test facility at the Swiss Federal Institute of Technology in Lausanne (EPFL). The experiments are expected to begin in early 1984. An Overview of different hybrid blanket design concepts proposed to date is first given. The technological and economic implications of the different blanket design philosophies are discussed to provide the basis and rationale for the thorium fast-fission blanket design concept selected for the first series of experiments. Detailed description, dimensions, and characteristics of the selected blanket design are given. The neutronic optimization studies on which the design is based are described in detail. Instrumentation and measurement techniques to be used in LOTUS are described elsewhere

  9. An aqueous lithium salt self-cooled blanket and shield for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Sawan, M.E.; Sviatoslavsky, I.N.; Kulcinski, G.L.

    1989-03-01

    A low cost low risk tritium breeding concept has been developed for ITER. This concept is based on dissolving lithium compounds in the water coolant. This makes it possible to breed tritium in all the shield zones, and results in tritium self-efficiency as well as enhanced magnet protection. The design that maximizes the outboard tritium breeding ratio utilizes a 40 cm thick zone of Be balls followed by an 80 cm thick zone of steel balls. Single size balls are used to minimize the pressure drop. The overall TBR excluding tritium bred in the test modules is 1.1. The inlet coolant temperature is 40/sup 0/C and the temperature rise is 15/sup 0/C in the first wall and 35/sup 0/C in the blanket.

  10. Microbiological sampling of spacecraft cabling, antennas, solar panels and thermal blankets

    Science.gov (United States)

    Koukol, R. C.

    1973-01-01

    Sampling procedures and techniques described resulted from various flight project microbiological monitoring programs of unmanned planetary spacecraft. Concurrent with development of these procedures, compatibility evaluations were effected with the cognizant spacecraft subsystem engineers to assure that degradation factors would not be induced during the monitoring program. Of significance were those areas of the spacecraft configuration for which special handling precautions and/or nonstandard sample gathering techniques were evolved. These spacecraft component areas were: cabling, high gain antenna, solar panels, and thermal blankets. The compilation of these techniques provides a historical reference for both the qualification and quantification of sampling parameters as applied to the Mariner Spacecraft of the late 1960's and early 1970's.

  11. Reaction rates in blanket assemblies of a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    To validate neutronics calculation for the blanket design of fusion-fission hybrid reactor, experiments for measuring reaction rates inside two simulating assemblies are performed. Two benchmark assemblies were developed for the neutronics experiments. A D-T fusion neutron source is placed at the center of the setup. One of them consists of three layers of depleted uranium shells and two layers of polyethylene shells, and these shells are arranged alternatively. The 238U capture reaction rates are measured using depleted uranium foils and an HPGe gamma spectrometer. The fission reaction rates are measured using a fission chamber coated with depleted uranium. The other assembly consists of depleted uranium and LiH shells. The tritium production rates are measured using the lithium glass scintillation detector which is placed in the LiH region of the assembly. The measured reaction rates are compared with the calculated ones predicted using MCNP code, and C/E values are obtained. (authors)

  12. A Markov blanket-based method for detecting causal SNPs in GWAS

    Directory of Open Access Journals (Sweden)

    Han Bing

    2010-04-01

    Full Text Available Abstract Background Detecting epistatic interactions associated with complex and common diseases can help to improve prevention, diagnosis and treatment of these diseases. With the development of genome-wide association studies (GWAS, designing powerful and robust computational method for identifying epistatic interactions associated with common diseases becomes a great challenge to bioinformatics society, because the study of epistatic interactions often deals with the large size of the genotyped data and the huge amount of combinations of all the possible genetic factors. Most existing computational detection methods are based on the classification capacity of SNP sets, which may fail to identify SNP sets that are strongly associated with the diseases and introduce a lot of false positives. In addition, most methods are not suitable for genome-wide scale studies due to their computational complexity. Results We propose a new Markov Blanket-based method, DASSO-MB (Detection of ASSOciations using Markov Blanket to detect epistatic interactions in case-control GWAS. Markov blanket of a target variable T can completely shield T from all other variables. Thus, we can guarantee that the SNP set detected by DASSO-MB has a strong association with diseases and contains fewest false positives. Furthermore, DASSO-MB uses a heuristic search strategy by calculating the association between variables to avoid the time-consuming training process as in other machine-learning methods. We apply our algorithm to simulated datasets and a real case-control dataset. We compare DASSO-MB to other commonly-used methods and show that our method significantly outperforms other methods and is capable of finding SNPs strongly associated with diseases. Conclusions Our study shows that DASSO-MB can identify a minimal set of causal SNPs associated with diseases, which contains less false positives compared to other existing methods. Given the huge size of genomic dataset

  13. Preliminary thermo-mechanical analysis of ITER breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

    Thermo-mechanical analysis has been conducted on ITER breeding blanket taking into account thermo-mechanical characteristics peculiar to pebble beds. The features of the analysis are to adopt an elasto-plastic constitutive model for pebble beds and to take into account spatially varying thermal conductivity and heat transfer coefficient, especially in the Be pebble bed, depending on the stress. ABAQUS code and COUPLED TEMPERATURE-DISPLACEMENT procedure of the code are selected so that thermal conductivity is automatically calculated in each calculation point depending on the stress. The modified DRUCKER-PRAGER/Cap plasticity model for granular materials of the code is selected so as to deal with such mechanical features of pebble bed as shear failure flow and hydrostatic plastic compression, and capability of the model is studied. The thermal property-stress correlation used in the analysis is obtained based on the experimental results at FZK and the results of additional thermo-mechanical analysis performed here. The thermo-mechanical analysis of an ITER breeding blanket module has been performed for four conditions: case A; nominal case with spatial distribution of thermal conductivity and heat transfer coefficient in Be pebble bed depending on the stress, case B; constant thermal conductivity, case C; thermal conductivity = -20% of nominal case, and case D; thermal conductivity = +20% of nominal case. In the nominal case the temperature of breeding material (Li{sub 2}ZrO{sub 3}) ranges from 317degC to 554degC and the maximum temperature of Be pebble bed is 446degC. It is concluded that the temperature distribution is within the current design limits. Though the analyses performed here are preliminary, the results exhibit well the qualitative features of the pebble bed mechanical behaviors observed in experiments. For more detail quantitative estimates of the blanket performance, further investigation on mechanical properties of pebble beds by experiment

  14. Three-dimensional SPH analysis of impact responses of ceramic/metal composite armors%陶瓷/金属复合装甲冲击响应的三维SPH法分析

    Institute of Scientific and Technical Information of China (English)

    张伟; 胡德安; 韩旭; 谭柱华

    2011-01-01

    采用自编三维SPH程序对弹体侵彻陶瓷/金属复合装甲问题进行了数值模拟.为了描述金属和陶瓷材料非线性变形及损伤特性,在程序中嵌入Johnson-Cook和Johnson-Holmquist Ⅱ本构模型及Mie-Gruneisen状态方程.SPH计算得到的典型位移随时间变化曲线与实验结果吻合较好.通过数值模拟合理再现了弹体侵蚀、陶瓷锥形成、飞溅和背板隆起撕裂等物理过程.研究了倾角对复合装甲抗侵彻性能的影响,结果表明,存在一个最优的靶板倾角,使得陶瓷/金属复合装甲的抗侵彻性能最佳.%A three-dimensional smoothed particle hydrodynamics (SPH) code was developed to simulate the penetration of ceramic/metal composite armors by projectiles. The Johnson-Cook and John-son-Holmquist Ⅱ constitutive models and the Mie-Griineisen equation of state were incorporated into the developed code to describe the nonlinear deformation and damage features of metal and ceramic materials. The typical displacement-time curves computed by the developed SPH code are in good a-greement with the experimental results. The numerical simulations can represent the major features of the composite armor system such as, projectile erosion, ceramic conoid formation and splash, backing plate failure and so on. And the obliquity effects on the penetration-resistance ability of the composite armor were investigated. The results indicate that there exists an optimal obliquity for the penetration-resistance ability.

  15. Vacuum hot-pressed beryllium and TiC dispersion strengthened tungsten alloy developments for ITER and future fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang, E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Chen, Jiming; Lian, Youyun; Wu, Jihong; Xu, Zengyu; Zhang, Nianman; Wang, Quanming; Duan, Xuro [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Wang, Zhanhong; Zhong, Jinming [Northwest Rare Metal Material Research Institute, CNMC, Ningxia Orient Group Co. Ltd.,No.119 Yejin Road, Shizuishan City, Ningxia,753000 (China)

    2013-11-15

    Beryllium and tungsten have been selected as the plasma facing materials of the ITER first wall (FW) and divertor chamber, respectively. China, as a participant in ITER, will share the manufacturing tasks of ITER first-wall mockups with the European Union and Russia. Therefore ITER-grade beryllium has been developed in China and a kind of vacuum hot-pressed (VHP) beryllium, CN-G01, was characterized for both physical, and thermo-mechanical properties and high heat flux performance, which indicated an equivalent performance to U.S. grade S-65C beryllium, a reference grade beryllium of ITER. Consequently CN-G01 beryllium has been accepted as the armor material of ITER-FW blankets. In addition, a modification of tungsten by TiC dispersion strengthening was investigated and a W–TiC alloy with TiC content of 0.1 wt.% has been developed. Both surface hardness and recrystallization measurements indicate its re-crystallization temperature approximately at 1773 K. Deuterium retention and thermal desorption behaviors of pure tungsten and the TiC alloy were also measured by deuterium ion irradiation of 1.7 keV energy to the fluence of 0.5–5 × 10{sup 18} D/cm{sup 2}; a main desorption peak at around 573 K was found and no significant difference was observed between pure tungsten and the tungsten alloy. Further characterization of the tungsten alloy is in progress.

  16. Neutral gas blanket effects in a gaseous divertor

    International Nuclear Information System (INIS)

    The gaseous divertor employs a neutral gas blanket to absorb the plasma heat flux in the divertor chamber. This novel method for resolving the heat loading problem in a conventional divertor system is simulated experimentally. In our operational range (nsub(e) 13 cm-3, Tsub(e) <= 5 eV) it is demonstrated that the localized plasma heat flux is scattered relatively uniformly with neutral pressures of a few microns. At large neutral pressures the plasma stream is neutralized without touching a material wall. Plasma pumping inhibits neutral backflow and can sustain a neutral pressure difference comparable to the plasma pressure. Effective divertor channel conductance is measured to be reduced by a factor of six. (orig.)

  17. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

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

    Science.gov (United States)

    Bosia, G.; Ragona, R.

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-10

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

  20. Cosmetic wastewater treatment by upflow anaerobic sludge blanket reactor

    International Nuclear Information System (INIS)

    Anaerobic treatment of pre-settled cosmetic wastewater in batch and continuous experiments has been investigated. Biodegradability tests showed high COD and solid removal efficiencies (about 70%), being the hydrolysis of solids the limiting step of the process. Continuous treatment was carried out in an upflow anaerobic sludge blanket reactor. High COD and TSS removal efficiencies (up to 95% and 85%, respectively) were achieved over a wide range of organic load rate (from 1.8 to 9.2 g TCOD L-1 day-1). Methanogenesis inhibition was observed in batch assays, which can be predicted by means of a Haldane-based inhibition model. Both COD and solid removal were modelled by Monod and pseudo-first order models, respectively.

  1. Cosmetic wastewater treatment by upflow anaerobic sludge blanket reactor

    Energy Technology Data Exchange (ETDEWEB)

    Puyol, D.; Monsalvo, V.M.; Mohedano, A.F. [Seccion de Ingenieria Quimica, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain); Sanz, J.L. [Departamento de Biologia Molecular, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain); Rodriguez, J.J., E-mail: juanjo.rodriguez@uam.es [Seccion de Ingenieria Quimica, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain)

    2011-01-30

    Anaerobic treatment of pre-settled cosmetic wastewater in batch and continuous experiments has been investigated. Biodegradability tests showed high COD and solid removal efficiencies (about 70%), being the hydrolysis of solids the limiting step of the process. Continuous treatment was carried out in an upflow anaerobic sludge blanket reactor. High COD and TSS removal efficiencies (up to 95% and 85%, respectively) were achieved over a wide range of organic load rate (from 1.8 to 9.2 g TCOD L{sup -1} day{sup -1}). Methanogenesis inhibition was observed in batch assays, which can be predicted by means of a Haldane-based inhibition model. Both COD and solid removal were modelled by Monod and pseudo-first order models, respectively.

  2. JAERI/U.S. collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Phase IIa and IIb experiments of JAERI/U.S. Collaborative Program on Fusion Blanket Neutronics have been performed using the FNS facility at JAERI. The phase IIa experimental systems consist of the Li2O test region, the rotating neutron target and the Li2CO3 container. In phase IIb, a beryllium layer is added to the inner wall to investigate a multiplier effect. Measured parameters are source characteristics by a foil activation method and spectrum measurements using both NE-213 and proton recoil counters. The measurements inside the Li2O region included tritium production rates, reaction rate by foil activation and neutron spectrum measurements. Analysis for these parameters was performed by using two dimensional discrete ordinate codes DOT3.5 and DOT-DD, and a Monte Carlo code MORSE-DD. The nuclear data used were based on JENDL3/PR1 and PR2. ENDF/B-IV, V and the FNS file were used as activation cross sections. The configurations analysed for the test region were a reference, a beryllium front and a beryllium sandwiched systems in phase IIa, and a reference and a beryllium front with first wall systems in phase IIb. This document describes the results of analysis and comparison between the calculations and the measurements. The prediction accuracy of key parameters in a fusion reactor blanket are examined. The tritium production rates can be well predicted in the reference systems but are fairly underestimated in the system with a beryllium multiplier. Details of experiments and the experimental techniques are described separately in the another report. (author)

  3. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6Li in order to reduce parasite neutron captures in there. (orig./HP)

  4. The LA-0 project of experimental investigations of ADTT blankets with fluoride salts

    International Nuclear Information System (INIS)

    There was an idea of experimental investigations of accelerator-driven transmuter blanket cores with fluoride salts on the experimental reactor LR-0 in the Nuclear Research Institute Rez plc, the Czech Republic, which has been further developed and elaborated. The results of a preliminary proposal and assessment of possibility of a reconstruction of the LR-0 core for experiments simulating blankets with molten fluoride salt fuel were reported. There is a brief survey of essential conclusions of both those stages presented in this paper. The current status of the complex project called LA-0 which has been proposed to start in 1996 will be given in some details. The project consists of all the basic regions -neutron source and subcritical assembly design and other engineering problems including fluoride chemical technology - of such a system. Nevertheless, the main attention, in this paper, is focused on the methodology of all the regions, i.e. methods of neutronic characteristic predictions, both computational and experimental, as well as some methods selected in the other topical regions are mentioned. Among the methods being taken into account for the first phases of the experimental program the measurements of neutronic characteristics by use of oscillating samples method will be given at the first position. There is also an illustrative description of the existing LR-0 experimental reactor given in the paper. Finally, there are future stages of the project outlined in a rough frame only and their benefits as well as drawbacks critically discussed in the concluding part of the paper. 5 refs., 4 figs

  5. Biological nutrient removal by internal circulation upflow sludge blanket reactor after landfill leachate pretreatment.

    Science.gov (United States)

    Abood, Alkhafaji R; Bao, Jianguo; Abudi, Zaidun N

    2013-10-01

    The removal of biological nutrient from mature landfill leachate with a high nitrogen load by an internal circulation upflow sludge blanket (ICUSB) reactor was studied. The reactor is a set of anaerobic-anoxic-aerobic (A2/O) bioreactors, developed on the basis of an expended granular sludge blanket (EGSB), granular sequencing batch reactor (GSBR) and intermittent cycle extended aeration system (ICEAS). Leachate was subjected to stripping by agitation process and poly ferric sulfate coagulation as a pretreatment process, in order to reduce both ammonia toxicity to microorganisms and the organic contents. The reactor was operated under three different operating systems, consisting of recycling sludge with air (A2/O), recycling sludge without air (low oxygen) and a combination of both (A2/O and low oxygen). The lowest effluent nutrient levels were realised by the combined system of A2/O and low oxygen, which resulted in effluent of chemical oxygen demand (COD), NH3-N and biological oxygen demand (BOD5) concentrations of 98.20, 13.50 and 22.50 mg/L. The optimal operating conditions for the efficient removal of biological nutrient using the ICUSB reactor were examined to evaluate the influence of the parameters on its performance. The results showed that average removal efficiencies of COD and NH3-N of 96.49% and 99.39%, respectively were achieved under the condition of a hydraulic retention time of 12 hr, including 4 hr of pumping air into the reactor, with dissolved oxygen at an rate of 4 mg/L and an upflow velocity 2 m/hr. These combined processes were successfully employed and effectively decreased pollutant loading. PMID:24494501

  6. The complete mitochondrial genome of the armored catfish, Hypostomus plecostomus (Siluriformes: Loricariidae).

    Science.gov (United States)

    Liu, Shikai; Zhang, Jiaren; Yao, Jun; Liu, Zhanjiang

    2016-05-01

    The complete mitochondrial genome of the armored catfish, Hypostomus plecostomus, was determined by next generation sequencing of genomic DNA without prior sample processing or primer design. Bioinformatics analysis resulted in the entire mitochondrial genome sequence with length of 16,523 bp. The H. plecostomus mitochondrial genome is consisted of 13 protein-coding genes, 22 tRNA genes, 2 rRNA genes, and 1 control region, showing typical circular molecule structure of mitochondrial genome as in other vertebrates. The whole genome base composition was estimated to be 31.8% A, 27.0% T, 14.6% G, and 26.6% C, with A/T bias of 58.8%. This work provided the H. plecostomus mitochondrial genome sequence which should be valuable for species identification, phylogenetic analysis and conservation genetics studies in catfishes. PMID:25329264

  7. The complete mitochondrial genome of the armored catfish, Hypostomus plecostomus (Siluriformes: Loricariidae).

    Science.gov (United States)

    Liu, Shikai; Zhang, Jiaren; Yao, Jun; Liu, Zhanjiang

    2016-05-01

    The complete mitochondrial genome of the armored catfish, Hypostomus plecostomus, was determined by next generation sequencing of genomic DNA without prior sample processing or primer design. Bioinformatics analysis resulted in the entire mitochondrial genome sequence with length of 16,523 bp. The H. plecostomus mitochondrial genome is consisted of 13 protein-coding genes, 22 tRNA genes, 2 rRNA genes, and 1 control region, showing typical circular molecule structure of mitochondrial genome as in other vertebrates. The whole genome base composition was estimated to be 31.8% A, 27.0% T, 14.6% G, and 26.6% C, with A/T bias of 58.8%. This work provided the H. plecostomus mitochondrial genome sequence which should be valuable for species identification, phylogenetic analysis and conservation genetics studies in catfishes.

  8. Armored brain in patients with hydrocephalus after shunt surgery: review of the literatures.

    Science.gov (United States)

    Taha, Mahmoud M

    2012-01-01

    Armored brain or chronic calcified subdural hematoma is a rare complication of cerebrospinal fluid diversion with few cases reported in the literature. Seventeen patients with this pathology have been published. A complete review of the literatures regarding this topic has been collected and discussed. The author also presents a 12- year old boy with triventricular hydrocephalus who had undergone ventriculoperitoneal medium pressure shunt system since birth. The patient presented to our clinic with a 2-year history of seizures. The patient was conscious and without neurological deficits on examination. Computed tomography of the brain showed bilateral high density mass with surface calcification. X ray skull and MRI confirmed the calcified subdural hematoma bilaterally. We preferred conservative treatment and the patient continued his antiepileptic treatment. At one year follow up, the patient had the same neurological state. The case highlights the importance of frequent follow up CT brain after shunt surgery.

  9. Armor plate protection for the Doublet III vacuum vessel for neutral beam heating

    International Nuclear Information System (INIS)

    The design of vacuum vessel armor plate for neutral beam systems presents a number of challenges to the engineer. Heat fluxes of several hundred watts/cm2 must be handled on a routine basis during normal plasma operations, and a factor of ten increase in these fluxes can occur during plasma disruptions. At the present time, a graphite tile system appears to be the best candidate for such a situation. Heat fluxes in excess of 4 kW/cm2 can be routinely sustained and the material sputtered or evaporated from the surface has a low atomic number. The system proposed for Doublet III will provide valuable data for the designers of future fusion reactors and will also provide proof-of-principle demonstrations for such machines as TFTR and JET

  10. Eddy current induced electromagnetic loads on shield blankets during plasma disruptions in ITER: A benchmark exercise

    International Nuclear Information System (INIS)

    According to recent updates of ITER shield blanket design, electromagnetic loads during the plasma disruption are being evaluated to verify the mechanical confidence and reliability. As a course of such evaluations, a benchmark activity for the electromagnetic analysis, coordinated by ITER Organization, is underway between ITER parties to compare the calculation results for disruption loads on the blankets. In this paper, we present calculation results for the electromagnetic loads on the simplified but practical model of ITER shield blankets with respect to six representative disruption scenarios of which ITER distributes simulation results based on the DINA code as a reference of the design and analysis. Commercial finite element method software, ANSYS/EmagTM, was employed to evaluate the eddy current on the blanket modules with the 40o sector model for major conducting structure of the tokamak including double-walled vacuum vessel, triangular support, and vertical targets of divertors. An interface between ANSYS/EmagTM and plasma simulator was implemented with a conversion tool assigning the plasma current density on the ANSYS elements corresponding to the current filaments in DINA outputs. Discussions are made of the possible improvement of the blanket model taking more realistic blanket configuration into account at the cost of the moderate increase in computational time. A final remark is given of the possibility of incorporating halo currents into ANSYS disruption simulations, which are major sources of electromagnetic loads on in-vessel components including blankets.

  11. 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.

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

    International Nuclear Information System (INIS)

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

  13. R and D activities of the liquid breeder blanket in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won, E-mail: dwlee@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Eo Hwak; Kim, Suk Kwon; Yoon, Jae Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer MARS and GAMMA were developed for He coolant and liquid breeder analysis. Black-Right-Pointing-Pointer FMS/FMS and Be/FMS joining methods were developed and verified with high heat flux test. Black-Right-Pointing-Pointer High temperature and pressure nitrogen and He loops were constructed for heat transfer experiment for developed codes validation. Black-Right-Pointing-Pointer A PbLi breeder loop was constructed for components, MHD, and corrosion tests. Black-Right-Pointing-Pointer A chamber for tritium extraction with a gas-liquid contact method was constructed. - Abstract: A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5-1.0 MW/m{sup 2}. Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and

  14. Preliminary Analysis of Liquid Metal MHD Pressure Drop in the Blanket for the FDS

    Institute of Scientific and Technical Information of China (English)

    王红艳; 吴宜灿; 何晓雄

    2002-01-01

    Preliminary analysis and calculation of liquid metal Li17Pb83 magnetohydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket. To decrease the liquid metal MHD pressure drop, Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts. The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed. Finally, the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.

  15. First adaptation of the European ceramic B. I. T. blanket design to the updated DEMO specifications

    Energy Technology Data Exchange (ETDEWEB)

    Anzidei, L.; Cecchi, P.; Cevolani, S.; Gallina, M.; Petrizzi, L.; Rado, V.; Talarico, C.; Violante, V.; Vettraino, V.; Zampaglione, V. (Associazione Euratom-ENEA sulla Fusione, Frascati (Italy)); Proust, E.; Giancarli, L.; Raepsaet, X.; Szczepanski, J.; Vallette, F.; Baraer, L.; Bielak, B.; Mercier, J. (Commissariat a l' Energie Atomique, DRN/DMT/SERMA, C.E.N. Saclay, 91 - Gif-sur-Yvette (France))

    1991-12-01

    The DEMO specifications defined so as to ensure the consistency of the various blanket conceptual design studies performed within the framework of the European Test Blanket Programme have been recently updated. A very first attempt has been made to adapt the European Ceramic Breeder Inside-Tube DEMO blanket to these new specifications. Two solutions have been investigated. The first would ensure tritium self-sufficiency of the plant with a large safety margin. The other one, which fully preserves the design simplicity and reliability of the initial design, appears to be somewhat marginal from the tritium breeding capability point of view, but to offer good improvement prospects. (orig.).

  16. Beryllium data base for in-pile mockup test on blanket of fusion reactor, (1)

    International Nuclear Information System (INIS)

    Beryllium has been used in the fusion blanket designs with ceramic breeder as a neutron multiplier to increase the net tritium breeding ratio (TBR). The properties of beryllium, that is physical properties, chemical properties, thermal properties, mechanical properties, nuclear properties, radiation effects, etc. are necessary for the fusion blanket design. However, the properties of beryllium have not been arranged for the fusion blanket design. Therefore, it is indispensable to check and examine the material data of beryllium reported previously. This paper is the first one of the series of papers on beryllium data base, which summarizes the reported material data of beryllium. (author)

  17. Realization of a flat fission power density in a hybrid blanket over long operation periods

    International Nuclear Information System (INIS)

    A straightforward numerical graphical method is applied to achieve a flat fission power density (FPD) in a hybrid blanket by using a mixed fuel (ThO2 and natural UO2) with variable fractions of the fuel components in the radial direction. The neutronic analysis is carried out on a blanket with a hard neutron spectrum in the fissionable zone by simply omitting the moderating beryllium neutron multiplier. Mainly due to this precaution in the blanket design, the FPD could be kept quasi-constant over a relatively long plant lifetime

  18. Evaluation of Cortaderia selloana (Capim-dos-pampas) blankets as sorbent materials for oil spills in simulated hydro equipment; Estudo do desempenho de tecidos e mantas para utilizacao como sorventes para petroleo

    Energy Technology Data Exchange (ETDEWEB)

    Bonetti, T.F.; Sydenstricker, T.H.D. [Universidade Federal do Parana (UFPR), Curitiba, PR (Brazil)], e-mail: thais@demec.ufpr.br; Amico, S.C. [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil)

    2006-07-01

    Oil spills in aquatic environments may cause serious economy losses and severe environmental impact which both drive the development of commercial systems (e.g. sorbents) to control these accidents. One way of using sorbents is to encapsulate them with an involucre or cover, i.e. producing blankets. The focus of this research is to evaluate the key characteristics of interest (aerial density, water and oil sorption, mechanical strength and cost) of different materials to use as covers for blankets and to prepare blankets and compare their performance when made with various core materials, such as Cortaderia selloana fibers and different commercial sorbents. A simulated aqueous body with stream was used for the sorption experiments, where the oil and water phases were circulated and forced to pass under the blankets. On the sorption tests, the fibers of Cortaderia selloana reached a performance lower to that of commercial sorbents, mainly due to their low density and high volume (difficult packing), nevertheless a clear trend was noted, heavier blankets with higher sorption periods lead to higher sorption. (author)

  19. The influence of external source intensity in accelerator/target/blanket system on conversion ratio and fuel cycle

    Science.gov (United States)

    Kochurov, Boris P.

    1995-09-01

    The analysis of neutron balance relation for a subcritical system with external source shows that a high ratio of neutron utilization (conversion ratio, breeding ratio) much exceeding similar values for nuclear reactors (both thermal or fast spectrum) is reachable in accelerator/target/blanket system with high external neutron source intensity. An accelerator/target/blanket systems with thermal power in blanket about 1850 Mwt and operating during 30 years have been investigated. Continual feed up by plutonium (fissile material) and Tc-99 (transmuted material) was assumed. Accelerator beam intensity differed 6.3 times (16 mA-Case 1, and 100 mA-Case 2). Conversion ratio (CR) was defined as the ratio of Tc-99 nuclei transmuted to the number of Pu nuclei consumed. The results for two cases are as follows: Case 1Case 2CR 0.77 1.66N(LWR) 8.6 19.1Power MWt(el) 512 225 where N(LWR)-number of LWRs(3000 MWt(th)) from which yearly discharge of Tc-99 is transmuted during 30 years. High value of conversion ratio considerably exceeding 1 (CR=1.66) was obtained in the system with high source intensity as compared with low source system (CR=0.77). Net output of electric power of high source intensity system is about twice lower due to consumption of electric power for accelerator feed up. The loss of energy for Tc-99 transmutation is estimated as 40 Mev(el)/nuclei. Yet high conversion ratio (or breeding ratio) achievable in electronuclear installations with high intensity of external source can effectively be used to close fuel cycle (including incineration of wastes) or to develop growing nuclear power production system.

  20. The gas-cooled Li2O moderator/breeder canister blanket for fusion-synfuels

    International Nuclear Information System (INIS)

    A new integrated power and breeding blanket is described. The blanket incorporates features that make it suitable for synthetic fuel production. It is matched to the thermal and electrical requirements of the General Atomic water-splitting process for producing hydrogen. The fusion reaction is the Tandem Mirror Reactor (TMR) using Mirror Advanced Reactor Study (MARS) physics. The canister blanket is a high temperature, pressure balanced, crossflow heat exchanger contained within a low activity, independently cooled, moderate temperature, first wall structural envelope. The canister uses Li2O as the moderator/breeder and helium as the coolant. ''In situ'' tritium control, combined with slip stream processing and self-healing permeation barriers, assures a hydrogen product essentially free of tritium. The blanket is particularly adapted to synfuels production but is equally useful for electricity production or co-generation

  1. Fast Breeder Blanket Facility FBBF. Annual report, January 1, 1981-December 31, 1981

    International Nuclear Information System (INIS)

    This annual report contains a summmary of fission rate, spectra, and gamma-ray heating rate measurements made in the first blanket of the Purdue Fast Breeder Blanket Facility. The first blanket consisted of aluminum clad, natural UO2 fuel rods with a secondary cladding of stainless steel or aluminum. The blanket was arranged in two concentric regions around the neutron source and converter regions. A neutron diffusion code, 2DB, and a Monte Carlo code, VIM, both using homogeneous cross section groups have been used to calculate the reaction rates. Calculated to experimental values for a number of important reactions are presented. A modified method of applying Bondarenko self-shielding factors to correct for the self shielding of resonance energy neutrons in aluminum, stainless steel and UO2 has improved the agreement between the calculations and experiment, but does not account for all of the differences

  2. Integrated-blanket-coil (IBC) concept applied to the OH-coil for spherical tori

    International Nuclear Information System (INIS)

    This concept combines blanket and coil functions into a single component. The objectives of the concept are to: (1) provide design options, (2) simplify overall configuration, (3) enhance compactness, and (4) reduce costs. Some drawings of the system are given

  3. Main maintenance operations for Test Blanket Systems in ITER TBM port cells

    International Nuclear Information System (INIS)

    Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues

  4. Acoustic contributions of a sound absorbing blanket placed in a double panel structure: Absorption Versus Transmission

    CERN Document Server

    Doutres, Olivier; 10.1121/1.3458845

    2010-01-01

    The objective of this paper is to propose a simple tool to estimate the absorption vs. transmission loss contributions of a multilayered blanket unbounded in a double panel structure and thus guide its optimization. The normal incidence airborne sound transmission loss of the double panel structure, without structure-borne connections, is written in terms of three main contributions; (i) sound transmission loss of the panels, (ii) sound transmission loss of the blanket and (iii) sound absorption due to multiple reflections inside the cavity. The method is applied to four different blankets frequently used in automotive and aeronautic applications: a non-symmetric multilayer made of a screen in sandwich between two porous layers and three symmetric porous layers having different pore geometries. It is shown that the absorption behavior of the blanket controls the acoustic behavior of the treatment at low and medium frequencies and its transmission loss at high frequencies. Acoustic treatment having poor sound ...

  5. An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

    Science.gov (United States)

    Yu, Guanying; Liu, Xufeng; Liu, Songlin

    2016-10-01

    The Chinese Fusion Engineering Tokamak Reactor (CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder (WCCB) blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic (RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement. supported by National Natural Science Foundation of China (No. 11175207) and the National Magnetic Confinement Fusion Program of China (No. 2013GB108004)

  6. Normal Operation (NO) of APT Blanket System and its Components Based on Initial Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  7. ARPA TRP turboalternator development: Quarterly Report, July--September 1995. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    This report presents information on 4 major tasks: under-armor auxiliary power unit (UAAPU) demonstration on an M1A1 tank; low-cost turbogenerator design; negligible emission catalytic combustor development; and turbogenerator demonstration on an electric bus.

  8. Climate-driven expansion of blanket bogs in Britain during the Holocene

    Directory of Open Access Journals (Sweden)

    A. V. Gallego-Sala

    2015-10-01

    Full Text Available Blanket bog occupies approximately 6 % of the area of the UK today. The Holocene expansion of this hyperoceanic biome has previously been explained as a consequence of Neolithic forest clearance. However, the present distribution of blanket bog in Great Britain can be predicted accurately with a simple model (PeatStash based on summer temperature and moisture index thresholds, and the same model correctly predicts the highly disjunct distribution of blanket bog worldwide. This finding suggests that climate, rather than land-use history, controls blanket-bog distribution in the UK and everywhere else. We set out to test this hypothesis for blanket bogs in the UK using bioclimate envelope modelling compared with a database of peat initiation age estimates. We used both pollen-based reconstructions and climate model simulations of climate changes between the mid-Holocene (6000 yr BP, 6 ka and modern climate to drive PeatStash and predict areas of blanket bog. We compiled data on the timing of blanket-bog initiation, based on 228 age determinations at sites where peat directly overlies mineral soil. The model predicts large areas of northern Britain would have had blanket bog by 6000 yr BP, and the area suitable for peat growth extended to the south after this time. A similar pattern is shown by the basal peat ages and new blanket bog appeared over a larger area during the late Holocene, the greatest expansion being in Ireland, Wales and southwest England, as the model predicts. The expansion was driven by a summer cooling of about 2 °C, shown by both pollen-based reconstructions and climate models. The data show early Holocene (pre-Neolithic blanket-bog initiation at over half of the sites in the core areas of Scotland, and northern England. The temporal patterns and concurrence of the bioclimate model predictions and initiation data suggest that climate change provides a parsimonious explanation for the early Holocene distribution and later

  9. New progress on design and R and D for solid breeder test blanket module in China

    Energy Technology Data Exchange (ETDEWEB)

    Feng, K.M., E-mail: fengkm@swip.ac.cn; Zhang, G.S.; Hu, G.; Chen, Y.J.; Feng, Y.J.; Li, Z.X.; Wang, P.H.; Zhao, Z.; Ye, X.F.; Xiang, B.; Zhang, L.; Wang, Q.J.; Cao, Q.X.; Zhao, F.C.; Wang, F.; Liu, Y.; Zhang, M.C.

    2014-10-15

    Highlights: • The new progress on design and R and D of Chinese solid breeder TBM are introduced. • The mock-up fabrication and component tests for Chinese HCCB TBM have being developed. • The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CFL-1 are being prepared. • The fabrication of 1/3 sized mock-up is being carried-out. • The key technology development is proceeding to the large-scale mock-up fabrication. - Abstract: ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R and D activities for each TBM module with the auxiliary system are introduced. The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R and D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.

  10. Wash resistance and repellent properties of Africa University mosquito blankets against mosquitoes

    OpenAIRE

    Lukwa, N; A. Makuwaza; T. Chiwade; S.L. Mutambu; M. Zimba; P. Munosiyei

    2013-01-01

    The effect of permethrin-treated Africa University (AU) mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes) and repellence (ability to prevent ≥80% of mosquito bites) properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up t...

  11. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2010-06-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  12. Implementation of two-phase tritium models for helium bubbles in HCLL breeding blanket modules

    OpenAIRE

    Fradera, Jordi; Sedano, L.A.; Mas de les Valls Ortiz, Elisabet; Batet Miracle, Lluís

    2011-01-01

    Tritium self-sufficiency requirement of future DT fusion reactors involves large helium production rates in the breeding blankets; this might impact on the conceptual design of diverse fusion power reactor units, such as Liquid Metal (LM) blankets. Low solubility, long residence-times and high production rates create the conditions for Helium nucleation, which could mean effective T sinks in LM channels. A model for helium nano-bubble formation and tritium conjugate transport phen...

  13. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2007-08-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  14. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Zhongliang; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Chen, Chong; Li, Min; Zhou, Guangming

    2015-06-15

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements.

  15. Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

    Institute of Scientific and Technical Information of China (English)

    HUANG Qun-Ying; LI Jian-Gang; CHEN Yi-Xue

    2004-01-01

    @@ Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB)to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDS-FBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW. yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.

  16. High levels of exotic armored scales on imported avocados raise concerns regarding USDA-APHIS' phytosanitary risk assessment.

    Science.gov (United States)

    Morse, J G; Rugman-Jones, P F; Watson, G W; Robinson, L J; Bi, J L; Stouthamer, R

    2009-06-01

    Between 1914 and 2007, a quarantine protected California avocado, Persea americana Mill., groves from pests that might be introduced into the state along with fresh, imported avocados. Soon after Mexican avocados were first allowed entry on 1 February 2007, live specimens of several species of armored scales (Hemiptera: Diaspididae) not believed to be present in California were detected on 'Hass' avocados entering the state from Mexico. Initially, the California Department of Food and Agriculture (CDFA) prevented avocados infested with these scales from entering the state or required that they be fumigated with an approved treatment such as methyl bromide. After a Science Advisory Panel meeting in May 2007, U.S. Department of Agriculture-Animal and Plant Health Inspection Service (USDA-APHIS) reaffirmed its position that armored scales on shipments of fruit for consumption (including avocados) pose a "low risk" for pest establishment. In compliance with APHIS protocols, as of 18 July 2007, CDFA altered its policy to allow shipments of scale-infested avocados into the state without treatment. Here, we report on sampling Mexican avocados over an 8-mo period, September 2007-April 2008. An estimated 67 million Mexican Hass avocados entered California over this period. Based on samples from 140 trucks containing approximately 15.6% of this volume of fruit, we estimate that approximately 47.6 million live, sessile armored scales and an additional 20.1 million live eggs and crawlers were imported. We found eight probable species of armored scales in the samples, seven of these are not believed to occur in California; 89.3% of the live scales were Abgrallaspis aguacatae Evans, Watson and Miller, a recently described species. In contrast to the USDA-APHIS opinion, we believe the volume of shipments and levels of live scales they contain present a significant risk to California's US$300 million avocado industry and to other crops that might become infested by one or more of

  17. Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

    1999-08-01

    Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

  18. Detailed 3-D nuclear analysis of ITER blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, T.D., E-mail: tdbohm@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Sawan, M.E.; Marriott, E.P.; Wilson, P.P.H. [University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, M.; Bullock, J. [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-10-15

    In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.

  19. Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Humrickhouse, Paul Weston [Idaho National Laboratory; Merrill, Brad Johnson [Idaho National Laboratory

    2014-11-01

    It is envisioned that tritium will be extracted from DCLL blankets using a vacuum permeator. We derive here an analytical solution for the extraction efficiency of a permeator tube, which is a function of only two dimensionless numbers: one that indicates whether radial transport is limited in the PbLi or in the solid membrane, and another that is the ratio of axial and radial transport times in the PbLi. The permeator efficiency is maximized by decreasing the velocity and tube diameter, and increasing the tube length. This is true regardless of the mass transport correlation used; we review several here and find that they differ little, and the choice of correlation is not a source of significant uncertainty here. The PbLi solubility, on the other hand, is a large source of uncertainty, and we identify upper and lower bounds from the literature data. Under the most optimistic assumptions, we find that a ferritic steel permeator operating at 550 °C will need to be at least an order of magnitude larger in volume than previous conceptual designs using niobium and operating at higher temperatures.

  20. Physics aspects of metal fuelled fast reactors with thorium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Mohapatra, D.K., E-mail: dina@igcar.gov.in; Singh, S.S.; Riyas, A.; Mohanakrishnan, P.

    2013-12-15

    Metal fuelled fast breeder reactors (MFBR) with high breeding ratio will play a major role in meeting the high nuclear power growth envisaged in India. In this regard several conceptual reactor designs with alloys of U–Pu–Zr fuel have been suggested for commercial operations. This study focusses on the physics design aspects of a sodium cooled U–Pu–6%Zr fuelled 1000 MWe fast breeder reactor, which can attain a breeding ratio of nearly 1.5. The calculation results on reactor kinetics and safety parameters of the 1000 MWe MFBR are presented. The changes in the breeding ratio by introduction of thorium in the blankets of the MFBR are also investigated. Burnup analyses are carried out to compare the core burnup effects in MOX and metal fuelled FBRs. Since the MOX fuelled 500 MWe prototype fast breeder is getting constructed at IGCAR, for burnup comparisons a MFBR of similar design is considered. The results of this study indicate that the loss of reactivity in the metal core with burnup is less than half that of a MOX core and its breeding ratio remains nearly constant. It is also found that the isotopic composition of plutonium (Pu-vector composition) remains more steady with burnup in a metal core.

  1. Upflow anaerobic sludge blanket reactor--a review.

    Science.gov (United States)

    Bal, A S; Dhagat, N N

    2001-04-01

    . Concentrated waste (usually sewage sludge) can be added continuously or periodically (semi-batch operation), where it is mixed with the contents of the reactor. Theoretically, the conventional digester is operated as a once-through, completely mixed, reactor. In this particular mode of operation the hydraulic retention time (HRT) is equal to the solids retention time (SRT). Basically, the required process efficiency is related to the sludge retention time (SRT), and hence longer SRT provided, results in satisfactory population (by reproduction) for further waste stabilization. By reducing the hydraulic retention time (HRT) in the conventional mode reactor, the quantity of biological solids within the reactor is also decreased as the solids escape with the effluent. The limiting HRT is reached when the bacteria are removed from the reactor faster than they can grow. Methanogenic bacteria are slow growers and are considered the rate-limiting component in the anaerobic digestion process. The first anaerobic process developed, which separated the SRT from the HRT was the anaerobic contact process. In 1963, Young and McCarty (1968) began work, which eventually led to the development of the anaerobic upflow filter (AF) process. The anaerobic filter represented a significant advance in anaerobic waste treatment, since the filter can trap and maintain a high concentration of biological solids. By trapping these solids, long SRT's could be obtained at large waste flows, necessary to anaerobically treat low strength wastes at nominal temperatures economically. Another anaerobic process which relies on the development of biomass on the surfaces of a media is an expanded bed upflow reactor. The primary concept of the process consists of passing wastewater up through a bed of inert sand sized particles at sufficient velocities to fluidize and partially expand the sand bed. One of the more interesting new processes is the upflow anaerobic sludge blanket process (UASB), which was developed

  2. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for test blanket module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation four different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port two TBMs will be accommodated, in the port 16 will be the European helium-cooled pebble bed blanket. In different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for the exchange of TBMs. Two TBMs are mounted onto one common frame, into a port plug (PP), which offers a standardised interface to the vacuum vessel (VV). It is cantilevered with a flange to VV port extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs is also the standard operation of ITER. Several components of the helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system are integrated into the ancillary equipment unit (AEU), which during the operation will connect the port plug to the subsystems. The bigger part of the AEU is accommodated in the port cell and the rest part of it is penetrated into the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection/disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected here. A

  3. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for Test Blanket Module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation 4 different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port will be two TBMs accomodated, in the port 16 will be the the European Helium Cooled Pebble Bed blanket. In the different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for exchange the TBMs. Two TBMs are mounted into one common frame, into a Port Plug (PP), which offers a standardised interface to the Vacuum Vessel (VV). It is cantilevered with a flange to VV Port Extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs are also standard operation of ITER. Several components of the Helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system, etc. All of them is integrated into the Ancillary Equipment Unit (AEU) which during operation will connect the port plug to the sub systems. The bigger part of the AEU is accomodated in the Port Cell and the rest part of it is penetrate to the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection / disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected

  4. Development and tests of molybdenum armored copper components for MITICA ion source

    Energy Technology Data Exchange (ETDEWEB)

    Pavei, Mauro, E-mail: mauro.pavei@igi.cnr.it; Marcuzzi, Diego; Rizzolo, Andrea; Valente, Matteo [Consorzio RFX, Corso Stati Uniti, 4, I-35127 Padova (Italy); Böswirth, Bernd; Greuner, Henri [Max-Planck-Institut für Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany)

    2016-02-15

    In order to prevent detrimental material erosion of components impinged by back-streaming positive D or H ions in the megavolt ITER injector and concept advancement beam source, a solution based on explosion bonding technique has been identified for producing a 1 mm thick molybdenum armour layer on copper substrate, compatible with ITER requirements. Prototypes have been recently manufactured and tested in the high heat flux test facility Garching Large Divertor Sample Test Facility (GLADIS) to check the capability of the molybdenum-copper interface to withstand several thermal shock cycles at high power density. This paper presents both the numerical fluid-dynamic analyses of the prototypes simulating the test conditions in GLADIS as well as the experimental results.

  5. 植被恢复用植生卷材制造技术及其应用%Manufacture Technology of Vegetative Blanket of Natural Fiber for Revegetation and Its Use

    Institute of Scientific and Technical Information of China (English)

    郭文静; 赵平; 王正; 范留芬

    2011-01-01

    综述以植物纤维为主要原料的复合植生卷材在边坡绿化中的应用优势,介绍复合植生卷材的特点以及以无纺布或纸、农作物秸秆、木纤维为主要原料的不同复合植生卷材的制造技术及其特点,概述复合植生卷材的国内外技术发展应用状况,并提出复合植生卷材在植被恢复和边坡绿化等领域的应用前景及其在我国生态环境综合治理中低碳加工、循环利用的新思路。%The application advantages of vegetative blanket made from natural fiber in slope vegetation were summarized. Both the characteristics of compounded vegetative blanket as well as the manufacture technology and characteristics of different types of compounded vegetative blanket made from paper or nonwoven fabrics, crop stalks and wood fiber respectively were described. The technology development and application of compounded vegetative blanket at home and abroad were also reviewed. Finally, the paper prospected the application of compounded vegetative blanket in revegetation and slope vegetation, and proposed the new thought that corapounded vegetative blanket can be used in China's eological environment improvement to achieve low-carbon and recycling use.

  6. Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Mexico, D.F. (Mexico); Francois, Juan Luis [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico)]. E-mail: jlfl@fi-b.unam.mx; Martin-del-Campo, Cecilia [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana, Avenida San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico)

    2005-04-15

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the {sup 233}U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly.

  7. Challenges of Engineering Grain Boundaries in Boron-Based Armor Ceramics

    Science.gov (United States)

    Coleman, Shawn P.; Hernandez-Rivera, Efrain; Behler, Kristopher D.; Synowczynski-Dunn, Jennifer; Tschopp, Mark A.

    2016-06-01

    Boron-based ceramics are appealing for lightweight applications in both vehicle and personnel protection, stemming from their combination of high hardness, high elastic modulus, and low density as compared to other ceramics and metal alloys. However, the performance of these ceramics and ceramic composites is lacking because of their inherent low fracture toughness and reduced strength under high-velocity threats. The objective of the present article is to briefly discuss both the challenges and the state of the art in experimental and computational approaches for engineering grain boundaries in boron-based armor ceramics, focusing mainly on boron carbide (B4C) and boron suboxide (B6O). The experimental challenges involve processing these ceramics at full density while trying to promote microstructure features such as intergranular films to improve toughness during shock. Many of the computational challenges for boron-based ceramics stem from their complex crystal structure which has hitherto complicated the exploration of grain boundaries and interfaces. However, bridging the gaps between experimental and computational studies at multiple scales to engineer grain boundaries in these boron-based ceramics may hold the key to maturing these material systems for lightweight defense applications.

  8. SIRE (sight-integrated ranging equipment): an eyesafe laser rangefinder for armored vehicle fire control systems

    Science.gov (United States)

    Keeter, Howard S.; Gudmundson, Glen A.; Woodall, Milton A., II

    1991-04-01

    The Sight Integrated Ranging Equipment (SIRE) incorporates an eyesafe laser rangefinder into the M-36 periscope used in tactical armored vehicles, such as the Commando Stingray light tank. The SIRE unit provides crucial range data simultaneously to the gunner and fire control computer. This capability greatly reduces 'time-to-fire', improves first-round hit probability, and increases the overall effectiveness of the vehicle under actual and simulated battlefield conditions. The SIRE can provide target range up to 10-km, with an accuracy of 10-meters. The key advantage of the SIRE over similar laser rangefinder systems is that it uses erbium:glass as the active lasing medium. With a nominal output wavelength of 1.54-microns, the SIRE can produce sufficient peak power to penetrate long atmospheric paths (even in the presence of obscurants), while remaining completely eyesafe under all operating conditions. The SIRE is the first eyesafe vehicle-based system to combine this level of accuracy, maximum range capability, and fire control interface. It simultaneously improves the accuracy and confidence of the operator, and eliminates the ocular hazard issues typically encountered with laser rangefinder devices.

  9. Environmental assessment of depleted uranium used in military armor-piercing rounds in terrestrial systems.

    Science.gov (United States)

    Stanley, Jacob K; Coleman, Jessica G; Brasfield, Sandra M; Bednar, Anthony J; Ang, Choo Y

    2014-06-01

    Depleted uranium (DU) from the military testing and use of armor-piercing kinetic energy penetrators has been shown to accumulate in soils; however, little is known about the toxicity of DU geochemical species created through corrosion or weathering. The purpose of the present study was to assess the toxic effects and bioaccumulation potential of field-collected DU oxides to the model terrestrial invertebrates Eisenia fetida (earthworm) and Porcellio scaber (isopod). Earthworm studies were acute (72 h) dermal exposures or 28-d spiked soil exposures that used noncontaminated field-collected soils from the US Army's Yuma and Aberdeen Proving Grounds. Endpoints assessed in earthworm testing included bioaccumulation, growth, reproduction, behavior (soil avoidance), and cellular stress (neutral red uptake in coelomocytes). Isopod testing used spiked food, and endpoints assessed included bioaccumulation, survival, and feeding behavior. Concentration-dependent bioaccumulation of DU in earthworms was observed with a maximum bioaccumulation factor of 0.35; however, no significant reductions in survival or impacts to cellular stress were observed. Reproduction lowest-observed-effect concentrations (LOEC) of 158 mg/kg and 96 mg/kg were observed in Yuma Proving Ground and a Mississippi reference soil (Karnac Ferry), respectively. Earthworm avoidance of contaminated soils was not observed in 48-h soil avoidance studies; however, isopods were shown to avoid food spiked with 12.7% by weight DU oxides through digital tracking studies. PMID:24549573

  10. Hazardous Traffic Event Detection Using Markov Blanket and Sequential Minimal Optimization (MB-SMO).

    Science.gov (United States)

    Yan, Lixin; Zhang, Yishi; He, Yi; Gao, Song; Zhu, Dunyao; Ran, Bin; Wu, Qing

    2016-01-01

    The ability to identify hazardous traffic events is already considered as one of the most effective solutions for reducing the occurrence of crashes. Only certain particular hazardous traffic events have been studied in previous studies, which were mainly based on dedicated video stream data and GPS data. The objective of this study is twofold: (1) the Markov blanket (MB) algorithm is employed to extract the main factors associated with hazardous traffic events; (2) a model is developed to identify hazardous traffic event using driving characteristics, vehicle trajectory, and vehicle position data. Twenty-two licensed drivers were recruited to carry out a natural driving experiment in Wuhan, China, and multi-sensor information data were collected for different types of traffic events. The results indicated that a vehicle's speed, the standard deviation of speed, the standard deviation of skin conductance, the standard deviation of brake pressure, turn signal, the acceleration of steering, the standard deviation of acceleration, and the acceleration in Z (G) have significant influences on hazardous traffic events. The sequential minimal optimization (SMO) algorithm was adopted to build the identification model, and the accuracy of prediction was higher than 86%. Moreover, compared with other detection algorithms, the MB-SMO algorithm was ranked best in terms of the prediction accuracy. The conclusions can provide reference evidence for the development of dangerous situation warning products and the design of intelligent vehicles. PMID:27420073

  11. Landfill Leachate Treatment Using Hybrid Up-Flow Anaerobic Sludge Blanket (HUASB Reactor

    Directory of Open Access Journals (Sweden)

    Mohd Bharudin Ridzuan

    2013-11-01

    Full Text Available Abstract: The Effect of the development process in the country would lead the increment of the solid wastes production. Malaysia as a developing country is also could not escape from the problem in its solid waste management. An important problem that associated to landfill is the production of leachate. Leachate contains dangerous substances such as organic matters, heavy metals, Nitrogen Ammonia and other materials that could pollute underground water source. The aim of the paper was to study landfill leachate treatment efficiency using Hybrid Upflow Anaerobic Sludge Blanket (HUASB reactor in lab-scale. This research was investigate the pollutant content in landfill leachate and determines the percentage of nutrient removal. Parameters used for this research, were Biochemical Oxygen Demand (BOD, Chemical Oxygen Demand (COD, Suspended Solid (SS, Total Nitrogen (TN, and Total Phosphorus (TP. The experiments were carried out in lab scaled constructed reactor,  30 days duration which samples for test had been taken each 3 days intervals. The results showed that HUASB reactor were capable in removal several parameters. It has great ability in removal of Total Phosphorus and Suspended Solid with 90.60% and 80.70% each. The result of COD removal showed an encouraging removal graph, with average percentage removal 73.70%. Average percentage removal for BOD is 64%. Total Nitrogen was less remove nutrient with average percentage removal 50.32%. From the results, it showed that HUASB reactor capable to remove organic pollutants from landfill leachate.

  12. The coolant purification system of the European test blanket modules: Preliminary design

    Energy Technology Data Exchange (ETDEWEB)

    Ciampichetti, A., E-mail: andrea.ciampichetti@enea.i [ENEA CR Brasimone, FPN-FISING, 40032 Camugnano (Bolivia, Plurinational State of) (Italy); Aiello, A.; Coccoluto, G. [ENEA CR Brasimone, FPN-FISING, 40032 Camugnano (Bolivia, Plurinational State of) (Italy); Ricapito, I. [Fusion for Energy, 08019 Barcelona (Spain); Liger, K. [CEA, DEN, DTN/STPA/LPC, Cadarache, 13108 St Paul-lez-Durance (France); Demange, D. [Karlsruhe Institute of Technology, ITEP-TLK, Postfach 36 40, 76021 Karlsruhe (Germany); Moreno, C. [EURATOM-CIEMAT Association, 28040 Madrid (Spain)

    2010-12-15

    The HCPB (Helium Cooled Pebble Bed) and HCLL (Helium Cooled Lithium Lead) Test Blanket Modules (TBMs), developed in EU to be tested in ITER, adopt helium at 80 bar as primary coolant. This paper contains a conceptual design of the TBMs Coolant Purification System (CPS) based on the need to remove permeated tritium and gas impurities. The following steps have been considered: identification of CPS design requirements; review of the purification systems developed for Helium Cooled Fission Reactors and proposed for DEMO Fusion Reactor; selection of the most promising technologies for CPS; indications on instrumentation and procedures for tritium balance. The proposed solution is a three-stage process constituted by an oxidiser to convert Q{sub 2} and CO to Q{sub 2}O and CO{sub 2}, an adsorption step, performed on molecular sieve at room temperature to remove Q{sub 2}O and CO{sub 2}, and a final step performed on a heated getter to remove residual impurities.

  13. Hazardous Traffic Event Detection Using Markov Blanket and Sequential Minimal Optimization (MB-SMO)

    Science.gov (United States)

    Yan, Lixin; Zhang, Yishi; He, Yi; Gao, Song; Zhu, Dunyao; Ran, Bin; Wu, Qing

    2016-01-01

    The ability to identify hazardous traffic events is already considered as one of the most effective solutions for reducing the occurrence of crashes. Only certain particular hazardous traffic events have been studied in previous studies, which were mainly based on dedicated video stream data and GPS data. The objective of this study is twofold: (1) the Markov blanket (MB) algorithm is employed to extract the main factors associated with hazardous traffic events; (2) a model is developed to identify hazardous traffic event using driving characteristics, vehicle trajectory, and vehicle position data. Twenty-two licensed drivers were recruited to carry out a natural driving experiment in Wuhan, China, and multi-sensor information data were collected for different types of traffic events. The results indicated that a vehicle’s speed, the standard deviation of speed, the standard deviation of skin conductance, the standard deviation of brake pressure, turn signal, the acceleration of steering, the standard deviation of acceleration, and the acceleration in Z (G) have significant influences on hazardous traffic events. The sequential minimal optimization (SMO) algorithm was adopted to build the identification model, and the accuracy of prediction was higher than 86%. Moreover, compared with other detection algorithms, the MB-SMO algorithm was ranked best in terms of the prediction accuracy. The conclusions can provide reference evidence for the development of dangerous situation warning products and the design of intelligent vehicles. PMID:27420073

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

    International Nuclear Information System (INIS)

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

  15. Thermal Analysis on Conceptual K-DEMO Breeding Blanket on parallel flow configuration

    International Nuclear Information System (INIS)

    The feasibility study consists the design guidelines and requirements for the K-DEMO(the Korean Fusion DEMOnstration reactor). It is possible to design flexible and realistic concepts of the demonstration fusion power plant. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the breeding blanket for the K-DEMO reactor. Recently, a new breeding blanket concept has been proposed for the K-DEMO reactor, and preliminary feasibility studies are actively ongoing. Design concept of parallel plate-type blanket of K-DEMO and thermal limits on components was identified, in this study. It was concluded that an acceptable thermal design was achieved in the proposed breeding blanket design. However, some design improvements in the geometry of the breeding blanket are ongoing. In aspect of pressure drop and outlet temperature, some analysis need to be conducted. Stage of study on the K-DEMO reactor is in the preliminary concept definition

  16. The integrated-blanket-coil concept applied to the spherical torus

    International Nuclear Information System (INIS)

    The purpose of this study is to investigate the potential of reactor embodiments based on the combination of two novel concepts, the integrated-blanket-coil (IBC) concept and the spherical torus (ST) concept. The IBC concept involves the combination of blanket and coil functions into a single component. The ST concept is based on the operation of the tokamak magnetic configuration at low aspect ratio. Two applications of the IBC concept to the ST are presented: (1) the IBC as the outer blanket and TF coil return legs (Application 1); and (2) the IBC as the inner blanket and OH solenoid coil set (Application 2). The application 1 IBC/ST yields reactor embodiments operating with high mass power density (≅500 kWe/MT) while, at the same time, yielding moderate neutron wall loading (≅5 MW/m/sup 2/) and modest electrical output (400-500 MWe). It is anticipated that fusion reactors operating in the above parameter space will exhibit attractive economic potential. When compared with a conventional ST design (copper TF coils and separate blanket), the Aplication 1 IBC/ST design exhibits a factor of two improvement in mass power density. The Application 2 IBC/ST provides an inductive current drive option which yields burn times in the range 3-4 minutes with no penalties in inboard breeding and energy recovery

  17. Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO

    International Nuclear Information System (INIS)

    Highlights: •A conceptual architectural model for a vertical maintenance DEMO is presented. •Novel concepts for a set of DEMO remote handling equipment are put forward. •Remote maintenance of a multi module segment blanket is found to be feasible. •The criticality of space in the vertical port is highlighted. -- Abstract: The anticipated high neutron flux, and the consequent damage to plasma-facing components in DEMO, results in the need to regularly replace the tritium breeding and radiation shielding blanket. The current European multi module segment (MMS) blanket concept favours a less invasive small port entry maintenance system over large sector transport concepts, because of the reduced impact on other tokamak systems – particularly the magnetic coils. This paper presents a novel conceptual remote maintenance strategy for a Vertical Maintenance Scheme DEMO, incorporating substantiated designs for an in-vessel mover, to detach and attach the blanket segments, and cask-housed vertical maintenance devices to open and close access ports, cut and join service connections, and extract blanket segments from the vessel. In addition, a conceptual architectural model for DEMO was generated to capture functional and spatial interfaces between the remote maintenance equipment and other systems. Areas of further study are identified in order to comprehensively establish the feasibility of the proposed maintenance system

  18. Stability of LMR oxide pins and blanket rods during run-beyond-cladding-break (RBCB) operation

    International Nuclear Information System (INIS)

    Since 1981, the U.S. Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan have collaborated on an operational reliability testing program in the Experimental Breeder Reactor II. The tests were designed to determine the irradiation behavior of liquid-metal reactor (LMR) oxide pins and blanket rods during steady-state, transient, and run-beyond-claddin-breach (RBCB) operation. Phase I tests completed in 1987 involved current LMR oxide designs and claddings; the phase II tests begun in 1988 concentrate on advanced LMR designs, large-diameter pins (7.5 mm), and advance cladding alloys. The cladding breaches in these tests have been readily detected by fission-gas and delayed-neutron (DN) precursor release. The condition of the fuel pin has been monitored by these releases during RBCB operation. A variety of failures have been intentionally studied in the RBCB portion of the program for operating times of up to 142 full-power days; also, several failure types have been incidentally experienced during the transient tests. Types of failure have included those induced by gas-pressure loading either naturally or by prethinning of the cladding defects, and fuel-cladding mechanical interaction (FCMI)-induced failures or secondary failures caused by the formation of low-density fuel-sodium reaction product (FSRP). This paper summarizes this experience with regard to LMR oxide fuel stability during RBCB operation

  19. Simultaneous degradation of cyanide and phenol in upflow anaerobic sludge blanket reactor.

    Science.gov (United States)

    Kumar, M Suresh; Mishra, Ram Sushil; Jadhav, Shilpa V; Vaidya, A N; Chakrabarti, T

    2011-07-01

    Coal coking, precious metals mining and nitrile polymer industries generate over several billion liters of cyanide-containing waste annually. Economic and environmental considerations make biological technologies attractive for treatment of wastes containing high organic content, in which the microbial cultures can remove concentrations of organics and cyanide simultaneously. For cyanide and phenol bearing waste treatment, an upflow anaerobic sludge blanket reactor has been developed, which successfully removed free cyanide 98% (with feed concentration of 20 mg 1(-1)) in presence of phenol. The effect of cyanide on phenol degradation was studied with varying concentrations of phenol as well as cyanide under anaerobic conditions. This study revealed that the methanogenic degradation of phenol can occur in the presence of cyanide concentration 30-38 mg 1(-1). Higher cyanide concentration inhibited the phenol degradation rate. The inhibition constant Ki was found to be 38 mg 1(-1) with phenol removal rate of 9.09 mg 1(-1.) x h.

  20. Analysis of the separation of protium from blanket tritium-product streams

    International Nuclear Information System (INIS)

    The case is considered in which the blanket product stream has been purified to the point where only protium, tritium, and a small quantity of deuterium remain. A cryogenic distillation cascade concept developed specifically to handle this enrichment problem is shown. The concept is based on a series of distillation columns and equilibrators capable of producing a protium-rich stream containing less than 1000 appm T and a tritium-rich stream containing less than 2000 appm H. It is envisioned that both of these streams could be blended with streams of comparable composition in the mainstream position of the fuel cycle without further processing. The computational analysis of the cascade was based on a fixed arrangement of columns and equilibrators and a fixed number of theoretical plates per columns, since these features are less easily varied in an actual system than reflux ratios and flow rates. In order to test the flexibility of this conceptual enruchment system to adjust to variations of the H/T ratio in the feed, H/T values of 0.333, 1.00, and 3.00 were investigated

  1. Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

    International Nuclear Information System (INIS)

    Highlights: • Optimized water distribution in ITER blanket modules is presented. • All key challenging constraints are included. • The methodology and the successful result are presented. - Abstract: This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements. The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented

  2. High temperature blankets for non-electrical/electrical applications of fusion reactors: Annual report, [1983

    International Nuclear Information System (INIS)

    During FY '83 the Li2O solid-breeder, helium-cooled canister blanket emerged as the LLNL-UW choice for driving the low-temperature (2, high-temperature outer zone for driving the GA hydrogen synfuel process. Providing 3-dimensional neutronics analysis of power deposition and tritium breeding in both blankets was an important part of the UW-Rowe and Assoc. work. In both the LLNL-UW and MARS studies, the fusion driver as the Axi-Cell, A-cell version of the tandem mirror reactor (TMR). Physics parameters consistent with the synfuel interface were determined as part of the work. Defining and analyzing the thermal-electric interfaces between the TMR and the synfuel process continues to be of prime importance. The analysis of thermal transport and energy conversion in the interface, as well as thermal hydraulics analysis of the blanket, were part of the UW-Rowe Assoc. work

  3. A novel duplex real-time reverse transcriptase-polymerase chain reaction assay for the detection of hepatitis C viral RNA with armored RNA as internal control

    Directory of Open Access Journals (Sweden)

    Meng Shuang

    2010-06-01

    Full Text Available Abstract Background The hepatitis C virus (HCV genome is extremely heterogeneous. Several HCV infections can not be detected using currently available commercial assays, probably because of mismatches between the template and primers/probes. By aligning the HCV sequences, we developed a duplex real-time reverse transcriptase-polymerase chain reaction (RT-PCR assay using 2 sets of primers/probes and a specific armored RNA as internal control. The 2 detection probes were labelled with the same fluorophore, namely, 6-carboxyfluorescein (FAM, at the 5' end; these probes could mutually combine, improving the power of the test. Results The limit of detection of the duplex primer/probe assay was 38.99 IU/ml. The sensitivity of the assay improved significantly, while the specificity was not affected. All HCV genotypes in the HCV RNA Genotype Panel for Nucleic Acid Amplification Techniques could be detected. In the testing of 109 serum samples, the performance of the duplex real-time RT-PCR assay was identical to that of the COBAS AmpliPrep (CAP/COBAS TaqMan (CTM assay and superior to 2 commercial HCV assay kits. Conclusions The duplex real-time RT-PCR assay is an efficient and effective viral assay. It is comparable with the CAP/CTM assay with regard to the power of the test and is appropriate for blood-donor screening and laboratory diagnosis of HCV infection.

  4. Neutronic Analysis of Flux Dispersion in a Multi-Layered, (D-T) Driven Hybrid Blanket

    OpenAIRE

    İPEK, Osman

    2000-01-01

    The concept of `hybrid blanket' is based on the placement of the nuclear fuel layer, which is a fertile material and fissionable by the fusion neutrons, at the front or the rear sides of the tritium breeding zone so that, in addition to gaining fission energy, a fissile fuel is produced. The neutronic flux distribution (neutron spectrum) along the radial direction varies with the type of material and the geometry used in the blanket, and also according to whether it is multi-layered or s...

  5. Use of Ball Blanket in attention-deficit/hyperactivity disorder sleeping problems

    DEFF Research Database (Denmark)

    Hvolby, Allan; Bilenberg, Niels

    2011-01-01

    Objectives: Based on actigraphic surveillance, attention-deficit/hyperactivity disorder (ADHD) symptom rating and sleep diary, this study will evaluate the effect of Ball Blanket on sleep for a sample of 8-13-year-old children with ADHD. Design: Case-control study. Setting: A child and adolescent...... psychiatric department of a teaching hospital. Participants: 21 children aged 8-13 years with a diagnosis of ADHD and 21 healthy control subjects. Intervention: Sleep was monitored by parent-completed sleep diaries and 28 nights of actigraphy. For 14 of those days, the child slept with a Ball Blanket. Main...

  6. Study on compact design of remote handling equipment for ITER blanket maintenance

    International Nuclear Information System (INIS)

    In the ITER, the neutrons created by D-T reactions activate structural materials, and thereby, the circumstance in the vacuum vessel is under intense gamma radiation field. Thus, the in-vessel components such as blanket are handled and replaced by remote handling equipment. The objective of this report is to study the compactness of the remote handling equipment (a vehicle/manipulator) for the ITER blanket maintenance. In order to avoid the interferences between the blanket and the equipment during blanket replacement in the restricted vacuum vessel, a compact design of the equipment is required. Therefore, the compact design is performed, including kinematic analyses aiming at the reduction of the sizes of the vehicle equipped with a manipulator handling the blanket and the rail for the vehicle traveling in the vacuum vessel. Major results are as follows: 1. The compact vehicle/manipulator is designed concentration on the reduction of the rail size and simplification of the guide roller mechanism as well as the reduction of the gear diameter for vehicle rotation around the rail. Height of the rail is reduced from 500 mm to 400 mm by a parameter survey for weight, stiffness and stress of the rail. The roller mechanism is divided into two simple functional mechanisms composed of rollers and a pad, that is, the rollers support relatively light loads during rail deployment and vehicle traveling while a pad supports heavy loads during blanket replacement. Regarding the rotation mechanism, the double helical gear is adopted, because it has higher contact ratio than the normal spur gear and consequently can transfer higher force. The smaller double helical gear, 996 mm in diameter, can achieve 26% higher output torque, 123.5 kN·m, than that of the original spur gear of 1,460 mm in diameter, 98 kN·m. As a result, the manipulator becomes about 30% lighter, 8 tons, than the original weight, 11.2 tons. 2. Based on the compact design of the vehicle/manipulator, the

  7. Activation Analysis for a He/LiPb dual Coolant Blanket for DEMO Reactor

    OpenAIRE

    Catalán, J.P.; Ogando Serrano, Francisco; Sanz Gonzalo, Javier

    2010-01-01

    The objective of the Spanish national project TECNO_FUS is to generate a conceptual design of a DCLL (Dual-Coolant Lithium-Lead) blanket for the DEMO fusion reactor. The dually-cooled breeding zone is composed of He/Pb-15.7 6Li and SiC as liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer-97) for the blanket and austenitic steel (316LN) for the Vacuum Vessel (VV). The goal of this work is to analyze the radioactive waste production by the neutron-i...

  8. RF test blanket sub-module with ceramic breeder and helium cooling for test in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kovalenko, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)]. E-mail: koval@nikiet.ru; Kapyshev, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Leshukov, A. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Poliksha, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Shatalov, G. [Russian Research Center ' Kurchatov Institute' , Kurchatov Square 1, 123182 Moscow (Russian Federation); Strebkov, Yu. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Strizhov, A. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Sviridenko, M. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)

    2006-02-15

    International thermonuclear experimental reactor (ITER) is anticipated as the only one step to DEMO fusion reactor. One of its main objectives is to demonstrate the availability and integration of technologies essential for a fusion reactor by testing of components for a future reactor including the test blanket modules (TBM) with different types of breeding materials. RF proposed to divide the TBM on two parts and to use two independent test blanket sub-modules (TBSM) which fixed on the frame in ITER horizontal experimental port for testing. CHC TBSM design description, its mechanical attachment on the frame, and principle schemes of helium cooling system and tritium cycle system are presented in this paper.

  9. Blanket concept of water-cooled lithium lead with beryllium for the SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    As an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. In SlimCS, the net tritium breeding ratio (TBR) supplied from WCLL blanket was not enough because the thickness of blanket in SlimCS was limited to about 0.5 m so as to allocate the conducting shell position near the plasma for high beta access and vertical stability of plasma. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier to reach a required TBR (≥ 1.05). Considering the operating temperature of blanket materials, a double pipe structure was adopted. The nuclear and thermal analysis were carried out by a nuclear-thermal-coupled code, ANIHEAT and DOHEAT so that blanket materials were appropriately arranged to satisfy the acceptable operation temperatures. The temperatures of materials were kept in appropriate range for the neutron wall load Pn = 5 MW/m2. It was found that the local TBR of WCLL with Be blanket was comparable with that of solid breeder blanket. (author)

  10. 77 FR 66597 - Chevron U.S.A. Inc.; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-11-06

    .... in Cameron Parish, Louisiana. Chevron states that DOE/FE has issued a number of blanket... U.S.A. Inc.; Application for Blanket Authorization To Export Previously Imported Liquefied Natural Gas on a Short-Term Basis AGENCY: Office of Fossil Energy, DOE. ACTION: Notice of application....

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

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

  13. 防弹材料及装备V50测试方法研究%Review of VS0 ballistic test method for bullet resistant materials and armor component

    Institute of Scientific and Technical Information of China (English)

    李茂辉; 黄献聪; 王雷; 周宏

    2011-01-01

    The V50 ballistic test method is the common way for testing and evaluating the ballistic performance of bullet resistant materials and armor component. This paper reviews the basic concept, developing course, main characteristics and application status of V50 ballistic test method. The future research direction and development trends are prospected.%V50测试方法是各种防弹材料及装备防弹性能评价测试的常用方法.在国内外相关参考文献研究的基础上,对V50指标及其测试方法的基本概念、发展沿革、主要特点和应用现状4个方面进行综述,对今后研究发展进行展望.

  14. On the use of double-walled tubes as a means to improve safety and availability of the EU DEMO Water-Cooled Pb-17Li blanket

    International Nuclear Information System (INIS)

    The impact on the blanket reliability and availability of both double-walled tube and welded joint failures in the Water-Cooled Pb-17Li Demo Single-Box blanket reference design is examined. The pertinence of employing a leak detection system is analysed and its contribution to the blanket safety and availability is evaluated. The contribution of welded joints to the blanket safety and availability is evaluated in normal operation. (orig.)

  15. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  16. Mathematical modeling of upflow anaerobic sludge blanket (UASB) reactor treating domestic wastewater.

    Science.gov (United States)

    Elmitwalli, Tarek

    2013-01-01

    Although the upflow anaerobic sludge blanket (UASB) reactor has been widely applied for domestic wastewater treatment in many developing countries, there is no sufficient mathematical model for proper design and operation of the reactor. An empirical model based on non-linear regression was developed to represent the physical and chemical removal of suspended solids (SS) in the reactor. Moreover, a simplified dynamic model based on ADM1 and the empirical model for SS removal was developed for anaerobic digestion of the entrapped SS and dissolved matter in the wastewater. The empirical model showed that effluent suspended chemical oxygen demand (COD(ss)) concentration is directly proportional to the influent COD(ss) concentration and inversely proportional to both the hydraulic retention time (HRT) of the reactor and wastewater temperature. For obtaining sufficient COD(ss) removal, the HRT of the UASB reactor must be higher than 4 h, and higher HRT than 12 h slightly improved COD(ss) removal. The dynamic model results showed that the required time for filling the reactor with sludge mainly depends on influent total chemical oxygen demand (COD(t)) concentration and HRT. The influent COD(t) concentration, HRT and temperature play a crucial role on the performance of the reactor. The results indicated that shorter HRT is needed for optimization of COD(t) removal, as compared with optimization of COD(t) conversion to methane. Based on the model results, the design HRT of the UASB reactor should be selected based on the optimization of wastewater conversion and minimization of biodegradable SS accumulation in the sludge bed, not only based on COD removal, to guarantee a stable reactor performance.

  17. Results of R and D for lithium/vanadium breeding blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L.; Reed, C.B.; Park, J.H. [Argonne National Lab., IL (United States); Kirillov, I.R. [D.V. Efremov Scientific Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Strebkov, Yu.S. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Rusanov, A.E. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Votinov, S.N. [A.A. Bochvar Inst. of Non-Organic Materials, Moscow (Russian Federation)

    1997-04-01

    The self-cooled lithium/vanadium blanket concept has several attractive features for fusion power systems, including reduced activation, resistance to radiation damage, accommodation of high heat loads and operating to temperatures of 650--700 C. The primary issue associated with the lithium/vanadium concept is the potentially high MHD pressure drop experienced by the lithium as it flows through the high magnetic field of the tokamak. The solution to this issue is to apply a thin insulating coating to the inside of the vanadium alloy to prevent the generation of eddy currents within the structure that are responsible for the high MHD forces and pressure drop. This paper presents progress in the development of an insulator coating that is capable of operating in the severe fusion environment, progress in the fabrication development of vanadium alloys, and a summary of MHD testing. A large number of small scale tests of vanadium alloy specimens coated with CaO and AlN have been conducted in liquid lithium to determine the resistivity and stability of the coating. In-situ measurements in lithium have determined that CaO coatings, {approximately} 5 {micro}m thick, have resistivity times thickness values exceeding 10{sup 6} {Omega}-cm{sup 2}. These results have been used to identify fabrication procedures for coating a large vanadium alloy (V-4Cr-4Ti) test section that was tested in the ALEX (Argonne Liquid metal Experiment) facility. Similar test sections have been produced in both Russia and the US.

  18. Design of Vibration Absorber using Spring and Rubber for Armored Vehicle 5.56 mm Caliber Rifle

    Directory of Open Access Journals (Sweden)

    Aditya Sukma Nugraha

    2014-12-01

    Full Text Available This paper presents a design of vibration absorber using spring and rubber for 5.56 mm caliber rifle armored vehicle. Such a rifle is used in a Remote-Controlled Weapon System (RCWS or a turret where it is fixed using a two degree of freedom pan-tilt mechanism. A half car lumped mass dynamic model of armored vehicles was derived. Numerical simulation was conducted using fourth order Runge Kutta method. Various types of vibration absorbers using spring and rubber with different configurations are installed in the elevation element. Vibration effects on horizontal direction, vertical direction and angular deviation of the elevation element was investigated. Three modes of fire were applied i.e. single fire, semi-automatic fire and automatic fire. From simulation results, it was concluded that the parallel configuration of damping rubber type 3, which has stiffness of 980,356.04 (N/m2 and damping coefficient of 107.37 (N.s/m, and Carbon steel spring whose stiffness coefficient is 5.547 x 106 (N/m2 provides the best vibration absorption. 

  19. Sediment mobility and bed armoring in the St Clair River: insights from hydrodynamic modeling

    Science.gov (United States)

    Liu, Xiaofeng; Parker, Gary; Czuba, Jonathan A.; Oberg, Kevin; Mier, Jose M.; Best, James L.; Parsons, Daniel R.; Ashmore, Peter; Krishnappan, Bommanna G.; Garcia, Marcelo H.

    2012-01-01

    The lake levels in Lake Michigan-Huron have recently fallen to near historical lows, as has the elevation difference between Lake Michigan-Huron compared to Lake Erie. This decline in lake levels has the potential to cause detrimental impacts on the lake ecosystems, together with social and economic impacts on communities in the entire Great Lakes region. Results from past work suggest that morphological changes in the St Clair River, which is the only natural outlet for Lake Michigan-Huron, could be an appreciable factor in the recent trends of lake level decline. A key research question is whether bed erosion within the river has caused an increase in water conveyance, therefore, contributed to the falling lake level. In this paper, a numerical modeling approach with field data is used to investigate the possibility of sediment movement in the St Clair River and assess the likelihood of morphological change under the current flow regime. A two-dimensional numerical model was used to study flow structure, bed shear stress, and sediment mobility/armoring over a range of flow discharges. Boundary conditions for the numerical model were provided by detailed field measurements that included high-resolution bathymetry and three-dimensional flow velocities. The results indicate that, without considering other effects, under the current range of flow conditions, the shear stresses produced by the river flow are too low to transport most of the coarse bed sediment within the reach and are too low to cause substantial bed erosion or bed scour. However, the detailed maps of the bed show mobile bedforms in the upper St Clair River that are indicative of sediment transport. Relatively high shear stresses near a constriction at the upstream end of the river and at channel bends could cause local scour and deposition. Ship-induced propeller wake erosion also is a likely cause of sediment movement in the entire reach. Other factors that may promote sediment movement, such as ice

  20. 75 FR 33803 - Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-06-15

    ... Energy Regulatory Commission Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization June 8, 2010. Take notice that on June 1, 2010, Sabine Pipe Line LLC (Sabine), 4800 Fournace Place, Bellaire... L. Kirk, Regulatory Specialist, Chevron Pipe Line Company, 4800 Fournace Place, Bellaire,...

  1. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  2. 77 FR 38622 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-06-28

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on June 4, 2012, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  3. 78 FR 68835 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-11-15

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on October 31, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  4. 78 FR 25264 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-04-30

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on April 16, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  5. 78 FR 53746 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-08-30

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on August 13, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  6. 77 FR 14517 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-03-12

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 21, 2012 Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 State Highway 56,...

  7. 78 FR 13663 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-02-28

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 11, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, P.O. Box...

  8. 75 FR 8053 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-02-23

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization February 16, 2010. Take notice that on January 29, 2010, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700...

  9. 32 CFR Appendix D to Part 505 - Exemptions; Exceptions; and DoD Blanket Routine Uses

    Science.gov (United States)

    2010-07-01

    ... 32 National Defense 3 2010-07-01 2010-07-01 true Exemptions; Exceptions; and DoD Blanket Routine Uses D Appendix D to Part 505 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS ARMY PRIVACY ACT PROGRAM Pt. 505, App. D Appendix D...

  10. Study of the effects of corrugated wall structures due to blanket modules around ICRH antennas

    Energy Technology Data Exchange (ETDEWEB)

    Dumortier, Pierre; Louche, Fabrice; Messiaen, André; Vervier, Michel [LPP-ERM/KMS, EURATOM-Belgian State Association, TEC partner, B-1000 Brussels (Belgium)

    2014-02-12

    In future fusion reactors, and in ITER, the first wall will be covered by blanket modules. These blanket modules, whose dimensions are of the order of the ICRF wavelengths, together with the clearance gaps between them will constitute a corrugated structure which will interact with the electromagnetic waves launched by ICRF antennas. The conditions in which the grooves constituted by the clearance gaps between the blanket modules can become resonant are studied. Simple analytical models and numerical simulations show that mushroom type structures (with larger gaps at the back than at the front) can bring down the resonance frequencies, which could lead to large voltages in the gaps between the blanket modules and perturb the RF properties of the antenna if they are in the ICRF operating range. The effect on the wave propagation along the wall structure, which is acting as a spatially periodic (toroidally and poloidally) corrugated structure, and hence constitutes a slow wave structure modifying the wall boundary condition, is examined.

  11. Transient electromagnetic and dynamic structural analyses of a blanket structure with coupling effects

    International Nuclear Information System (INIS)

    Transient electromagnetic and dynamic structural analyses of a blanket structure in the fusion experimental reactor (FER) under a plasma disruption event and a vertical displacement event (VDE) have been performed to investigate the dynamic structural characteristics and the feasibility of the structure. Coupling effects between eddy currents and dynamic deflections have also been taken into account in these analyses. In this study, the inboard blanket was employed because of our computer memory limitation. A 1/192 segment model of a full torus was analyzed using the analytical code, EDDYCUFF. In the plasma disruption event, the maximum magnetic pressure caused by eddy currents and poloidal fields was 1.2MPa. The maximum stress intensity by this magnetic pressure was 114MPa. In the VDE, the maximum magnetic pressure was 2.4MPa and the maximum stress intensity was 253MPa. This stress was somewhat beyond the allowable stress limit. Therefore, the blanket structure and support design should be reviewed to reduce the stress to a suitable value. In summary, the dynamic structural characteristics and design issues of the blanket structure have been identified. (orig.)

  12. Stochastic modeling to determine the economic effects of blanket, selective, and no dry cow therapy

    NARCIS (Netherlands)

    Huijps, K.; Hogeveen, H.

    2007-01-01

    In many countries, blanket dry cow therapy (DCT) is the standard way to dry off cows. Because of concerns about antibiotic resistance, selective DCT is proposed as an alternative. The economic consequences of different types of DCT were studied previously, but variation between input traits and diff

  13. 75 FR 38092 - The Dow Chemical Company; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-07-01

    ... The Dow Chemical Company (Dow), requesting blanket authorization to export liquefied natural gas (LNG... with its principal place of business in Midland, Michigan. Dow is an international chemical and...\\ The Dow Chemical Company, DOE/FE Order No. 2754 issued February 25, 2010. Current Application In...

  14. 77 FR 53874 - The Dow Chemical Company; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-09-04

    ... application (Application), filed on July 13, 2012, by The Dow Chemical Company (Dow), requesting blanket... comments are invited. \\1\\ The Dow Chemical Company, DOE/FE Order No. 2859 (October 5, 2010) extends through... chemical and plastics manufacturing company with operations in a number of U.S. states. Dow owns...

  15. 77 FR 31004 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-05-24

    ... Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on May 9, 2012, Southern Natural Gas Company (Southern), 569 Brookwood Village, Suite....210 of the Federal Energy Regulatory Commission's regulations under the Natural Gas Act (NGA),...

  16. 76 FR 18216 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2011-04-01

    ... Federal Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on March 16, 2011, Southern Natural Gas Company (Southern), Post Office Box 2563... and 157.216 of the Commission's Regulations under the Natural Gas Act (NGA) as amended, to abandon...

  17. 75 FR 13535 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-03-22

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization March 16, 2010. Take notice that on March 12, 2010, Northern Natural Gas Company (Northern), 1111 South... External Affairs, Northern Natural Gas Company, 1111 South 103rd Street, Omaha, Nebraska 68124, at...

  18. 75 FR 3232 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-01-20

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization January 8, 2010. Take notice that on December 30, 2009, Northern Natural Gas Company (Northern), 1111... sections 157.205 and 157.214 of the Commission's regulations under the Natural Gas Act for authorization...

  19. Two-dimensional cross-section sensitivity and uncertainty analysis for fusion reactor blankets

    International Nuclear Information System (INIS)

    A two-dimensional sensitivity and uncertainty analysis for the heating of the TF coil for the FED (fusion engineering device) blanket was performed. The uncertainties calculated are of the same order of magnitude as those resulting from a one-dimensional analysis. The largest uncertainties were caused by the cross section uncertainties for chromium

  20. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Science.gov (United States)

    2010-07-01

    ... Blanket Routine Uses (a) Routine Use—Law Enforcement. If a system of records maintained by a DoD Component, to carry out its functions, indicates a violation or potential violation of law, whether civil... decision concerning the hiring or retention of an employee, the issuance of a security clearance,...