WorldWideScience

Sample records for armored blanket development

  1. 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...

  2. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

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

  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. Development of the armor tile replacement system for FER

    International Nuclear Information System (INIS)

    Armor tiles attached to a first wall of the Fusion Experimental Reactor (FER) must be replaced several times by a remote replacement system since they are damaged due to plasma during reactor operation. Since positions of armor tiles are changed due to a thermal deformation and so on, the replacement system must be capable of adapting to the position changes. A remote control system has been developed to automatically replace armor tiles of first walls with a single manipulator arm for FER. The system is composed of a robot manipulator and an end-effector (a tile replacement hand), which has a tile gripper, a nut-runner to rotate fixing bolts of tiles and a vision sensor to measure positions of them. The system can replace the tiles by means of a visual feedback control using the vision sensor in case the tile positions have changed due to plant operations. Tests have proven that the remote control system is useful and practicable. (author)

  6. Remote handling system development of armor tile replacement for FER

    International Nuclear Information System (INIS)

    A number of armor tiles are attached to the first wall of the Fusion Experimental Reactor (FER) in order to protect the first wall against severe heat/particle loads from plasma during its operation. Although the armor tiles are made of heat-resisting materials such as graphite, they are eroded and damaged due to the loads and thus they are categorized into scheduled maintenance component. A remote handling system is required to replace a large number of tiles rapidly in the highly activated circumstance and has to be capable for adjusting a manipulator's motion taking into account a thermal deformation of the first wall and/or a positioning error of a manipulator for the remote handling system. For this purpose, a remote handling system of the armor tile replacement with a visual feedback control has been fabricated and this paper describes an experimental system and the performance test results

  7. 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)

  8. 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)

  9. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  12. 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.

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

    International Nuclear Information System (INIS)

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

  14. 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.)

  15. 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

  16. Fabrication techniques development of test blanket module based on CLAM

    International Nuclear Information System (INIS)

    The Reduced Activation Ferritic/Martensitic steels (RAFMs) are considered as the primary candidate structural material for the DEMO fusion reactor and the first fusion power plant. China Low Activation Martensitic (CLAM) steel, a version of RAFMs, is being developed in ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences), under wide collaboration with many institutes and universities in China and overseas. The designs of FDS (Fusion Design Study) series liquid LiPb blankets for fusion reactors and corresponding Dual Functional Lithium Lead (DFLL) Test Blanket Module (TBM) in International Thermonuclear Experimental Reactor (ITER) are being carried out in ASIPP. And CLAM steel is chosen as the primary candidate structural material in these designs. So the fabrication techniques for DFLL TBM with CLAM are or urgently needed to be studied in detail. The fabrication of DFLL TBM mainly includes the manufacturing of the First Wall (FW), the Cooling Plates (CP) and the joining of the FW and CPs. Currently, solid Hot Isostatic Pressing (HIP) bonding and uniaxial diffusion bonding method are the most promising candidate fabrication method for the FW and CP. Experiments of HIP and unixial diffusion bonding of CLAM/CLAM were carried out and good joints were obtained. As for the joining technique of FW and CPs, the fusion welding techniques such as Tungsten Inert Gas welding, Laser welding and Electron Beam welding are candidates. Preliminary experiments on these welding techniques were performed. The simulation of thermal process by Gleeble 2000 was also carried out. Results of these experiments are summarized and further R and D plan on blanket fabrication techniques is also stated. (authors)

  17. Development of remote replacement system for armor tiles of first wall of FER

    International Nuclear Information System (INIS)

    A remote system has been developed to replace automatically armor tiles of first walls with a single manipulator arm for the Fusion Experimental Reactor (FER). The system is composed of a manipulator arm and an end-effector (a tile replacement hand), which have a gripper of the tiles, a nutrunner to rotate attatching bolts of them and a vision sensor to measure positions of them. The system can replace the tiles by means of a visual feedback system using vision sensor, even if the positions of the tiles would have changed. As a result of tests, it has been proved that the end-effector is useful and the control system is practicable. (author)

  18. Welding techniques development of CLAM steel for Test Blanket Module

    International Nuclear Information System (INIS)

    Fabrication techniques for Test Blanket Module (TBM) with CLAM are being under development. Effect of surface preparation on the HIP diffusion bonding joints was studied and good joints with Charpy impact absorbed energy close to that of base metal have been obtained. The mechanical properties test showed that effect of HIP process on the mechanical properties of base metal was little. Uniaxial diffusion bonding experiments were carried out to study the effect of temperature on microstructure and mechanical properties. And preliminary experiments on Electron Beam Welding (EBW), Tungsten Inert Gas (TIG) Welding and Laser Beam Welding (LBW) were performed to find proper welding techniques to assemble the TBM. In addition, the thermal processes assessed with a Gleeble thermal-mechanical machine were carried out as well to assist the fusion welding research.

  19. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  20. 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)

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

    Science.gov (United States)

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. PMID:22112596

  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. Development of a virtual reality simulator for the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    The authors developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robotic simulation software, ENVISION. The simulator is connected to the control system of the manipulator, which was developed as part of the blanket maintenance system during the Engineering Design Activity (EDA), and can reconstruct the positions of the manipulator and blanket module using position data transmitted from motors through a LAN. In addition, it can provide virtual visual information (e.g., about the interface structures behind the blanket module) by making the module transparent on the screen. It can also be used for confirming a maintenance sequence before the actual operation. The simulator will be modified further, with addition of other necessary functions, and will finally serve as a prototype of the actual simulator for the blanket remote handling system, which will be procured as part of an in-kind contribution

  4. 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

  5. Materials development for ITER shielding and test blanket in China

    International Nuclear Information System (INIS)

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

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

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. Design and analysis of ITER shield blanket

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  10. Development of conductively cooled first wall armor and actively cooled divertor structure for ITER/FER

    International Nuclear Information System (INIS)

    Based on the design requirements for the plasma facing components in ITER/FER, we have performed design studies on the conductively cooled first wall armor and the divertor plate with sliding supports. The full-scale armor tiles were fabricated for heat load tests, and good thermal performances were obtained in heat load tests of 0.2-0.4 MW/m2. It is shown by the thermomechanical analysis on the divertor plate that thermal stresses and bending deformation are reduced significantly by using the sliding supports. The divertor test module with the sliding supports has been fabricated to investigate its fabricability and to verify the functions of the sliding supports during a high heat load of about 10 MW/m2. (orig.)

  11. Development of insulating coatings for liquid metal blankets

    International Nuclear Information System (INIS)

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

  12. 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.)

  13. 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....

  14. 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.

  15. 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

  16. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    International Nuclear Information System (INIS)

    This paper presents results of conceptual design activities and associated R and D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R and D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  17. Phase III experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    A pseudo-line source has been realized by using an accelerator based D-T point neutron source. The pseudo-line source is obtained by time averaging of continuously moving point source or by superposition of finely distributed point sources. The line source is utilized for fusion blanket neutronics experiments with an annular geometry so as to simulate a part of a tokamak reactor. The source neutron characteristics were measured for two operational modes for the line source, continuous and step-wide modes, with the activation foil and the NE213 detectors, respectively. In order to give a source condition for a successive calculational analysis on the annular blanket experiment, the neutron source characteristics was calculated by a Monte Carlo code. The reliability of the Monte Carlo calculation was confirmed by comparison with the measured source characteristics. The shape of the annular blanket system was a rectangular with an inner cavity. The annular blanket was consist of 15 mm-thick first wall (SS304) and 406 mm-thick breeder zone with Li2O at inside and Li2CO3 at outside. The line source was produced at the center of the inner cavity by moving the annular blanket system in the span of 2 m. Three annular blanket configurations were examined; the reference blanket, the blanket covered with 25 mm thick graphite armor and the armor-blanket with a large opening. The neutronics parameters of tritium production rate, neutron spectrum and activation reaction rate were measured with specially developed techniques such as multi-detector data acquisition system, spectrum weighting function method and ramp controlled high voltage system. The present experiment provides unique data for a higher step of benchmark to test a reliability of neutronics design calculation for a realistic tokamak reactor. (J.P.N.)

  18. Neutronics R and D efforts in support of the European breeder blanket development programme

    International Nuclear Information System (INIS)

    The EU fusion technology programme considers two blanket development lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles as breeder material and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy acting both as breeder and neutron multiplier. The long-term strategy aims at providing validated engineering designs of breeder blankets for a fusion power demonstration reactor (DEMO). As an important intermediate step, the breeder blankets need to be tested in a real fusion environment as provided by ITER. HCPB and HCLL Test Blanket Modules (TBM) have been accordingly designed for tests in dedicated ITER blanket ports. The nuclear design and performance of the breeder blanket modules rely on the results provided by neutronics design calculations. Validated computational tools and qualified nuclear data are required for high prediction accuracies including reliable uncertainty assessments. Complementary to the application of established standard tools and data for design analysis, a dedicated neutronics R and D effort is therefore conducted in the EU. This includes the development of dedicated computational tools, the generation of high quality nuclear data and their validation through integral experiments. The recent neutronic design efforts have been devoted to the European DEMO reactor study comprising (i) Monte Carlo based pre-analysis for the dimensioning of the shielding system, (ii) the generation of a generic CAD based Monte Carlo geometry model, and (iii) performance analysis for HCLL and HCPB based DEMO variants. The recent focus of the validation effort is on neutronics TBM mock-up experiments. The first experiment of this kind was performed on a TBM mock-up of the HCPB breeder blanket. The follow-up experiment on a neutronics HCLL TBM mock-up is currently under preparation. Computational pre-analysis were performed to optimise the design of the mock

  19. 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)

  20. 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

  1. Improved structure and long-life blanket concepts for heliotron reactors

    International Nuclear Information System (INIS)

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity COE. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armor tiles that soften the neutron energy spectrum incident on the self-cooled Flibe-RAF blanket. In this adaptation of the Spectral-shifter and Tritium breeder Blanket (STB) concept a local tritium breeding ratio TBR over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the super-conducting magnet coils is also significantly improved. Using the constant cross sections of helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armor tiles. The key R and D issues to develop the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated. (author)

  2. First-wall and blanket engineering development for magnetic-fusion reactors

    International Nuclear Information System (INIS)

    A number of programs in the USA concerned with materials and engineering development of the first wall and breeder blanket systems for magnetic-fusion power reactors are described. Argonne National Laboratory has the lead or coordinating role, with many major elements of the research and engineering tests carried out by a number of organizations including industry and other national laboratories

  3. 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

  4. Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

    International Nuclear Information System (INIS)

    Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R and D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R and D progress on these areas are introduced here

  5. Preliminary design of test facilities for tritium breeding blanket development, (1)

    International Nuclear Information System (INIS)

    This report describes the results of the preliminary design of outpile test facilities which are used for development of tritium breeding blanket with ceramic breeding material. The facilities which were designed are as follows; High heat flux test facility, Thermal-hydraulic test facility, Integrity test facility, Fabrication Technology Development Facility. This design study was performed by Kawasaki Heavy Industries, Ltd. under the contract to Fusion Research System Laboratory. (author)

  6. An overview on tritium permeation barrier development for WCLL blanket concept

    International Nuclear Information System (INIS)

    The reduction of tritium permeation through blanket structural materials and cooling tubes has to be carefully evaluated to minimise radiological hazards. A strong effort has been made in the past to select the best technological solution for the realisation of tritium permeation barriers (TPB) on complex structures not directly accessible after the completion of the manufacturing process. The best solution was identified in aluminium rich coatings, which form Al2O3 at their surface. Two technologies were selected as reference for the realisation of coating in the WCLL blanket concept: the chemical vapour deposition (CVD) process developed on laboratory scale by CEA, and the hot dipping (HD) process developed by FZK. The results obtained during three years of tests on CVD and HD coated specimens in gas and liquid metal phase are summarised and discussed

  7. Development of high temperature fusion blanket with LiPb-SiC and its socio-economic aspects

    International Nuclear Information System (INIS)

    This paper describes recent results of the development of SiC-LiPb blanket in Kyoto University with advanced SiC composite, and its implication for high temperature blanket preferable from socio-economic aspects. Blanket is the interface between fusion energy and outside world, i.e., industry, public and environment. Material and energy balances, such as fuel supply and waste discharge, or supply of product energy from fusion plants are characterized by the specific blanket concepts, and performance and characteristics of blankets are considered to dominate the feature of fusion energy that should respond to the requirements of the sponsors and public. Thus development strategy for blanket concepts are affected by the aspects of socio-economics such as; environmental release, assumed accidental scenario, rad-waste, and deployment to the market in each countries/area. Selections of breeders, coolants, and structural materials are based on such consideration, and ITER/TBM is expected to foster the evolving concepts. Combination of LiPb, helium and SiC is of particular interests for a demo blanket concept, because it is expected to be feasible in early stage of development such as TBM by Dual Coolant Lithium Lead concept as an intermediate step, and would eventually achieve high operating temperature above current fission reactors'. In the power plant design in Europe, US and Japan, Lithium lead will work as breeder and primary heat medium, and SiC composites are used for flow insert, enclosure and Intermediate Heat Exchanger to generate high temperature helium that is considered as a medium for power generation including possible hydrogen production. Staged development strategy can be planned; and experiments to verify compatibility, tritium permeability, MHD pressure drop and heat transfer characteristics are being evaluated with LiPb-He loop in Kyoto University. Hydrogen production process with this blanket is also experimentally studied. Based on the results

  8. 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

  9. Development of electrically insulating coatings on vanadium alloys for lithium-cooled blankets

    International Nuclear Information System (INIS)

    The self-cooled lithium blanket concept with a vanadium structure offers a potential for high performance with attractive safety and environmental features. Based on blanket design studies, it became apparent that electrically insulating duct walls would be required to reduce the magnetohydrodynamic (MHD) pressure drop for liquid metal-cooled blankets for high magnetic field fusion devices. As a result, development of insulator coatings was recommended as the most appropriate approach for resolving this issue. Oxides such as CaO, Y2O3, BeO, MgO, MgAl2O4, and Y3Al2O12 and nitrides such as AlN, BN and Si3N2 were initially considered potential candidate coating materials. Based on results of scoping studies, CaO and AlN have been selected as primary candidates for further development. Progress on the development of CaO and AlN coatings, including in-situ formation and electrical properties measurements, are summarized in this paper

  10. 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

  11. Development of vanadium base alloys for fusion first-wall/blanket applications

    International Nuclear Information System (INIS)

    Vanadium alloys have been identified as a leading candidate material for fusion first-wall/blanket applications. Certain vanadium alloys exhibit favorable safety and environmental characteristics, good fabricability, high temperature and heat load capability, good compatibility with liquid metals and resistance to irradiation damage effects. The current focus is on vanadium alloys with (3-5)% Cr and (3-5)% Ti with a V-4Cr-4Ti alloy as the leading candidate. Preliminary results indicate that the crack-growth rates of certain alloys are not highly sensitive to irradiation. Results from the Dynamic Helium Charging Experiment (DHCE) which simulates fusion relevant helium/dpa ratios are similar to results from neutron irradiated material. This paper presents an overview of the recent results on the development of vanadium alloys for fusion first wall/blanket applications

  12. 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.

  13. NDE of hybrid armor structures using acoustography

    International Nuclear Information System (INIS)

    The US Army is investigating the use of composite materials to deliver lightweight and more effective armor protection systems to soldiers and other army assets. However, widespread use of such hybrid armor will require a reliable but fast NDE methodology to ensure integrity of these components during manufacturing and while in service. Traditional ultrasonic inspection of such hybrid armor structures may prove to be very effective, but point-by-point ultrasonic scanning is inherently time-consuming and manufacturing slowdowns could develop in high-volume production of such armor systems. In this paper, we report on the application of acoustography for the NDE of hybrid armor structures. Acoustography differs from conventional ultrasonic testing in that test objects are inspected in full field, analogously to real time x-ray imaging. The approach uses a novel, super high resolution large area acousto-optic (AO) sensor, which allows image formation through simple ultrasound shadow casting, analogous to x-ray image formation. This NDE approach offers significant inspection speed advantage over conventional point-by-point ultrasonic scanning procedures and is well-suited for high volume production. We will report initial results on a number of hybrid armor plate specimens employing composite materials that are being investigated by the US Army. Acoustography NDE results will also be verified using other complimentary NDE methods.

  14. Fabrication of prototype mockups of ITER shielding blanket with separable first wall

    International Nuclear Information System (INIS)

    Design of shielding blanket for ITER-FEAT applies the first wall which has the separable structure from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. This report summarizes the demonstrative fabrication of the ITER shielding blanket with separable first wall performed for the shielding blanket fabrication technology development, under the task agreement of G 16 TT 108 FJ (T420-2) in ITER Engineering Design Activity Extension Period. The objectives of the demonstrative fabrication are: to demonstrate the comprehensive fabrication technique in a large scale component (e.g the joining techniques for the beryllium armor/copper alloy and copper alloy/SS, and the slotting method of the FW and shield block); to develop an improved fabrication method for the shielding blanket based on the ITER-FEAT updated design. In this work, the fabrication technique of full scale separable first wall shield blanket was confirmed by fabricating full width Be armored first wall panel, full scale of 1/2 shield block with poloidal cooling channels. As the R and D for updated cooling channel configuration, the fabrication technique of the radial channel shield block was also demonstrated. Concluding to the all R and D results, it was demonstrated successfully that the fabrication technique and optimized conditions in the results obtained under the task agreement of G 16 TT 95 FJ (T420-1) was applicable to the prototype of the separable first wall blanket module. Additionally, basic echo data of ultra-sonic test method (UT) was obtained to show the applicability of UT method for in tube access detection of defect on the Cu alloy/SS tube interface. (author)

  15. Development of a two-dimensional nuclear-thermal-coupled analysis code for conceptual blanket design of fusion reactors

    International Nuclear Information System (INIS)

    For conceptual blanket design of fusion reactors, a two-dimensional (2-D) nuclear-thermal-coupled analysis code, DOHEAT, has been developed. A remarkable feature of the code is a user-friendly graphical interface which allows to prepare an appropriate analysis model of a variety of blanket concepts with different materials in a short time. Furthermore, the code has a strong point that it provides the cross-sectional distribution of temperature, neutron/gamma fluxes, nuclear heating and tritium breeding ratio (TBR) inside the blanket after the completion of a single run. Throughout the DEMO design work in JAEA, it has proven that DOHEAT is extremely valuable in guiding the design study toward a wide variety of blanket concepts.

  16. Reproductive behavior, development and eye regression in the cave armored catfish, Ancistrus cryptophthalmus Reis, 1987 (Siluriformes: Loricariidae), breed in laboratory

    OpenAIRE

    Sandro Secutti; Eleonora Trajano

    2009-01-01

    The troglobitic armored catfish, Ancistrus cryptophthalmus (Loricariidae, Ancistrinae) is known from four caves in the São Domingos karst area, upper rio Tocantins basin, Central Brazil. These populations differ in general body shape and degree of reduction of eyes and of pigmentation. The small Passa Três population (around 1,000 individuals) presents the most reduced eyes, which are not externally visible in adults. A small group of Passa Três catfish, one male and three females, reproduced...

  17. Fabrication of the full scale separable first wall of ITER shielding blanket

    International Nuclear Information System (INIS)

    Shielding blanket for ITER-FEAT applies the unique first wall structure which is separable from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. Such unique features of blanket structure required technological clarification from the technical base of the previous achievement of the blanket module fabrication development. Previously, within the EDA Task T216+, a prototype for the no.4 Primary Wall Module of the ITER Shield Blanket with integrated first wall has been manufactured by forging and drilling and the first wall has been manufactured and joined to the shield block by Hot Isostatic Pressing (HIP) in one step process. This work has been performed to clarify the remaining R and D issues which have not been covered in the previous R and D. This report summarizes the demonstrative fabrication of the real scale separable first wall for ITER shielding blanket designed for ITER-FEAT, together with the essential technology developments such as, the slit grooving of the first wall with beryllium armor and SS shield block and fabrication of a partial mockup of beryllium armored first wall panel with built-in cooling channels. This work has been performed under the task agreement of G 16 TT 95 FJ (T420-1) in ITER Engineering Design Activity Extension Period. By the demonstration of the Be armor joining to the first wall panel, the joining technique of Be and DSCu developed previously, was shown to be applicable to the realistic structure of first wall panel. Also, the slit grooving by an end-mill method and an electron discharge machining method have been applied to the first wall mockup with Be armor tiles and demonstrated the applicability within the design tolerance. As the slit grooving technique

  18. Limitations on blanket performance

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 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.)

  1. Neutronics R&D efforts in support of the European breeder blanket development programme

    Science.gov (United States)

    Fischer, U.; Batistoni, P.; Klix, A.; Kodeli, I.; Leichtle, D.; Perel, R. L.

    2009-06-01

    The recent progress in the R&D neutronics efforts spent in the EU to support the development of the HCLL and HCPB breeder blankets is presented. These efforts include neutronic design activities performed in the framework of the European DEMO reactor study, validation efforts by means of neutronics mock-up experiments using 14 MeV neutron generators and the development of dedicated computational tools such as the conversion software McCad for the automatic generation of a Monte Carlo geometry model from available CAD data, and the MCSEN code for Monte Carlo based calculations of sensitivities and uncertainties by using the track length estimator. The supporting validation effort is devoted to the capability of the neutronics tools and data to predict the tritium production and other nuclear responses of interest in neutronics mock-up experiments. Such an experiment has been conducted on a HCPB mock-up while another on a HCLL mock-up is in progress.

  2. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  3. 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

  4. 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)

  5. 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.

  6. 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...

  7. 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.

  8. 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.

  9. 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)

  10. 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.)

  11. Reproductive behavior, development and eye regression in the cave armored catfish, Ancistrus cryptophthalmus Reis, 1987 (Siluriformes: Loricariidae, breed in laboratory

    Directory of Open Access Journals (Sweden)

    Sandro Secutti

    2009-09-01

    Full Text Available The troglobitic armored catfish, Ancistrus cryptophthalmus (Loricariidae, Ancistrinae is known from four caves in the São Domingos karst area, upper rio Tocantins basin, Central Brazil. These populations differ in general body shape and degree of reduction of eyes and of pigmentation. The small Passa Três population (around 1,000 individuals presents the most reduced eyes, which are not externally visible in adults. A small group of Passa Três catfish, one male and three females, reproduced spontaneously thrice in laboratory, at the end of summertime in 2000, 2003 and 2004. Herein we describe the reproductive behavior during the 2003 event, as well as the early development of the 2003 and 2004 offsprings, with focus on body growth and ontogenetic regression of eyes. The parental care by the male, which includes defense of the rock shelter where the egg clutch is laid, cleaning and oxygenation of eggs, is typical of many loricariids. On the other hand, the slow development, including delayed eye degeneration, low body growth rates and high estimated longevity (15 years or more are characteristic of precocial, or K-selected, life cycles. In the absence of comparable data for close epigean relatives (Ancistrus spp., it is not possible to establish whether these features are an autapomorphic specialization of the troglobitic A. cryptophthalmus or a plesiomorphic trait already present in the epigean ancestor, possibly favoring the adoption of the life in the food-poor cave environment. We briefly discuss the current hypotheses on eye regression in troglobitic vertebrates.O cascudo troglóbio (exclusivamente subterrâneo, Ancistrus cryptophthalmus (Loricariidae, Ancistrinae é conhecido de quatro cavernas na área cárstica de São Domingos, bacia do alto rio Tocantins, Goiás. Estas populações diferem quanto ao formato geral do corpo e grau de redução dos olhos e da pigmentação cutânea. A pequena população encontrada na caverna Passa Tr

  12. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

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

  13. Development of tailorable advanced blanket insulation for advanced space transportation systems

    Science.gov (United States)

    Calamito, Dominic P.

    1987-01-01

    Two items of Tailorable Advanced Blanket Insulation (TABI) for Advanced Space Transportation Systems were produced. The first consisted of flat panels made from integrally woven, 3-D fluted core having parallel fabric faces and connecting ribs of Nicalon silicon carbide yarns. The triangular cross section of the flutes were filled with mandrels of processed Q-Fiber Felt. Forty panels were prepared with only minimal problems, mostly resulting from the unavailability of insulation with the proper density. Rigidizing the fluted fabric prior to inserting the insulation reduced the production time. The procedures for producing the fabric, insulation mandrels, and TABI panels are described. The second item was an effort to determine the feasibility of producing contoured TABI shapes from gores cut from flat, insulated fluted core panels. Two gores of integrally woven fluted core and single ply fabric (ICAS) were insulated and joined into a large spherical shape employing a tadpole insulator at the mating edges. The fluted core segment of each ICAS consisted of an Astroquartz face fabric and Nicalon face and rib fabrics, while the single ply fabric segment was Nicalon. Further development will be required. The success of fabricating this assembly indicates that this concept may be feasible for certain types of space insulation requirements. The procedures developed for weaving the ICAS, joining the gores, and coating certain areas of the fabrics are presented.

  14. Al-based anti-corrosion and T-permeation barrier development for future DEMO blankets

    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.; Holstein, N.; Zimmermann, H. [Karlsruhe Institute of Technology, Hermann von Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2011-10-01

    In the Helium-Cooled-Liquid-Lead (HCLL) design of Test-Blanket-Modules (TBM's) for a future fusion power plant Pb-15.7Li is used as liquid breeder which is in direct contact with the structure material, e.g. EUROFER steel. Compatibility testing showed that high corrosion attack appears and that the dissolved steel components form precipitates with a high risk of system blockages. A reliable operation needs coatings as corrosion barriers. The earlier developed Hot-Dip Aluminisation (HDA) process has shown that Al-based scales can act as anti-corrosion as well as T-permeation barriers. Meanwhile two advanced electro-chemically based processes for deposition of Al-scales were successfully developed. The first (ECA = Electro-Chemical Al-deposition) is working with an organic electrolyte and the second one (ECX = Electro-Chemical-X-metal-deposition) is based on ionic liquids. Coatings in the {mu}m-range were deposited homogeneously with exact controllable thicknesses. Metallurgical investigations showed the successful generation of protective scales and compatibility testing demonstrated the barrier function.

  15. Al-based anti-corrosion and T-permeation barrier development for future DEMO blankets

    Science.gov (United States)

    Krauss, W.; Konys, J.; Holstein, N.; Zimmermann, H.

    2011-10-01

    In the Helium-Cooled-Liquid-Lead (HCLL) design of Test-Blanket-Modules (TBM's) for a future fusion power plant Pb-15.7Li is used as liquid breeder which is in direct contact with the structure material, e.g. EUROFER steel. Compatibility testing showed that high corrosion attack appears and that the dissolved steel components form precipitates with a high risk of system blockages. A reliable operation needs coatings as corrosion barriers. The earlier developed Hot-Dip Aluminisation (HDA) process has shown that Al-based scales can act as anti-corrosion as well as T-permeation barriers. Meanwhile two advanced electro-chemically based processes for deposition of Al-scales were successfully developed. The first (ECA = Electro-Chemical Al-deposition) is working with an organic electrolyte and the second one (ECX = Electro-Chemical-X-metal-deposition) is based on ionic liquids. Coatings in the μm-range were deposited homogeneously with exact controllable thicknesses. Metallurgical investigations showed the successful generation of protective scales and compatibility testing demonstrated the barrier function.

  16. Al-based anti-corrosion and T-permeation barrier development for future DEMO blankets

    International Nuclear Information System (INIS)

    In the Helium-Cooled-Liquid-Lead (HCLL) design of Test-Blanket-Modules (TBM's) for a future fusion power plant Pb-15.7Li is used as liquid breeder which is in direct contact with the structure material, e.g. EUROFER steel. Compatibility testing showed that high corrosion attack appears and that the dissolved steel components form precipitates with a high risk of system blockages. A reliable operation needs coatings as corrosion barriers. The earlier developed Hot-Dip Aluminisation (HDA) process has shown that Al-based scales can act as anti-corrosion as well as T-permeation barriers. Meanwhile two advanced electro-chemically based processes for deposition of Al-scales were successfully developed. The first (ECA = Electro-Chemical Al-deposition) is working with an organic electrolyte and the second one (ECX = Electro-Chemical-X-metal-deposition) is based on ionic liquids. Coatings in the μm-range were deposited homogeneously with exact controllable thicknesses. Metallurgical investigations showed the successful generation of protective scales and compatibility testing demonstrated the barrier function.

  17. Helium Gas Permeability of SiC/SiC Composite Developed for Blanket Component

    International Nuclear Information System (INIS)

    To employ SiC/SiC composite as blanket components of a fusion reactor, permeation behavior of helium gas has to be investigated since the helium is used as the coolant. For this purpose, a vacuum system consisting of upstream and downstream chambers was fabricated for the measurement of permeability, and the permeability was measured for several SiC/SiC composite materials recently developed. For the pressure range from 102 to 105 Pa in the upstream chamber, the pressure rise due to the permeation of helium in the downstream chamber linearly increased with the pressure of the upstream chamber. Then, the permeability was roughly constant for the pressure range of the upstream chamber.The permeabilities of SiC/SiC composites produced by polymer impregnation and pyrolysis (PIP), hot pressing (HP) and melt-infiltration-finished PIP were 5 x 10-5, 4 x 10-6, and 9 x 10-7 m2/s, respectively. In the matrix structure of the SiC/SiC composite made by the PIP method with a high permeability, cracking in the matrix and pores of micron size were observed. Compared to these materials, SiC/SiC composites produced by liquid phase sintering using submicron or nanopowder of β-SiC and the HP method had extremely low permeabilities. The permeability of the SiC/SiC composite made by using submicron or nanopowder of β-SiC became 1.5 x 10-9 or 4 x 10-11 m2/s.Based upon the present data, the helium gas flow was analyzed for a blanket module consisting of only SiC/SiC composite. If a vacuum pump is attached to the module, the helium leak into a plasma can be ignored, compared to the heliums produced by fusion reactions. Hence, the entire module can be made by only SiC/SiC composite, from a viewpoint of helium permeation

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

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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...

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

    International Nuclear Information System (INIS)

    The Japan-USA collaborative program, JUPITER-II, has made significant progress in a research program titled 'The irradiation performance and system integration of advanced blanket' through a six-year plan for 2001-2006. The scientific concept of this program is to study the elemental technology in macroscopic system integration for advanced fusion blankets based on an understanding of the relevant mechanics at the microscopic level. The program has four main research emphases: (1)Flibe molten salt system: Flibe handling, reduction-oxidation control by Be and Flibe tritium chemistry; thermofluid flow simulation experiment and numerical analysis. (2)Vanadium /Li system: MHD ceramics coating of vanadium alloys and compatibility with Li; neutron irradiation experiment in Li capsule and radiation creep. (3)SiC/He system: Fabrication of advanced composites and property evaluation; thermomechanics of SiC system with solid breeding materials; neutron irradiation experiment in He capsule at high temperatures. (4)Blanket system modeling: Design-based integration modeling of Flibe system and V/Li system; multiscale materials system modeling including He effects. This paper describes the perspective of the program including the historical background, the organization and facilities, and the task objectives. Important recent results are reviewed

  5. Comparison analysis of fusion breeder blanket concepts

    International Nuclear Information System (INIS)

    Based on the wide survey, the development status and key issues of fusion breeder blanket concepts are summarized. Two types of blanket concepts, i.e. solid and liquid breeder blanket, were compared and assessed in terms of engineering feasibility, tritium recovery and control, economic and safety aspects, etc. The advantages and disadvantages of the two types of blanket concepts were clarified from the viewpoint of technology realization and development potential. This study may act as a valuable reference for fusion blanket concept selection and design. (authors)

  6. 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

  7. 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.)

  8. The development of a direct insulation layer for the liquid metal cooled fusion reactor blanket

    International Nuclear Information System (INIS)

    The suppression of MHD pressure drops in the channels, in which liquid metal is flowing in a strong magnetic field, is necessary to get a sufficient cooling effect in the self-cooled liquid metal blanket or similar arrangements of a blanket structure. The MHD effects can significantly be reduced by means of electrical insulation of the flowing liquid metal against the structural material. The insulating material has to provide a resistivity of ≥ 25 Ωm, it has to be compatible with the liquid metal and should be sufficiently stable against irradiation damage and fracture due to thermal and mechanical cycling stresses. The liquid metal blanket fluid, Pb-17Li eutectic alloy, has the capacity to reduce the oxide layers which can be formed on austenitic and martensitic steels by means of high-temperature oxidation. It does not react with alumina in the temperature range of interest. Thus, the covering of structural material with alumina would be a solution of the problem of direct insulation of the structural material. Though several methods are known to cover steels with alumina layers, such methods do not appear to be feasible for the covering of the inner side of a large tubing system. The covering of the structural material with aluminum and the subsequent oxidation of this surface seems to open a way for the solution of this problem. Though the packing procedure of alitizing was known to offer a possibility to form surface layers rich in aluminum, the alternative method of hot-dip aluminizing was applied, since this procedure has the potential for the use in large dimensions and particularly for aluminizing inner sides of tubes

  9. 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

  10. Research and development on ceramic coatings for fusion reactor liquid blankets

    International Nuclear Information System (INIS)

    Fabrication and properties of three kinds of ceramic coatings for fusion reactor liquid blankets such as Cr2O3-SiO2, Al2O3 and Y2O3 are reviewed, focused on the activities in the University of Tokyo. A composite coating of SiO2 particles in a Cr2O3 matrix was prepared on the surface of SUS316 by the chemically densified method. It has very low tritium permeability, and it is quite effective as a tritium permeation barrier without the presence of molten Li17-Pb83 alloy. Alumina (Al2O3) coating was prepared on the surface of SUS316 by the hot-dipping and oxidation method. It showed very large corrosion-resistivity to Li17-Pb83 and very small tritium permeability. Yttria (Y2O3) coating was formed on the surface of SUS316 by the plasma-spraying method. It also has a possibility as a ceramic coating for liquid blankets if crack-free coating is made on the surface of piping materials. (author)

  11. Research and development on ceramic coatings for fusion reactor liquid blankets

    International Nuclear Information System (INIS)

    Fabrication and properties of three kinds of ceramic coatings for fusion reactor liquid blankets such as Cr2O3-SiO2, Al2O3 and Y2O3 are reviewed, focusing on the activities in the University of Tokyo. A composite coating of SiO2 particles in a Cr2O3 matrix was prepared on the surface of SUS316 by the chemically densified method. It has a very low tritium permeability and is quite effective as a tritium permeation barrier without the presence of the molten Li17-Pb83 alloy. An alumina (Al2O3) coating was prepared on the surface of SUS316 by the hot-dipping and oxidation method. It showed a very high corrosion-resistance to Li17-Pb83 and a very low tritium permeability. An yttria (Y2O3) coating was formed on the surface of SUS316 by the plasma-spraying method. The product has a poor compatibility with liquid lithium. However, since sintered Y2O3 is more resistant to degradation than plasma sprayed Y2O3, it may be possible to use Y2O3 as a ceramic coating for liquid blankets if crackfree coating is made on the surface of piping materials. (orig.)

  12. Development of a direct insulation layer for a self-cooled liquid metal fusion reactor blanket

    International Nuclear Information System (INIS)

    The magneto-hydrodynamic (MHD) pressure drop in a self-cooled liquid metal fusion reactor blanket can be significantly reduced by means of electrical insulation of the flowing liquid metal against the structural material. Since alumina is chemically stable against Pb-17Li at up to 500 C, direct insulation of MANET steel was performed by aluminizing the steel, using a hot-dip process followed by high-temperature oxidation. The aluminide layer formed in this process was insulating and stable against corrosion in flowing Pb-17Li alloy at 450 C for 10000h. Several temperature changes during the corrosion tests did not cause any damage of the insulating layers. Thus, the process of hot-dip aluminizing followed by optimized oxidation at high temperature can provide a direct insulation layer to suppress MHD effects. (orig.)

  13. 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.

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

    International Nuclear Information System (INIS)

    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. 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.

  16. 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...

  17. 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. PMID:25828209

  18. 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

  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. 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

  1. 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...

  2. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

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

  3. Development of a dynamic model of a Upflow Anaerobic Sludge Blanket (UASB) reactor

    International Nuclear Information System (INIS)

    'Full text:' The Upflow Anaerobic Sludge Blanket Reactor has been modeled in this paper using a semi-empirical correlation. The main objective of the modeling exercise was to examine the dynamics (hydrodynamics and transformation) of critical variables that are important for the reactor process design, performance and operation. These factors include some microbiological variables such as soluble (as COD) and biodegradable organics (as BOD), suspended solids (SS), and biomass concentration (as VSS). The hydraulic variables such as the upflow velocity of the liquid and by extension, upflow velocity of the solids and the volumetric fractions of the liquid (voidage) are also given considerations. The dynamics of external mass transfer of organics and inorganics and the transformation reactions within the reactor are modeled by five non-linear partial differential equations (PDE's), which define dependent variables in one-dimensional temporal and spatial planes, one integrand equation for the gas rate of flow and six algebraic equations that define the reactor performance (conversion rates and removal efficiencies). The PDE's are solved using backward finite differencing method. The discretized equations are computed using a simulation program that was written in the Borland C++ programming language. The model defines the dynamics of the UASB in a simplistic way but which, nonetheless, can be useful for engineering designs where the interest is in qualitative results and not quantitative outcomes. The simulation was validated with data derived from the operation of a pilot model UASB. The pilot model was operated using raw domestic wastewater as substrate, which had an average total COD of 492 mg/L, soluble COD of 172 mg/L, BOD of 220 mg/L, suspended solid of 0.235 g/L and a high sulphate content of 108 mg/L. The average nutrient concentrations were, NH3-N 20 mg/L and PO4 15 mg/L. The UASB was operated at HRT's of 10.968, 8.326 and 5.51 hours The average removal

  4. 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

  5. Materials for breeding blankets

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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.

  8. 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...

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

    International Nuclear Information System (INIS)

    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

  11. 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...

  12. 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...

  13. 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.

  14. Design of test blanket system for ITER module testing

    International Nuclear Information System (INIS)

    Test blanket systems to be installed in ITER for developing demo blankets have been investigated. One of the main engineering goals of ITER is to test tritium breeding blankets relevant to a power reactor. The test foreseen on modules include the demonstration of a breeding capability that would lead to tritium self-sufficiency in a reactor and extraction of a high grade heat suitable for electricity generation. To accomplish these goals, several ITER equatorial ports are available to test the test blanket systems, both in the basic performance phase (BPP) and the enhanced performance phase (EPP). Test blanket systems for water-cooled and helium-cooled type DEMO blankets with ceramic breeders, developed in Japan have been designed. The design activities include the neutronics, thermal and hydraulic analyses, and mechanical configuration considering port sharing, cooling systems and tritium recovery systems, and test blanket system compatible with the current ITER design has been developed. (author)

  15. 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.

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

    International Nuclear Information System (INIS)

    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

  17. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

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

  18. 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

  19. 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...

  20. Divertor and gas blanket impurity control study

    International Nuclear Information System (INIS)

    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

  1. 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...

  2. 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)

  3. 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

  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. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

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

  6. 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.

  7. The frontiers of research on fusion blanket technology

    International Nuclear Information System (INIS)

    Current topics concerning blanket technology are reviewed. In the chemical engineering/chemistry area, the qualitative and quantitative effects of mass transfer steps of tritium is important in the understanding of the behavior of bred tritium in the solid breeder blanket system. Such phenomena as adsorption, isotope exchange reactions, and water formation reaction at the grain surface produce profound effects on the behavior of the bred tritium in the blanket. Regarding the liquid system, the physical or chemical properties of Li, Li17Pb83 and Flibe as liquid blanket materials were compared. Some recent studies were introduced regarding tritium recovery from the liquid blanket materials, impurity removal from salts, ceramic coating of structural materials, and the vapor pressure of mixtures of metals or salts. Thermal hydraulic topics in relation to several candidate power reactor concepts are summarized. Emphasis is laid on the simultaneous removal of heat and tritium from the blanket and some aspects of forming effective power cycles are developed. (author)

  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. 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.)

  11. Breeding blanket concepts for fusion and materials requirements

    International Nuclear Information System (INIS)

    This paper summarizes the design and performances of recent breeding blanket concepts and identifies the key material issues associated with them. An assessment of different classes of concepts is carried out by balancing out the potential performance of the concepts with the risk associated with the required material development. Finally, an example strategy for blanket development is discussed

  12. Analysis of behind the armor ballistic trauma.

    Science.gov (United States)

    Wen, Yaoke; Xu, Cheng; Wang, Shu; Batra, R C

    2015-05-01

    The impact response of body armor composed of a ceramic plate with an ultrahigh molecular weight polyethylene (UHMWPE) fiber-reinforced composite and layers of UHMWPE fibers shielding a block of ballistic gelatin has been experimentally and numerically analyzed. It is a surrogate model for studying injuries to human torso caused by a bullet striking body protection armor placed on a person. Photographs taken with a high speed camera are used to determine deformations of the armor and the gelatin. The maximum depth of the temporary cavity formed in the ballistic gelatin and the peak pressure 40mm behind the center of the gelatin front face contacting the armor are found to be, respectively, ~34mm and ~15MPa. The Johnson-Holmquist material model has been used to simulate deformations and failure of the ceramic. The UHMWPE fiber-reinforced composite and the UHMWPE fiber layers are modeled as linear elastic orthotropic materials. The gelatin is modeled as a strain-rate dependent hyperelastic material. Values of material parameters are taken from the open literature. The computed evolution of the temporary cavity formed in the gelatin is found to qualitatively agree with that seen in experiments. Furthermore, the computed time histories of the average pressure at four points in the gelatin agree with the corresponding experimentally measured ones. The maximum pressure at a point and the depth of the temporary cavity formed in the gelatin can be taken as measures of the severity of the bodily injury caused by the impact; e.g. see the United States National Institute of Justice standard 0101.06-Ballistic Resistance of Body Armor. PMID:25676500

  13. 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.

  14. Experience of neutronic evaluation for in-pile tests of fusion blanket with the JMTR. Influence of impurities in Beryllium

    International Nuclear Information System (INIS)

    Impurities in beryllium for fusion blanket application such as armor of the first wall and neutron multiplier is one of important issues to perform its properties. In this study, tritium production rate evaluation was carried out on the base of JMTR irradiation test that simulated the breeding blanket and the impact of impurity on tritium production rate was discussed. Influence of impurities in S200F beryllium was therefore less than 1%, concerning tritium production rate. Accordingly, it is reasonable to estimate extremely small impact on tritium production rate in case of fusion application that uses high purity beryllium such as S65C. (author)

  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. 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.

  17. Strategy for solving a coupled problem of the electromagnetic load analysis and design optimization for local conducting structures to support the ITER blanket development

    International Nuclear Information System (INIS)

    Highlights: • We present the way of modeling transient electro-magnetic loads on local conductive domains in the large magnetic system. • Simplification is achieved by decomposing of the problem, multi-scale integral-differential modeling and use of integral parameters. • The intrinsic scale of loads on a localized conductor with eddy is quantified through the load susceptibility tensor. • Solution is searched as response of a simple equivalent dynamic simulator, using control theory methods. • The concept is exemplified with multi-scenario assessment of EM eddy loads on ITER blanket modules. - Abstract: The complexity of the electromagnetic (EM) response of the tokamak structures is one of the key and design-driving issues for the ITER. We consider the specifics of the assessment of ponderomotive forces, acting on local components of a large electro-physical device during electromagnetic transients. A strategy and approach is proposed for the operative EM loads modeling and analysis that enables design optimization at early phases of development. The paper describes a method of principal simplification of the mathematical model, based on the analysis and exploiting specific features and peculiarities of the relevant technical problem, determined by the design and operation of the device and system under consideration. The application of the method for predictive EM loads analysis and corresponding numerical calculations are exemplified for the localized ITER blanket components — shield modules. The example demonstrates the efficiency of EM load analysis in complex electromagnetic systems via a set of simplified models with different scope, contents and level of detail

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

    International Nuclear Information System (INIS)

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

  20. 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 ...

  1. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

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

  2. 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)

  3. 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.

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. 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.

  12. 76 FR 69764 - Stab-Resistant Body Armor Standard Workshop

    Science.gov (United States)

    2011-11-09

    ... of Justice Programs Stab-Resistant Body Armor Standard Workshop AGENCY: National Institute of Justice... Standards and Technology (NIST) are jointly hosting a workshop focused on the NIJ Stab-resistant Body Armor.... This workshop is being held specifically to discuss with interested parties recent progress made...

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

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li2O) and lithium zirconate (Li2ZrO3) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers

  15. Tailorable Advanced Blanket Insulation (TABI)

    Science.gov (United States)

    Sawko, Paul M.; Goldstein, Howard E.

    1987-01-01

    Single layer and multilayer insulating blankets for high-temperature service fabricated without sewing. TABI woven fabric made of aluminoborosilicate. Triangular-cross-section flutes of core filled with silica batting. Flexible blanket formed into curved shapes, providing high-temperature and high-heat-flux insulation.

  16. 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.)

  17. Achievements in the development of the Water Cooled Solid Breeder Test Blanket Module of Japan to the milestones for installation in ITER

    Science.gov (United States)

    Tsuru, Daigo; Tanigawa, Hisashi; Hirose, Takanori; Mohri, Kensuke; Seki, Yohji; Enoeda, Mikio; Ezato, Koichiro; Suzuki, Satoshi; Nishi, Hiroshi; Akiba, Masato

    2009-06-01

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows the recent achievements towards the milestones of ITER TBMs prior to the installation, which consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, targeting the detailed design final report in 2012, structure designs of the WCSB TBM and the interfacing components (common frame and backside shielding) that are placed in a test port of ITER and the layout of the cooling system are presented. As for the module qualification, a real-scale first wall mock-up fabricated by using the hot isostatic pressing method by structural material of reduced activation martensitic ferritic steel, F82H, and flow and irradiation test of the mock-up are presented. As for safety milestones, the contents of the preliminary safety report in 2008 consisting of source term identification, failure mode and effect analysis (FMEA) and identification of postulated initiating events (PIEs) and safety analyses are presented.

  18. Achievements in the development of the Water Cooled Solid Breeder Test Blanket Module of Japan to the milestones for installation in ITER

    International Nuclear Information System (INIS)

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows the recent achievements towards the milestones of ITER TBMs prior to the installation, which consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, targeting the detailed design final report in 2012, structure designs of the WCSB TBM and the interfacing components (common frame and backside shielding) that are placed in a test port of ITER and the layout of the cooling system are presented. As for the module qualification, a real-scale first wall mock-up fabricated by using the hot isostatic pressing method by structural material of reduced activation martensitic ferritic steel, F82H, and flow and irradiation test of the mock-up are presented. As for safety milestones, the contents of the preliminary safety report in 2008 consisting of source term identification, failure mode and effect analysis (FMEA) and identification of postulated initiating events (PIEs) and safety analyses are presented.

  19. 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...

  20. 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.

  1. 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

  2. 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)

  3. The Los Alamos accelerator driven transmutation of nuclear waste (ATW) concept development of the ATW target/blanket system

    International Nuclear Information System (INIS)

    The studies carried out in the frame of the Accelerator Driven Transmutation Technology (ADTT) program developed at Los Alamos in order to solve the nuclear waste problem and to build a new generation of safer and non-proliferant nuclear power plants, are presented

  4. 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.)

  5. 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

  6. 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)

  7. Preliminary Safety Analysis of Korea HCSB Test Blanket Module

    International Nuclear Information System (INIS)

    A Helium Cooled Solid Breeder (HCSB) blanket has been considered as one of the promising blanket for the fusion power demonstration plant. Therefore HCSB Test Blanket Module (TBM) testing in ITER is the most important milestone for the development of the blanket of the DEMO plant. Korea has developed the HCSB TBM with some features such as graphite reflector and simplified flow passage. The objective of this study was to evaluate the thermal and structural integrity of the HCSB TBM under the hypothetical accidental conditions such as cooling pipe break in TBM. The safety analysis was performed under conservative conditions based on the TBM design, which can be assumed by the similarity of the safety analysis of the ITER shielding blanket. Transient analysis model was used to calculate the temperature distribution for Loss of Coolant Accident (LOCA). Simplified analysis conditions were a) simultaneous plasma shutdown and LOCA b) LOCA and then after FW temperature reaches 1150 deg. plasma shutdown. Helium circuit behavior during the different LOCA scenarios was also evaluated. Finally the design modifications based on the analysis result and the related R-and-D of the HCSB blanket design for the application in a DEMO reactor were mentioned. (author)

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

    International Nuclear Information System (INIS)

    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. (author)

  9. Thermal hydraulics and mechanics research on fusion blanket system

    International Nuclear Information System (INIS)

    In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding / neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding / n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R and Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics. (author)

  10. 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

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

    Science.gov (United States)

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

    2014-09-01

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

  12. Nondestructive characterization of UHMWPE armor materials

    International Nuclear Information System (INIS)

    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.

  13. 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.

  14. Nondestructive characterization of UHMWPE armor materials

    Science.gov (United States)

    Chiou, Chien-Ping; Margetan, Frank J.; Barnard, Daniel J.; Hsu, David K.; Jensen, Terrence; Eisenmann, David

    2012-05-01

    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.

  15. 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)

  16. Design and thermal-hydraulic analysis of tokamak divertor armor tiles

    International Nuclear Information System (INIS)

    A prototype divertor armor tile design has been investigated using water-cooled ATJ graphite tiles fitted to a copper heat sink. Two-dimensional steady-state and 1-D time dependent heat transfer codes were developed to determine thermal design characteristics. A steady-state heat flux of 5 MW/m2 and a transient disruption load of 140 MJ/m2 over 100 micros were assumed for an ITER-type device operating in a radiative divertor configuration. For a tile fitted to the heat sink by a bonded-pin mechanism, the optimal armor thickness was determined to be 1.0 cm, with a 2.2 cm diameter coolant channel. The maximum steady state and disruption temperatures of the tile were determined to be 1,760 K and 4,800 K, respectively. LOCA analysis yielded that a 7 second response time would be needed after loss-of-coolant in the armor tile. The design is predicted to survive approximately 6 disruptions before tile replacement would be necessary

  17. 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

  18. Survey on experimental neutron physics of CTR blankets in the KFA

    International Nuclear Information System (INIS)

    The research program of the KFA on neutronic CTR-blanket problems is described by examples of experiments on blanket-model assemblies. The work is in strong correlation with theoretical evaluations of the problems. The main objective is the experimental test of nuclear data and calculational methods by comparison of the results. The properties of the experimental blanket models used are derived from the existing CTR conceptual design studies and from the theory of neutron fields. The limitations of the present blanket experiments are discussed. The development of appropriate measuring techniques is also described

  19. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

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

  20. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

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

  1. 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

  2. Recent MHD activities for blanket analysis at UPC

    International Nuclear Information System (INIS)

    In the frame of fusion reactor design definition, the detailed analysis of main flow parameters in liquid metal blankets is of utmost interest. Critical aspects are (1) tritium inventories and permeation rates, (2) heat extraction and maximum temperatures for material specifications and (3) MHD pressure drops. The aim of GREENER research group at UPC is to develop a CFD code, based on the OpenFOAM toolbox, able to deal with the main phenomena occurring at blanket channels (MHD coupling, heat transfer and tritium transport) and capable to quantify the above mentioned critical aspects. In parallel, CIMNE research group is developing its own MHD code, mainly focused on algorithm optimisation. The paper summarises the developing tools at each research group and compares their behaviour in a validation step using analytical solutions. In order to expose the applicability of the codes, some simulation results related with the HCLL-ITER/TBM blanket are exposed. Special focus is made on buoyancy flows in U-shaped channels and multi channel effect. Moreover, a preliminary flow analysis related with vertical banana-shape liquid metal channels is discussed, related with a new blanket design that is being considered as a progress of conceptual design refinement of dual-coolant liquid metal blankets (DEMO specifications).

  3. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

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

  4. A European proposal for an ITER water-cooled solid breeder blanket

    International Nuclear Information System (INIS)

    The water-cooled solid breeder blanket concept proposed here aims to replace the shielding blanket for the enhanced performance phase of the international thermonuclear experimental reactor (ITER). The nominal performances are as follows: an average neutron wall load of 1 MW m-2 which corresponds to a fusion power of about 1.5 GW, and an average neutron fluence of 1 MWy m-2. The proposed blanket concept has been designed to accept a power increase of about 30% and power transients up to 3-5 GW for a short time. This blanket concept is based on a breeder inside tube (BIT)-type blanket with poloidal breeding elements made of 316 L-type stainless steel and filled with lithium metazirconate and beryllium pebbles. Inlet and outlet water temperatures of 160 and 200 C have been considered with a medium-pressure cooling system during plasma burn. The diameters of the breeding elements are compatible with the space available in test fission reactor core channels, making in-pile testing, required for blanket development and qualification, easier. A conservative approach using qualified materials, a blanket concept easily testable in fission reactors and on-going mock-up testing, which can be qualified using the blanket test module during the basic performance phase of ITER, will allow the blanket reliability required for the enhanced performance phase to be achieved. (orig.)

  5. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    In nuclear fusion DEMO reactor, the blanket is required to provide the tritium breeding ratio (TBR) of more than unity by the neutron induced reaction in lithium in the blanket. To provide the TBR of more than unity is critical issue in the development of the blanket. Also in order to develop the blanket with low activation level, the evaluation of the induced activity with high precision is required by taking into account the sequential reactions induced by secondary charged particles. In order to evaluate these issues experimentally, neutronics experiments have been performed by using DT neutrons at JAERI/FNS. From the results of TBR experiment by using the mockup relevant to the DEMO blanket with multilayered structure composed of Be, Li2TiO3 and F82H, it was clarified that the TPR can be evaluated within 10 % uncertainty by using the Monte Carlo calculation. From the results of sequential reactions experiment for the test specimens simulating the cooling water pipe, it was found that the effective cross-sections due to the sequential reactions were increased in a form close to an exponential curve in the cooling water pipe with reducing the distance to the water. (author)

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

    International Nuclear Information System (INIS)

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

  7. 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...

  8. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

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

  9. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

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

  10. Fusion reactor blanket: neutronic studies in France

    International Nuclear Information System (INIS)

    The problem of effective tritium regeneration is a crucial issue for the fusion reactor, especially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty analysis. The results of these studies permit us to conclude that it is possible to expect an adequate tritium breeding ratio

  11. 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.

  12. HIP joining of Be/FMS for the Development of the ITER TBM First Wall

    International Nuclear Information System (INIS)

    The test blanket module (TBM) systems for the international thermonuclear experimental reactor (ITER) have been investigated with the aim to check on their safety, reliability and compatibility under a nuclear fusion state, i.e., tritium production and recovery, high-grade heat generation and radiation shielding. ITER participant teams are developing their own TBMs to be tested from a Day-1 operation of the ITER. Korea has also proposed a helium cooled molten lithium (HCML) and helium cooled solid breeder (HCSB) blanket. One of the main issues about the R and D on the TBM is to develop the fabrication technologies for the TBM first wall. The TBM first wall is multilayer components consisting of plasma facing armor materials and structural materials. Beryllium (Be) and ferritic/ martensitic steel (FMS) are the primary candidate alloys for the armor and structural materials of the TBM, respectively. For a successful fabrication of such complex components, the hot isostatic pressing (HIP) method has been considered as the most feasible method. In this study, Be and FMS were joined by HIP techniques, and several interlayer materials had been applied in order to manufacture high strength joints

  13. 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.

  14. 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. 

  15. EXOTIC: Development of ceramic tritium breeding materials for fusion reactor blankets. The behaviour of tritium in: lithium aluminate, lithium oxide, lithium silicates, lithium zirconates

    International Nuclear Information System (INIS)

    This report describes the results of six EXOTIC experiments comprising a total of 48 capsules. Samples of the candidate tritium breeding materials LiAlO2, Li2ZrO3, Li4SiO4, Li6Zr2O7, Li8ZrO6, Li2O and Li2SiO3 have been irradiated at different temperature levels and up to a maximum lithium burnup of about 3%. Tritium residence times of the various breeding materials have been determined from temperature transients performed during irradiation. After irradiation the tritium inventory has been determined from small samples of the various materials. From the out-of-pile tritium release experiments activation energies were determined. These activities have been performed at ECN within the framework of the European Fusion Technology Programme on Breeding Blankets. (orig.)

  16. Natural Curaua Fiber-Reinforced Composites in Multilayered Ballistic Armor

    Science.gov (United States)

    Monteiro, Sergio Neves; Louro, Luis Henrique Leme; Trindade, Willian; Elias, Carlos Nelson; Ferreira, Carlos Luiz; de Sousa Lima, Eduardo; Weber, Ricardo Pondé; Miguez Suarez, João Carlos; da Silva Figueiredo, André Ben-Hur; Pinheiro, Wagner Anacleto; da Silva, Luis Carlos; Lima, Édio Pereira

    2015-10-01

    The performance of a novel multilayered armor in which the commonly used plies of aramid fabric layer were replaced by an equal thickness layer of distinct curaua fiber-reinforced composites with epoxy or polyester matrices was assessed. The investigated armor, in addition to its polymeric layer (aramid fabric or curaua composite), was also composed of a front Al2O3 ceramic tile and backed by an aluminum alloy sheet. Ballistic impact tests were performed with actual 7.62 caliber ammunitions. Indentation in a clay witness, simulating human body behind the back layer, attested the efficacy of the curaua-reinforced composite as an armor component. The conventional aramid fabric display a similar indentation as the curaua/polyester composite but was less efficient (deeper indentation) than the curaua/epoxy composite. This advantage is shown to be significant, especially in favor of the lighter and cheaper epoxy composite reinforced with 30 vol pct of curaua fiber, as possible substitute for aramid fabric in multilayered ballistic armor for individual protection. Scanning electron microscopy revealed the mechanism associated with the curaua composite ballistic performance.

  17. Armor of cermet with metal therein increasing with depth

    Science.gov (United States)

    Wilkins, M.L.; Holt, A.C.; Cline, C.F.; Foreschner, K.E.

    1973-07-01

    The system described consists of a ceramic matrix having a gradient of fine ductile metallic particles dispersed therein in an amount of from 0.0%, commencing at the front or impact surface of the armor, to about 2 to 15% by volume along the interface to the back of the system. (auth)

  18. 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.)

  19. Nuclear Performance Analyses for HCPB Test Blanket Modules in ITER

    International Nuclear Information System (INIS)

    The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in the framework of the European Fusion Technology Programme for performance tests in ITER. The related efforts currently focus on the design optimisation of suitable Test Blanket Modules (TBM) and associated R-and-D activities. Four different HCPB TBM types are considered for addressing issues related to (i) electromagnetic transients (EM), (ii) neutronics and Tritium (NT), (iii) thermo-mechanical properties of the pebble beds (TM), and (iv) the integral performance of the blanket module (Plant Integration, PI). The lay-out of the NT and the PI modules has been entirely revised to represent the latest HCPB breeder blanket concept for fusion power reactors. A HCPB TBM consists of a steel box with an internal stiffening grid and small breeder units. The stiffening grid forms radially running open cells accommodating the breeder units (BU). The BU consists of a back plate with attached breeder canisters providing space for the breeder pebble beds. The space between the canisters and the stiffening plates is filled with Beryllium pebbles for the neutron multiplication. The latest design assumes two vertically arranged breeder containers per BU with a toroidal bed height of 10 and 24 mm, for NT and PI modules, respectively. Li4SiO4 is assumed as breeder material at 6Li enrichment levels between 40 at % (NT) and 90 at % (PI). This work is devoted to the neutronic, shielding and activation analyses performed recently for NT and PI variants of the HCPB TBM in ITER. The analyses are based on three-dimensional neutronic and activation calculations making use of a 20 degree torus sector model of ITER developed for Monte Carlo calculations with the MCNP code. The model includes a proper representation of the horizontal ITER test blanket port, the water cooled support frame with two integrated HCPB blanket test modules, the radiation shield and the port environment. Monte Carlo

  20. Investigation of neutronics for CH DEMO blanket with helium-cooled ceramics breeding concepts

    International Nuclear Information System (INIS)

    ITER TBM provides the strong support for the design, materials and technology of DEMO blanket. However, ITER TBM is quite different from a DEMO blanket in aspects of boundary conditions and neutron wall loading. It is very important to further clarify relations between ITER TBM and DEMO blanket. Neutronics of the blanket is theory basis for development of fusion reactor. In order to further identify the outline design for China ITER helium-cooled solid breeder (HCSB) test blanket module (TBM) in view of Chine DEMO goal, investigation of neutronics for a DEMO reactor with helium-cooled ceramics breeding blanket is investigated by means of three dimensional MCNP code. In this paper, the author attempts to explore the pathway from ITER TBM to DEMO blanket in view of neutronics design. (1) One-dimensional neutronics of three types of breeding blanket with 4 BZ (breeding zones) (Case 1), 2 BZ (Case 2) and 3 BZ (Case 3), respectively, are studied when neutron wall loading is assumed to be 0.78 MW/m2 on ITER and 2.64 MW/m2 on DEMO. Results show that TBR (tritium breeding ratio) of Case 1 is the smallest one that is adopted by CH ITER HCSB TBM. TBR of Case 3 is 1.43 and the largest one. Case 2 has the most simplified structure and the highest power density of 20.58 MW/m3 (ITER Wn=0.78 MW/m2), which is approach to 23.49 MW/m3 (DEMO: Wn=2.64 MW/m2), that is, TBM on ITER (Case 2) is probably used to test the characterization of DEMO blanket (Case 1). ITER TBM can approach DEMO blanket in a certain engineering parameters although ITER TBM cannot approach to DEMO blanket in overall engineering conditions. Three types of HCSB blankets are all useful according to different requirements of DEMO blanket. (2) 3D neutronics calculation is much necessary for defining tritium self-sufficiency of DEMO blanket. When Case1, Case 2, Case 3 is applied to CH HCSB DEMO blanket, TBR is 0.95, 1.04 and 1.11, respectively. In three cases, case 3 has the largest TBR more than 1.0 and is proposed

  1. Multi-dimensional neutronics analysis of the 'canister blanket' for NET

    International Nuclear Information System (INIS)

    At KfK a design of a helium-cooled ceramic breeder blanket, called 'canister blanket', has been developed for the NET fusion test reactor. In this report a detailed neutronic analysis of the 'canister blanket', based on one-, two- and three-dimensional Monte-Carlo calculations in the NET-III double null configuration, is presented. The main object refers to the three-dimensional analysis of a complete sector of the NET-reactor containing the 'canister blanket'. This concerns the poloidal distribution of the neutron wall load and the neutron fluxes at the first wall, the spatial distribution of the power density, the total power production and global effects on the tritium breeding ratio. It is shown that, in case of the 'canister blanket', a global tritium breeding ratio beyond 1.0 seems to be feasible for NET. (orig.)

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

    International Nuclear Information System (INIS)

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

  3. Compatibility problems in tritium breeding blankets

    International Nuclear Information System (INIS)

    Compatibility between tritium breeding materials (liquid or solid), neutron multiplier and structural steels is a concern for the choice of a tritium breeding blanket for NET. For solid tritium breeding blanket, it seems that the more severe compatibility problem is due to the interaction of beryllium with steel. As for the water-cooled Pb17Li blanket, the first results obtained in experimental conditions closed to the concept have evidenced lower corrosion rates than those measured in thermal convection loops

  4. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

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

  5. Range of blanket concepts from near term solutions to advanced concepts

    International Nuclear Information System (INIS)

    Breeding blankets are key components of fusion power plants and determine to a large degree their attractiveness. There is a large range of possible blanket concepts, characterized by breeding material, structural material, coolant, and geometrical arrangement. On one extreme side there are concepts requiring only a modest extrapolation of the present technology, but with limited attractiveness. On the other side there are concepts with the potential for very attractive plants, but involving a considerable development risk. Therefore, the selection of blanket concepts depends on the overall strategy for fusion power plant development. To ensure sufficient support over the long development time, it has to be shown in a credible way, that (a) there are feasible blanket concepts for fusion power plants which can be developed with a high confidence of success, and (b) the final product will be an attractive power plant which can be operated safely, with minimum impact on the environment, and at an acceptable cost of electricity. This requires an evaluation of the entire range of blanket concepts, starting from near term solutions and reaching up to very advanced designs with a potential for exceptionally high performance. Typical examples of candidate blanket concepts are compared in this study. The blanket development strategy developed in the EU is described as an example

  6. 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.

  7. Further adaptation of the European ceramic-B.I.T. blanket conceptual design to updated Demo specifications

    International Nuclear Information System (INIS)

    This paper presents the recent development studies on the adaptation of the European Ceramic Solid Breeder Inside Tube (BIT) Blanket to updated DEMO specifications. The adaptation work is in progress, since 1990, when a detailed comparison between two existing designs lead to the selection of an unique concept. The main new developments concern the separation in two parts of the inboard blanket segments at the level of the lower divertor, the consequent improvement of the blanket coverage, the simplification of maintenance operations, and finally the increased compactness of the blanket because of the inclusion of the shielding into the breeder assembly

  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. 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)

  10. 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.

  11. Lithium-cooled blankets for advanced tokamaks

    International Nuclear Information System (INIS)

    The main objective of the Tokamak Power System Studies (TPSS) at Argonne National Lab. during fiscal year 1985 was to explore innovative design concepts that have the potential for significant enhancement of the attractiveness of a tokamak-based power plant. Activities in the area of plasma engineering resulted in a reference reactor concept, which served as a model for the impurity control and first-wall/blanket/shield studies. The liquid-metal-cooled first-wall/blanket/shield design activity was centered around the vanadium alloy structure and liquid-lithium coolant leading blanket concept as identified by the Blanket Comparison and Selection Study (BCSS). A ferritic steel structure and a LiPb breeder were considered as backup options. The magnetohydrodynamics (MHD) effects associated with self-cooled liquid-metal blanket/first-wall systems are substantially reduced by the lower magnetic fields required for higher plasmas, the lower neutron wall loading resulting from reduced power output, and the smaller reactor size of the TPSS model reactor. Therefore, improved performance characteristics of self-cooled liquid-metal blanket concepts are achievable mainly because the design constraints are more relaxed compared to the BCSS guidelines. Key aspects of the designs evaluated in the current study include the following: (1) design simplicity; (2) use of the first wall as an impurity control device; (3) modular first-wall/blanket/reflector/shield construction; and (4) integrated first-wall/blanket/reflector/shield

  12. 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.

  13. 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

  14. Activation analysis of coolant water in ITER blanket and divertor

    International Nuclear Information System (INIS)

    Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. 16N and 17N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The 16N and 17N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced 17N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 1010 nuclide/cm3 and 1.05 × 1010 nuclide/cm3 of 16N, 8.93 × 106 nuclide/cm3 and 0.33 × 105 nuclide/cm3 of 17N. The decay gamma-rays from 16N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from 17N decay neutrons was much lower than that from fusion neutrons.

  15. Mechanisms of stability of armored bubbles: FY 1996 Final Report

    International Nuclear Information System (INIS)

    Theoretical and experimental studies examine how a coating, or open-quotes armor,close quotes of partially wetted solid particles can stabilize tiny bubbles against diffusion of gas into the surrounding liquid, in spite of the high capillary pressures normally associated with such bubbles. Experiments with polymethylmethacrylate (PNMA) beads and carbonated water demonstrate that armored bubbles can persist for weeks in liquid unsaturated with respect to the gas in the bubbles. This question is of concern regarding gas discharges from waste tanks at the Hanford reservation. The stresses on the solid-solid contacts between particles in such cases is large and could drive sintering of the particles into a rigid framework. Stability analysis suggests that a slightly shrunken bubble would not expel a solid particle from its armor to relieve stress and allow the bubble to shrink further. Expulsion of particles from more stressed bubbles at zero capillary pressure is energetically favored in some cases. It is not clear, however, whether this expulsion would proceed spontaneously from a small perturbation or require a large initial disturbance of the bubble. In some cases, it appears that a bubble would expel some particles and shrink, but the bubble would approach a final, stable size rather than disappear completely. This simplified analysis leaves out several factors. For instance, only one perturbation toward expelling a solid from the armor is considered; perhaps other perturbations would be more energetically favored than that tested. Other considerations (particle deformation, surface roughness, contact-angle hysteresis, and adhesion or physical bonding between adjacent particles) would make expelling solids more difficult than indicated by this theoretical study

  16. 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.)

  17. Disruption problematics in segmented-blanket concepts

    International Nuclear Information System (INIS)

    In tokamaks, the hostile operating environment originated by plasma disruption events requires that the first-wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence, there is a need to improve the safety features of the segmented-blanket design concepts in order to satisfy the disruption problematics.The present paper describes recent investigations on internal blanket reinforcement systems needed in order to improve the first-wall/blanket/shield structural design for next-step and commercial fusion reactors. Particularly in the context of SEAFP and ITER activities, representative 3-D CAD models of the inboard and outboard blanket regions and the related magnetomechanical simulations are illustrated. (orig.)

  18. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

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

  19. 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

  20. Nuclear analyses of Indian LLCB test blanket system in ITER

    International Nuclear Information System (INIS)

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no. 2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radio-active waste management, equipments maintenance and replacement strategies and nuclear safety. To predict the nuclear behaviour of LLCB test blanket module in ITER, nuclear responses like tritium production, nuclear heating, neutron fluxes and radiation damages are estimated. As a part of ITER machine, LLCB TBS has to follow certain nuclear shielding requirements i.e. shutdown dose rates should not exceed the defined limits in ITER premises (inside bio-shield ∼100 μSv/hr after 12 days cooling and outside bio-shield ∼10 μSv/hr after 1 day cooling). Hence nuclear analyses are performed to assess and optimize the shielding capability of LLCB TBS inside and outside bio-shield. To state the radio-activity level of LLCB TBS components which support the rad-waste and safety assessment, nuclear activation analyses are executed. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.1). The paper describes comprehensive nuclear performance of LLCB TBS in ITER. (author)

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    In the present study, thermal-hydraulic analyses for the blanket concept are being conducted using the Multidimensional Analysis of Reactor Safety (MARSKS) code, which has been used for the safety analysis of a pressurized water reactor. The purposes of the analyses are to verify the applicability of the code for the proposed blanket system, to investigate the departure of nucleate boiling (DNB) occurrence during the normal and transient conditions, and to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. In this paper, the thermal analysis results of the proposed blanket design using the MARS-KS code are presented for the normal operation and an accident condition of a reduced coolant flow rate. Afterwards, the plan for the whole blanket system analysis using MARSKS is introduced and the result of the first trial for the multiple blanket module analysis is summarized. In the present study, thermal-hydraulic analyses for the blanket concept were conducted using the MARS-KS code for a single blanket module. By comparing the MARS calculation results with the CFD analysis results, it was found that MARS-KS can be applied for the blanket thermal analysis with less number of computational meshes. Moreover, due to its capability on the two-phase flow analysis, it can be used for the transient or accident simulation where a phase change may be resulted in. In the future, the MARS-KS code will be applied for the anticipated transient and design based accident analyses. The investigation of the DNB occurrence during the normal and transient conditions will be of special interest of the analysis using it. After that, a methodology to simulate the entire blanket system was proposed by using the DLL version of MARS-KS. A supervisor program, which controls the multiple DLL files, was developed for the common header modelling. The program explicitly determines the flow rates of each module which can equalize

  3. 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.)

  4. 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.

  5. Morphometric structural diversity of a natural armor assembly investigated by 2D continuum strain analysis.

    Science.gov (United States)

    Varshney, Swati; Song, Juha; Li, Yaning; Boyce, Mary C; Ortiz, Christine

    2015-12-01

    Many armored fish scale assemblies use geometric heterogeneity of subunits as a design parameter to provide tailored biomechanical flexibility while maintaining protection from external penetrative threats. This study analyzes the spatially varying shape of individual ganoid scales as a structural element in a biological system, the exoskeleton of the armored fish Polypterus senegalus (bichir). X-ray microcomputed tomography is used to generate digital 3D reconstructions of the mineralized scales. Landmark-based geometric morphometrics is used to measure the geometric variation among scales and to define a set of geometric parameters to describe shape variation. A formalism using continuum mechanical strain analysis is developed to quantify the spatial geometry change of the scales and illustrate the mechanisms of shape morphing between scales. Five scale geometry variants are defined (average, anterior, tail, ventral, and pectoral fin) and their functional implications are discussed in terms of the interscale mobility mechanisms that enable flexibility within the exoskeleton. The results suggest that shape variation in materials design, inspired by structural biological materials, can allow for tunable behavior in flexible composites made of segmented scale assemblies to achieve enhanced user mobility, custom fit, and flexibility around joints for a variety of protective applications. PMID:26481418

  6. 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.

  7. 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.

  8. 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.

  9. ITER reference breeding blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, M. [ENEA, Frascati (Italy); Bianchi, A. [EFET, Ansaldo Ricerche, Genova (Italy); Celentano, G. [ENEA, ERG-FUS, Centro di Frascati, Via Enrico Fermi, 27, P.O. Box 65, I-00044, Frascati (IT)] [and others

    1999-11-01

    The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions. (orig.)

  10. ITER reference breeding blanket design

    International Nuclear Information System (INIS)

    The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions. (orig.)

  11. 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

  12. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m2) the average neutron power load on the first wall is below 1 MWm.2, which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  13. 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

  14. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

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

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. 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.

  2. Study of the tritium production in a 1-D blanket model with Monte Carlo methods

    OpenAIRE

    Cubí Ricart, Álvaro

    2015-01-01

    In this work a method to collapse a 3D geometry into a mono dimensional model of a fusion reactor blanket is developed and tested. Using this model, neutron and photon uxes and its energy deposition will be obtained with a Monte Carlo code. This results will allow to calculate the TBR and the thermal power of the blanket and will be able to be integrated in the AINA code.

  3. 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...

  4. 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....

  5. 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)

  6. Parasitoids and predators of armored scales in some orchards in Serbia

    OpenAIRE

    Draga Graora; Radoslava Spasić; Anđa Vučetić

    2009-01-01

    Five parasitoid and five predator species were registered on four armored scale species in apple, pear, peach and plum orchards at 16 sites in Serbia. The armored scales found were: Diaspidiotus perniciosus (Comstock), Pseudaulacaspis pentagona (Targioni-Tozzetti), Epidiaspis leperii (Signoret) and Lepidosaphes ulmi (L.). The parasitoid species detected were: Encarsia (=Prospaltella) perniciosi (Tow.), Encarsia (=Prospaltella) berlesei (How.), Aphytis proclia (Wal.), Aphytis mytilaspidis (Le ...

  7. 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。

  8. Markov Blanket Ranking using Kernel-based Conditional Dependence Measures

    OpenAIRE

    Strobl, Eric V.; Visweswaran, Shyam

    2014-01-01

    Developing feature selection algorithms that move beyond a pure correlational to a more causal analysis of observational data is an important problem in the sciences. Several algorithms attempt to do so by discovering the Markov blanket of a target, but they all contain a forward selection step which variables must pass in order to be included in the conditioning set. As a result, these algorithms may not consider all possible conditional multivariate combinations. We improve on this limitati...

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. Laser-induced micro-jetting from armored droplets

    Science.gov (United States)

    Marston, J. O.; Thoroddsen, S. T.

    2015-07-01

    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.

  14. Performance analysis of a full graphite armor for RFX

    International Nuclear Information System (INIS)

    The paper firstly summarises the reasons that led to the choice of covering the whole RFX vacuum vessel with solid graphite tiles, and gives a short description of the armor design. The thermal behaviour of the first wall under heat fluxes coming from plasma particle transport is then analysed; 1-D and 2-D finite element models give accurate predictions of the temperature within the tiles during the pulses and on the full wall during the dead times. The radiative and conductive contribution to the heat exchanges are analysed and compared. Finally, eddy currents and forces acting on the tiles and their supports during fast plasma current terminations are calculated

  15. Performance analysis of a full graphite armor for RFX

    International Nuclear Information System (INIS)

    A summary is presented of the reasons that led to the choice of covering the whole RFX vacuum vessel with solid graphite tiles, and a short description is given of the armor design. The thermal behaviour of the first wall under heat fluxes coming from plasma particle transport is analysed; 1-D and 2-D finite element models give accurate predictions of the temperature within the tiles during the pulses and on the full wall during the dead times. The radiative and conductive contribution to the heat exchanges are analysed and compared. Finally, eddy currents and forces acting on the tiles and their supports during fast plasma current terminations are calculated. (author)

  16. Integral experiments for verification of tritium production on the beryllium/lithium titanate blanket mock-up with a one-breeder layer

    International Nuclear Information System (INIS)

    The first series of integral experiments on the blanket mock-up with a one breeder layer was performed in support of the concept of the solid breeding blanket cooled with water, proposed by JAERI for application in the DEMO reactor. The mock-up for the first series of experiments was designed to be as simple as possible within the proposed blanket concept. Key objectives of the experiments were: to check how correctly the tritium production rate can be predicted in the breeder layer closest to the first wall, since this particular location is greatly affected by changes of incoming neutron spectra; to validate the modified experimental techniques for measurements of tritium production rate in conditions of quick gradient thermal neutron field inside the lithium titanate layer. The mock-up contains F82H, lithium titanate and beryllium layers, with respective thicknesses of 16 mm, 12 mm and 203 mm. An additional tungsten layer was installed in front of the first layer in order to simulate armor material. The mock-up, being placed inside the SS316 cylindrical enclosure, is shaped as a pseudo-cylindrical slab with an area-equivalent diameter of 628 mm. Integral experiments on the blanket mock-up irradiated by neutrons from the D-T source with and without the source reflector were executed. A detailed description of experimental results and an example of calculation analysis are presented. (author)

  17. Design study on water- and gas-cooled outboard shield blankets for NET

    International Nuclear Information System (INIS)

    The Karlsruhe Nuclear Research Center entered into an agreement with the Commission of the European Communities on execution of development work geared to shielding blankets for NET. The concept to be investigated concerned water-cooled shielding blankets and, as a backup solution, a variation with helium cooling. The NET standard version as of late 1985 was considered as the basis of the investigations. The results of the study prepared in cooperation with the Sulzer company, Winterthur, and relating to the outboard blanket are contained in this report, which shows that it is relatively easy to fabricate water-cooled blankets. The stresses acting on the material during operation as a result of temperature gradients and coolant pressure are low. By addition of lithium salts to the coolant a great potential of tritium generation is offered. On the other hand, helium cooling is associated with some difficulties in design and with higher expenditure in fabrication. However, these difficulties can probably be overcome. (orig.)

  18. 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)

  19. Calculation of the possible tritium production in irradiation positions of the FRJ-2(DIDO) for fusion blanket experiments

    International Nuclear Information System (INIS)

    In the field of tritium and fusion blanket technology possibly an important and early contribution to the development of a fusion reactor blanket can be obtained by irradiation experiments at the research reactor FRJ-2 in Juelich, Federal Republic of Germany. However, the tritium production rate of 0.2 x 1013 to 2 x 1013 cm-3s-1 and the power per volume of 2 to 20 W cm-3 characteristic for a fusion reactor blanket have to be realized. The present report shows the reachable tritium values calculated for different irradiation positions in the FRJ-2 for natural lithium as a breeder material considering the actual existing neutron spectrum. Based on these results we come to the conclusion that the specified blanket data can actually be reached and adjusted. Therefore irradiation experiments at the FRJ-2 would be able to supply basical results for the fusion blanket development. (orig.)

  20. 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.

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

    International Nuclear Information System (INIS)

    In the frame of the recent CEA studies aiming at the development, evaluation and comparison of solid breeder blanket concepts in view of their adaptation to NET, and the evaluation of specific questions related to the first wall design, the present paper examines first the performances of a helium cooled toroidal blanket design for NET, based on innovative Beryllium/Ceramics breeder rod elements. Neutronic and thermo-mechanical optimisation converges on a concept featured by a breeding capability in excess of 1.2, a reasonnable pumping power of 1 % and a narrow breeder temperature range (470 +- 30 0C of the breeder), the latter being largely independent of the power level. The final section of the paper is devoted to the evaluation of the heat load poloidal distribution and to the irradiation effects on first wall structural materials

  2. 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.

  3. 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

  4. 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

  5. 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

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

    International Nuclear Information System (INIS)

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

  7. Blanket management method for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    A method for reducing thermal striping in liquid metal fast breeder reactors by reducing temperature gradients between adjacent fuel and blanket assemblies by shuffling blanket assemblies at each refueling outage so as to progressively shuffle the blanket assemblies to the core periphery through multiple moves and to generally locate fresh blanket assemblies adjacent to exposed fuel assemblies and exposed blanket assemblies adjacent to fresh fuel. Additionally, assembly orificing is altered to provide less flow to blanket assemblies needing less flow due to an otherwise decreased temperature gradient and providing additional flow to fuel assemblies which need more flow to sufficiently reduce temperature gradients to prevent thermal striping. (author)

  8. 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

  9. Solid breeder blanket design and tritium breeding

    International Nuclear Information System (INIS)

    Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements - to breed one tritium per fusion neutron - are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li2O, LiAlO2, Li2ZrO3, Li4SiO4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the amount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket. (orig.)

  10. 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.

  11. In plain sight: the Chesapeake Bay crater ejecta blanket

    Directory of Open Access Journals (Sweden)

    D. L. Griscom

    2012-02-01

    during reentry.

    An idealized calculation of the CBIS ejecta-blanket elevation profile minutes after the impact was carried out founded on well established rules for explosion and impact-generated craters. This profile is shown here to match the volume of the upland deposits ≥170 km from the crater center. Closer to the crater, much of the "postdicted" ejecta blanket has clearly been removed by erosion. Nevertheless the Shirley and fossil-free Bacons Castle Formations, located between the upland deposits and the CBIS interior and veneering the present day surface with units ∼10–20 m deep, are respectively identified as curtain- and excavation-phase ejecta. The neritic-fossil-bearing Calvert Formation external to the crater is deduced to be of Eocene age (as opposed to early Miocene as currently believed, preserved by the armoring effects of the overlying CBIS ejecta composed of the (distal upland deposits and the (proximal Bacons Castle Formation. The lithofacies of the in-crater Calvert Formation can only have resulted from inward mass wasting of the postdicted ejecta blanket, vestiges of which (i.e. the Bacons Castle and Shirley Formations still overlap the crater rim and sag into its interior, consistent with this expectation.

    Because there appear to be a total of ∼10 000 km2 of CBIS ejecta lying on the present-day surface, future research should center the stratigraphic, lithologic, and petrologic properties of these ejecta versus both radial distance from the crater center (to identify ejecta from different ejection stages and circumferentially at fixed radial distances (to detect possible anisotropies relating the impact angle and direction of approach of the impactor.

    The geological units described here may comprise the best preserved, and certainly the most accessible, ejecta blanket of a major crater on the Earth's surface and therefore promise to be a boon to the field of impact geology. As a corollary, a major

  12. 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.

  13. 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...

  14. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

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

  15. 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.

  16. Development of discrete ordinates code supporting unstructured tetrahedral mesh and applied in neutronics analysis for the Korea Helium Cooled Ceramic Reflector Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Woon, E-mail: jwkim@kaeri.re.kr [Korea Atomic Energy Research Institute, 989 Daeduck-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, Cheol Woo; Lee, Young-Ouk; Lee, Dong-Won [Korea Atomic Energy Research Institute, 989 Daeduck-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Gwahangno, Yuseong-gu, Daejeon 305-333 (Korea, Republic of)

    2014-10-15

    Highlights: • We developed discrete ordinates transport solver which uses discontinuous finite element method (DFEM) as a spatial discretization to deal with an unstructured tetrahedral mesh. • For pre- and post-processing, Gmsh is used to generate unstructured tetrahedral mesh by importing CAD file (*.step) and visualize the calculation results. • We applied this code to a TBM neutronics analysis, the neutron flux distribution in the Korea HCCR TBM is compared with that of MCNPX, and visualized in a three-dimensional system domain. • Calculated total flux averaged over multilayer zone shows good agreement with that of MCNPX. - Abstract: The discrete ordinates code under development by KAERI uses an unstructured tetrahedral mesh, and thus it can be applied to solve the radiation transport in a complicated geometry. In addition, the geometry modeling process has become much easier because computational tetrahedral meshes are generated based on the CAD file by Gmsh. As our first phase of applying the code to a TBM neutronics analysis, the neutron flux distribution in the Korea HCCR TBM is compared with that of MCNPX, and visualized in a three-dimensional system domain. Visualization of the fluxes and associated reaction rates in the whole system with a single run is one of the merits of a deterministic method and is very useful for checking hot spots.

  17. Development of discrete ordinates code supporting unstructured tetrahedral mesh and applied in neutronics analysis for the Korea Helium Cooled Ceramic Reflector Test Blanket Module

    International Nuclear Information System (INIS)

    Highlights: • We developed discrete ordinates transport solver which uses discontinuous finite element method (DFEM) as a spatial discretization to deal with an unstructured tetrahedral mesh. • For pre- and post-processing, Gmsh is used to generate unstructured tetrahedral mesh by importing CAD file (*.step) and visualize the calculation results. • We applied this code to a TBM neutronics analysis, the neutron flux distribution in the Korea HCCR TBM is compared with that of MCNPX, and visualized in a three-dimensional system domain. • Calculated total flux averaged over multilayer zone shows good agreement with that of MCNPX. - Abstract: The discrete ordinates code under development by KAERI uses an unstructured tetrahedral mesh, and thus it can be applied to solve the radiation transport in a complicated geometry. In addition, the geometry modeling process has become much easier because computational tetrahedral meshes are generated based on the CAD file by Gmsh. As our first phase of applying the code to a TBM neutronics analysis, the neutron flux distribution in the Korea HCCR TBM is compared with that of MCNPX, and visualized in a three-dimensional system domain. Visualization of the fluxes and associated reaction rates in the whole system with a single run is one of the merits of a deterministic method and is very useful for checking hot spots

  18. Safety and environmental impact of the dual coolant blanket concept. SEAL subtask 6.2, final report

    International Nuclear Information System (INIS)

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four concepts under development, namely two of the solid breeder type and two of the liquid breeder type. At the Forschungszentrum Karlsruhe one blanket concept of each line has been pursued so far with the so-called dual coolant type representing the liquid breeder line. In the dual coolant concept the breeder material (Pb-17Li) is circulated to external heat exchangers to carry away the bulk of the generated heat and to extract the tritium. Additionally, the heavily loaded first wall is cooled by high pressure helium gas. The safety and environmental impact of the dual coolant blanket concept has been assessed as part of the blanket concept selection excercise, a European concerted action, aiming at selecting the two most promising concepts for futher development. The topics investigated are: (a) Blanket materials and toxic materials inventory, (b) energy sources for mobilisation, (c) fault tolerance, (d) tritium and activation products release, and (e) waste generation and management. No insurmountable safety problems have been identified for the dual coolant blanket. The results of the assessment are described in this report. The information collected is also intended to serve as input to the EU 'Safety and Environmental Assessment of Fusion longterm Programme' (SEAL). The unresolved issues pertaining to the dual coolant blanket which would need further investigations in future programmes are outlined herein. (orig.)

  19. 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.

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

    International Nuclear Information System (INIS)

    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 Li2TiO3 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

  1. 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.

  2. 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

  3. MHD issues related to the use of Lithium Lead eutectic as breeder material for blankets of fusion power plants

    OpenAIRE

    Aiello, G.; Mistrangelo, C.; Buehler, L; Mas de les Valls Ortiz, Elisabet; Aubert, J. -J.; Li-Puma, A.; Rapisarda, David; Del Nevo, A.

    2015-01-01

    The European Community is committed to the development of a DEMOnstration fusion power plant whose operation could start as soon as 2050. The blanket is one of the most critical components in a fusion reactor; three of the four blanket concepts currently under development are based on the use of the liquid eutectic alloy Pb-15.7Li. Since the blanket will operate under the strong magnetic eld used to con ne the plasma, electromagnetic forces will occur in the PbLi ow, giving rise to magnetohy...

  4. 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.

  5. Polythioether Particles Armored with Modifiable Graphene Oxide Nanosheets.

    Science.gov (United States)

    Rodier, Bradley J; Mosher, Eric P; Burton, Spencer T; Matthews, Rachael; Pentzer, Emily

    2016-06-01

    Facile and scalable fabrication methods are attractive to prepare materials for diverse applications. Herein, a method is presented to prepare cross-linked polymeric nanoparticles with graphene oxide (GO) nanosheets covalently attached to the surface. Alkene-modified GO serves as a surfactant in a miniemulsion polymerization, and the alkene functionalities of GO exposed to the oil-phase are incorporated into the polymer particle through thiol-ene reactions, leaving the unreacted alkene functional groups of the other face of GO available for further functionalization. The surface of GO-armored polymer particles is then modified with a small molecule fluorophore or carboxylic acid functional groups that bind to Fe2 O3 and TiO2 nanoparticles. This methodology provides a facile route to preparing complex hybrid composite materials. PMID:27076068

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

    DEFF Research Database (Denmark)

    Moghim, M. N.; Andersen, Thomas Lykke

    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 the...... 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 of the...... simple method based on measured data of Rec and hf is better than the presented stability equation based on wave momentum flux parameter. When using the best prediction formulae for recession and depth of intersection the simple method and the momentum flux stability equation provide similar uncertainty....

  7. 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

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

    International Nuclear Information System (INIS)

    Two breeding blanket are presently developed in Europe for the DEMO reactor: the first one, the Helium Cooled Lithium Lead (HCLL) uses a liquid breeder while the other , the Helium Cooled Pebble Bed (HCPB), uses a solid breeder in form of pebble bed. The modules of these blankets, called Test Blanket Modules (TBM) will be located in correspondence of the equatorial ports of ITER in order to be tested. ITER FEAT was designed with shielding blankets, therefore in the final stage of the experiment, in the foreseen tritium -deuterium operation phase, the tritium will be supplied to the reactor and not produced inside it. Since the production of tritium is of main importance for the feasibility of a nuclear fusion reactor, perhaps in the ITER final stage, the shielding blanket could be substituted by means of a breeding blanket. The geometry and composition of this breeding blanket would be, of course, similar to that of TBM which demonstrated to have the best performances. This paper illustrates a neutronic and thermal analysis of an hypothetical triziogen blanket for ITER FEAT made similar to a HCPB test module. The main aims of the performed analyses are to determine the Tritium Breeding Ratio (TBR) considering different solid breeders (Li4SiO4 and Li2TiO3) with different enrichment in 6Li and different structural materials (a 9%CRWVTa reduced activation ferritic martensitic steel (EUROFER) or ceramic matrix composites like SiCf/SiC). The breeding blanket design is compared considering the highest value of TBR and the verification of the temperature constraints ( 550 oC for the steel, 950 o C for the breeder and 650 oC for the Beryllium). The neutronic analyses have been performed by means of MCNP-4C code and the thermal analyses using the MSC-MARC code. A TBR about equal 1 was obtained with a SiCf/SiC structural material and a Li4SiO4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give useful

  9. 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.)

  10. 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

  11. 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...... ability of the wires. This may cause the wires to buckle in the circumferential pipe direction, when these are restrained against radial deformations by adjacent layers. In the present paper, a single armoring wire modeled as a long and slender curved beam embedded in a frictionless cylinder bent into a...

  12. 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。...

  13. 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.

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

    OpenAIRE

    Stephan Mark Spiller; Nils Rüther; Heide Friedrich

    2015-01-01

    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 obse...

  15. Tritium recovery from ceramic breeder blanket

    International Nuclear Information System (INIS)

    It is known that chemical forms of tritium released from ceramic breeders are T2O and T2. Among issues relevant to the tritium chemical form, tritium inventory is one of the major criteria in the selection of breeder material. The primary purpose of this report is to study the dependence of tritium inventory in a blanket with ceramic solid breeder on the tritium chemical form. In this light, tritium inventory in a Li2O blanket has been evaluated as a function of tritium chemical form under the conditions of the Japanese Fusion Experimental Reactor (FER). It was shown that in a blanket with Li2O as a breeder, which has a strong affinity to water vapor, the inventory due to T2O adsorption becomes quite large. In order to reduce the T2O adsorption inventory, conversion of the tritium chemical form through an isotope exchange reaction with hydrogen added to the sweep gas (T2O + 2 H2 → H2O + 2 HT) has been proposed, and its advantages and problems have been examined. Lithium hydroxide formation and mass transfer, which are considered to be inherent in the Li2O-T2O system and to be critical issues for the feasibility of a Li2O blanket, have been also discussed. (author)

  16. 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

  17. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    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.)

  18. 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

  19. 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%.

  20. Integral experiment on effects of large opening in fusion reactor blanket on tritium breeding using annular geometry

    International Nuclear Information System (INIS)

    An experiment involving a simulated blanket with an opening has been performed using the line source and annular blanket system developed under the JAERI/USDOE collaborative programme in order to examine the effects of the opening on neutronics parameters such as the tritium-breeding ratio. The annular test assembly was rectangular in shape and consisted of a lithium oxide blanket covered with graphite and SS304 which simulated the graphite armour plate and first wall in a fusion device. A large opening (376mm x 425.5mm) was made in the middle of the test blanket. This opening simulated a neutral beam injector.Tritium production rates and reaction rates were measured inside the blanket. Neutron spectra and reaction rates were also measured on the surfaces of both sides and without the opening of the inner cavity. The opening decreased the number of low energy neutrons contained in the cavity and especially decreased 6Li tritium production by 10% inside the blanket at the opposite side of the opening. The Monte Carlo code GMVP using the JENDL-3 nuclear data library predicted the measured nuclear parameters in the test blankets, such as the tritium production rate, to within 10% accuracy. (orig.)

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

    International Nuclear Information System (INIS)

    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%

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-09-01

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

  3. 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.

  4. Further neutronic analyses of the European ceramic B.I.T. blanket for Demo

    International Nuclear Information System (INIS)

    The present study concerns the most recent neutronic analyses of two design versions of the european ceramic B.I.T. blanket, jointly developed by ENEA and CEA since few years. The last year developments required a new 3-D geometry evaluations of the global TBR (Tritium Breeding Ratio). The results indicated that the ENEA version reaches a global TBR value of 1.13. The CEA version, in a 3-D model using a simplified description of the breeder module layout, reaches a TBR value of 1.12. Nuclear heat deposition density has been determined for all blanket components as a function of the poloidal co-ordinate. Shielding properties of this type of blanket have been analyzed

  5. 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

  6. 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)

  7. Design and analysis of breeding blanket with helium cooled solid breeder for ITER-TBM

    International Nuclear Information System (INIS)

    Test blanket module (TBM) is one of important components in ITER. Some of related blanket technologies of future fusion, such as tritium self-sufficiency, the exaction of high-grade heat, design criteria and safety requirements and environmental impacts, will be demonstrated in ITER-TBM. In ITER device, the three equatorial ports have allocated for TBM testing. China had proposed to develop independently the ITER-TBM with helium cooled solid breeder in 12th meeting of test blanket workgroup (TBWG-12). In this work, the preliminary design and analysis for Chinese HCSB TBM will be carried out. The TBM must be contains the function of the first wall, breeding blanket, shield and structure. Finally, in the period of preliminary investigation, HCSB TBM design adopt modularization concept which is helium as coolant and tritium purge gas, ferritic/martensitic steel as structural material, Lithium orthosilicate (Li4SiO4) as tritium breeder, beryllium pebble as neutron multiplier. TBM is allocated in standard vertical frame port. HCSB TBM consist of first wall, backplate, breeding sub-modules, caps, grid and support plate, and breeding sub-modules is arranged by layout of 2 x 6 in blanket box. In this paper, main components of HCSB TBM will be described in detail, also performance analysis of main components have been completed. (authors)

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

    International Nuclear Information System (INIS)

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

  9. 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

  10. 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

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

    Science.gov (United States)

    Chintapalli, Ravi Kiran; Mirkhalaf, Mohammad; Dastjerdi, Ahmad Khayer; Barthelat, Francois

    2014-09-01

    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. PMID:24613857

  12. 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)

  13. 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

  14. 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)

  15. 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.)

  16. 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.

  17. Novel method for sludge blanket measurements.

    Science.gov (United States)

    Schewerda, J; Förster, G; Heinrichmeier, J

    2014-01-01

    The most widely used methods for sludge blanket measurements are based on acoustic or optic principles. In operation, both methods are expensive and often maintenance-intensive. Therefore a novel, reliable and simple method for sludge blanket measurement is proposed. It is based on the differential pressure measurement in the sludge zone compared with the differential pressure in the clear water zone, so that it is possible to measure the upper and the lower sludge level in a tank. Full-scale tests of this method were done in the secondary clarifier at the waste water treatment plant in Hecklingen, Germany. The result shows a good approximation of the manually measured sludge level. PMID:24569276

  18. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

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

  19. 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 ...

  20. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R-and-D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  1. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R(and)D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  2. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Balaraman Kumar

    2010-06-01

    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 thickness direction. 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 optimising the torque for blankets from different manufacturers.

  3. 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.

  4. Neutronics Assessment of Molten Salt Breeding Blanket Design Options

    International Nuclear Information System (INIS)

    Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be ∼1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (∼3 kg) over the blanket lifetime of ∼3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied

  5. Fissile fuel breeding in DT fusion reactor blankets

    International Nuclear Information System (INIS)

    Results of neutronic evaluations of fissile fuel breeding in a variety of DT fusion hybrid-reactor blankets are presented. The blankets are of the fast-fission or fission-suppressed rather than fission-enhanced designs, i.e. in the blankets considered emphasis is on fissile fuel rather than power production. For 233U breeding, when Li metal is the coolant for the first wall and the graphite moderator and the tritium breeding constituent of the blanket, the number of atoms of 233U produced per fusion in blankets that could be of practical interest is in the range 0.5 - 0.68, with the lower value applying to water-cooled ThO2 fertile fuel, the upper to gas-cooled Th-metal fuel located next to the reactor first wall. Neutron multipliers like Pb or Be can increase the production to about 0.74. For 239Pu breeding, the production ratio in practical blankets is 0.6 - 1.64, with the best results being for gas, Na- or Li-metal-cooled U-metal fuels located adjacent to the first wall (the U is depleted uranium). Gas-cooled U-Th-metal blankets, optimized for 233U breeding, yield 0.76 atoms of 233U and 0.38 atoms of 239Pu. The blanket energy multiplication factors are in the range 1.6 - 2.5 for Th blankets, 2.5 - 9.0 for U blankets and approximately 5.5 for the U-Th-metal blanket. The tritium breeding ratio in all blankets is 1.075. Blankets with other first wall, coolant and tritium breeding constituents are also considered. The fusion power requirements of hybrids that could supply the fuel needs of thorium-burning CANDU power reactors, and the allowed costs for building the hybrids are indicated

  6. 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.

  7. First wall/blanket/shield design and optimization system

    International Nuclear Information System (INIS)

    First wall/blanket/shield design and optimization system (BSDOS) has been developed to provide a state-of-the-art design tool for fast accurate analysis. In addition, it has been designed to perform several other functions: (1) allowing comparison and evaluation studies for different concepts using the same data bases and ground rules, (2) permitting the use of any figure of merit in the evaluation studies, (3) optimizing the first wall/blanket/shield design parameters for any figure of merit under several design constraints, (4) permitting the use of different reactor parameters in the evaluation and optimization analyses, (5) allowing the use of improved eingineering data bases to study the impact on the design performance for planning future research and development, and (6) evaluating the effect of the data base uncertainties on the design performance. BSDOS is the first design and optimization system to couple the highly interacting neutronics, heat transfer, thermal hydraulics, stress analysis, radioactivity and decay-heat analyses, tritium balance, and capital cost. A brief description of the main features of BSDOS is given in this paper. Also, results from using BSDOS to perform design analysis for several reactor components are presented. 17 refs., 1 fig., 2 tabs

  8. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

  9. Flibe blanket concept for transmuting transuranic elements and long lived fission products

    International Nuclear Information System (INIS)

    A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

  10. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    International Nuclear Information System (INIS)

    Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

  11. 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.

  12. 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.

  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. 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

  17. 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.

  18. 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...

  19. 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.

  20. 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)

  1. 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.

  2. 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.

  3. Investigation on Ballistic Performance of Armor Ceramics against Long-Rod Penetration

    Science.gov (United States)

    Huang, Feng-Lei; Zhang, Lian-Sheng

    2007-12-01

    A series of depth-of-penetration (DOF) tests are carried out to investigate the ballistic performance of armor ceramics. Based on the experimental results, an improved differential efficiency factor (DEF) is presented, which demonstrates that the general ballistic efficiency index is independent of the ceramic thickness. It is also shown that the density, internal friction, and compression strength of ceramics are crucial factors that affect the ballistic performance of ceramics significantly through the interaction between the long-rod projectiles and thick-tile armor.

  4. 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

  5. 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).

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

    International Nuclear Information System (INIS)

    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 .deg. C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 .deg. 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)

  7. Heterogeneous structure effect on molten salt blanket neutronics

    Energy Technology Data Exchange (ETDEWEB)

    Grebyonkin, K.F.; Kandiev, Ya.Z.; Malyshkin, G.N.; Orlov, A.I. [Inst. of Technical Pysics, Chelyabinsk (Russian Federation). Dept. of Physics

    1997-09-01

    The report presents the results of the molten salt blanket neutronics calculations performed for researchers of a facility for accelerator-driven transmutation of long-lived radioactive wastes and plutonium conversion. Heterogeneous structure effect on molten salt blanket neutronics was studied through computation. 4 refs., 1 fig., 1 tab.

  8. 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.)

  9. Tritium confinement requirements for the water-cooled Pb-17Li blanket for demo

    International Nuclear Information System (INIS)

    In the water-cooled liquid Pb-17Li blanket concept for DEMO the limitation of tritium permeation from the breeder material into the cooling water will be required. In order to find out on what conditions this tritium permeation remains within reasonable limits, a 1-d FEM code was developed which evaluates the tritium partial pressure in Pb-17Li, the tritium inventory in the blanket material, and the tritium permeation from the Pb-17Li into the cooling water as a function of the permeation reduction factor of a barrier and the efficiency of the tritium extraction from Pb-17Li. With a parametric study the conditions were identified which allow a permeation rate of as little as 1 g.d-1 without pushing the requirements for permeation barriers and extraction efficiencies excessively far. an example is a barrier with a permeation reduction factor of 75 together with an extractor efficiency of approximately 83%. In these conditions the expected tritium inventory is attributed to approximately one third to the martensitic blanket structure (some ten grams) while two thirds will be found in the Pb-17Li. These inventory values are two orders of magnitude lower than in solid breeder blankets and are thus not considered a critical issue. 6 refs., 5 figs., 2 tabs

  10. 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.

  11. 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

  12. Fast-core thermal-blanket breeder reactor

    International Nuclear Information System (INIS)

    A preliminary assessment of the performance expected from a specific type of FCTB reactor, consisting of a gas-cooled fast system for the core and natural-uranium light-water thermal system for the blanket is reported. Both the core and the blanket use the 238U-Pu fuel cycle. When all the neutrons leaking out of the core reach the blanket, the blanket-to-core power ratio is estimated to be about 1.3. By reducing its water-to-fuel volume ratio below 1.5, the light water blanket can be designed to have a higher ksub(eff), while maintaining an equilibrium fissile fuel content. Compared with conventional FBRs, having the same power output, the FCTB reactor considered offers the following advantages: a lower fissile fuel content, easier and safer control, no need for Pu separation. (B.G.)

  13. 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

  14. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    International Nuclear Information System (INIS)

    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.

  15. 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.

  16. 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...

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

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) project began a new design phase called the Engineering Design Activity (EDA) which started in July 1992. A variety of blanket designs options were analyzed as a part of the U.S. ITER home team blanket option trade-off study (BOTS) which began in May 1993. The options considered were a self-cooled Li/V blanket, a helium cooled Li/V blanket and a water cooled 316 SS nonbreeding shield option. Detailed activation, dose rate and waste disposal rating calculations have been performed for these different ITER blanket design options based on a fluence of 3.0 MWa/m2 and an average neutron wall loading of 2.0 MW/m2. A continuous operation assumption was utilized in the analysis. The results of this work are presented in this conference

  18. 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)

  19. 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

  20. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  1. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Aqueous solution blanket using lithium salts such as LiNO3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

  2. 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

  3. 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.

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. Hybrid reactor blankets for constant energy multiplication and flat power distribution

    International Nuclear Information System (INIS)

    Two blanket design difficulties are usually attributed to the blanket neutronic properties: high peak-to-average power density ratio distribution and the variation of the energy multiplication with burnup. This work shows that blankets can be designed to have a constant energy multiplication and a flat power distribution. These features are illustrated for light water hydrid reactor blankets

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. Technologies and modelling issues for tritium processing in the European Test Blanket Systems and perspectives for DEMO

    International Nuclear Information System (INIS)

    Highlights: • Provided DEMO relevancy considerations on tritium processing technologies. • Provided updates on the main technologies present in the Test blanket System ancillary circuits. • Provided the main achievements for tritium transport modelling tools development. - Abstract: One of the main objectives of the experimental campaign on the Test Blanket Systems (TBS) in ITER is the demonstration of the efficient processing of the tritium generated in the Test Blanket Module (TBM). On the other side, efficient tritium processing in a TBS has deep implications on: (i) safe operation of TBS itself and whole ITER system; (ii) successful development and validation of tritium transport modelling codes; (iii) demonstration of DEMO relevancy of tritium processing technologies. This work describes various aspects of HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS activities related to TBS tritium management. After a short description of HCLL and HCPB blanket concepts and related TBS, the paper contains: 1.a presentation of the key tritium processing technologies in the current design baseline of the European TBS; 2.a discussion on the DEMO relevancy of some specific TBS tritium processing technologies; 3.an overview on the activities related to the tritium transport modelling tools that will be validated along the development of the TBM project, including experimental campaign in ITER, and used for supporting the DEMO Breeding Blanket design. These three items are connected each other since tritium-related data, generated through the experimental campaign in ITER and interpreted through suitable modelling tools, will be one of the most significant outcomes in support of the breeding blanket design for DEMO and beyond

  15. 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...

  16. 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...

  17. 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.

  18. 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.

  19. 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...

  20. A study of permeation barriers for Pb17Li breeding blankets

    International Nuclear Information System (INIS)

    Coatings on the surface in contact with Pb17Li can significantly decrease tritium permeation rate through structural steels and tritium inventory in the blanket. For this purpose, compatibility tests of oxidized and aluminized 316L and 1.4914 steels have been carried out up to 500 deg C in stagnant anisothermal Pb17Li. Only aluminized coatings exhibit a good behaviour, especially on 316L steel. Technological developments of the aluminization process have shown the possibility to coat an experimental loop

  1. 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...

  2. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

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

  3. 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

  4. FIRST STEP blanket structure and fuel assembly design

    International Nuclear Information System (INIS)

    FIRST STEP (Fusion, Inertial, Reduced Requirement Systems Test for Special Nuclear Material, Tritium, and Energy Production) is an Inertial Confinement Fusion (ICF) plant designed to produce tritium, SNM, and energy using near-term technology. It is an integrated facility that will serve as a test bed for fusion power plant technology. The design of the blanket structure and blanket fuel assembly for wetted-wall FIRST STEP reactors is presented here

  5. 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.

  6. 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.

  7. 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)

  8. 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.)

  9. Heatup event analyses of the water cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Water Cooled Solid Breeder (WCSB) Test Blanket Module (TBM) is being designed by JAEA as a primary candidate TBM of Japan. From the viewpoint of the safety, the TBM should be designed so that it does not damage the soundness of the vacuum vessel, the primary barrier for radioisotopes of the ITER. One of the major concerns on the safety of the TBM is temperature elevation due to coolant leakage into the neutron multiplier layer, beryllium, of the TBM. Since the chemical reaction of beryllium and water is an exothermic reaction and the reaction rate exponentially increases with the temperature increase, there is a possibility that the temperature of the TBM exceeds the maximum allowable temperature of its structural material. This paper describes the safety evaluation on the heatup events of the WCSB TBM and proposes the basic strategy to ensure safety, especially incorporating the chemical reaction between beryllium and water. Failure Mode Effect Analysis (FMEA) has been carried out to select the severest heatup events of the WCSB TBM, followed by one-dimensional analyses to evaluate the selected events. The analysis model includes thermal conduction in the TBM, thermal radiation from the TBM to a common frame, and thermal radiation from the TBM first wall to the first wall of the opposite blankets (shield blanket etc.). The sequences of the selected events are shown as follows; Loss of cooling of the TBM during plasma operation is assumed as an initial event. Temperature of the TBM totally increases, then a plasma disruption takes place when the temperature of the first wall armor reaches at a certain value, for example, its melting point of 1273 C. After the plasma disruption, temperature of the TBM decreases according to time and the event converges. However, if the pipe of cooling system in the TBM ruptures due to high temperature, chemical reaction between beryllium and water is activated and the TBM structure is possibly destroyed in the worst case. Therefore

  10. Heatup event analyses of the water cooled solid breeder test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Tsuru, Daigo; Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Agency (Japan)

    2007-07-01

    Water Cooled Solid Breeder (WCSB) Test Blanket Module (TBM) is being designed by JAEA as a primary candidate TBM of Japan. From the viewpoint of the safety, the TBM should be designed so that it does not damage the soundness of the vacuum vessel, the primary barrier for radioisotopes of the ITER. One of the major concerns on the safety of the TBM is temperature elevation due to coolant leakage into the neutron multiplier layer, beryllium, of the TBM. Since the chemical reaction of beryllium and water is an exothermic reaction and the reaction rate exponentially increases with the temperature increase, there is a possibility that the temperature of the TBM exceeds the maximum allowable temperature of its structural material. This paper describes the safety evaluation on the heatup events of the WCSB TBM and proposes the basic strategy to ensure safety, especially incorporating the chemical reaction between beryllium and water. Failure Mode Effect Analysis (FMEA) has been carried out to select the severest heatup events of the WCSB TBM, followed by one-dimensional analyses to evaluate the selected events. The analysis model includes thermal conduction in the TBM, thermal radiation from the TBM to a common frame, and thermal radiation from the TBM first wall to the first wall of the opposite blankets (shield blanket etc.). The sequences of the selected events are shown as follows; Loss of cooling of the TBM during plasma operation is assumed as an initial event. Temperature of the TBM totally increases, then a plasma disruption takes place when the temperature of the first wall armor reaches at a certain value, for example, its melting point of 1273 C. After the plasma disruption, temperature of the TBM decreases according to time and the event converges. However, if the pipe of cooling system in the TBM ruptures due to high temperature, chemical reaction between beryllium and water is activated and the TBM structure is possibly destroyed in the worst case. Therefore

  11. 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: .

  12. 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.

  13. 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...

  14. 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.

  15. 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

  16. 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.)

  17. 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.

  18. 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. PMID:19886461

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. Preconceptual engineering design for the APT 3He Target/Blanket concept

    International Nuclear Information System (INIS)

    A preconceptual engineering design has been developed for the 3He Target/Blanket (T/B) System for the Accelerator Production of Tritium Project. This concept uses an array of pressure tubes containing tungsten rods for the neutron spallation source and 3He gas contained in a metal tank and blanket tubes as the tritium production material. The engineering design is based on a physics model optimized for efficient tritium production. Principle engineering consideration were: provisions for cooling all materials including the 3He gas; containment of the gas and radionuclides; remote handling; material compatibility; minimization of 3He, D2O, and activated waste; modularity; and manufacturability. The design provides a basis for estimating the cost to implement the system

  5. 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.)

  6. Water chemistry control for the target/blanket region of the accelerator production of tritium

    International Nuclear Information System (INIS)

    High-energy particle interactions in the various components of the target/blanket region of the Accelerator Production of Tritium lead to heat generation and deposition. Heavy-water and light-water systems are used to cool the target/blanket system and associated equipment. Structural materials include Inconel alloy 718, aluminum-clad lead rods, aluminum tubes containing helium-3 and tritium gas, and stainless steel components. Proper coolant chemistry is required to maximize neutron production, minimize corrosion of components, and minimize activity buildup. Corrosion-related phenomena and development of coolant and moderator corrosion control for both power and defense fission reactors has been studied extensively over the past 50 years. Less is known, however, about cooling systems for accelerators where a variety of transient chemical species and spallation products may be formed. The following provides a discussion on the issues that need to be addressed for proper water chemistry control for the APT system

  7. 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.)

  8. 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 breeder-inside-tube (BIT) blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. The 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. (author) 8 refs.; 2 figs

  9. Numerical Analyses of Electromagnetic Forces on the ITER Blanket Module Shield Block During Major Disruptions

    International Nuclear Information System (INIS)

    Electromagnetic (EM) load is one of the key design drivers for the blanket shield block (SB) and other in-vessel components. In this article, an EM analysis method was developed to address the EM force on the SB. The plasma currents, which vary spatially and temporally, are loaded as a filament at each time point. The standard blanket module No.04 (BM04) under major disruption (MD) is selected to perform the analyses. The analyses results are validated by comparing currents on the passive structure. To better understand the effects of cooling channels and slits on the EM force, the case of SB without cooling channel and the case without slits are calculated to make comparisons. The results show that the slits play an important role in controlling the EM load on SB. (fusion engineering)

  10. Safety approach in the EU test blanket systems design

    International Nuclear Information System (INIS)

    Highlights: ► European Union (EU) Test Blanket Systems (TBS) safety objectives. ► EU TBS strategy. ► EU TBS project environment and status, risks and mitigations. ► EU TBS safety approach and its implementation. - Abstract: This paper reveals the safety strategy and approach developed and followed in the design of the two EU TBS describing its objectives, components and implementation. Addressing the safety in the early stage of the conceptual design of nuclear facilities is a well recognized international practice and industrial project-level requirement for the successful completion of the licensing process within expected project cost and schedule. The impact of the early development of the safety approach, its implementation and monitoring in the design of nuclear device like the TBS is not limited to the safety assessment and licensing activities only. Safety approach plays indispensible role in reducing the overall project risk. It infiltrates the entire design process through the unavoidable interfaces between the design features and its safety level. In reality the entire process of the TBS development, design, technological demonstration and implementation is affected by the project team safety culture.

  11. System engineering approach in the EU Test Blanket Systems Design Integration

    International Nuclear Information System (INIS)

    The complexity of the Test Blanket Systems demands diverse and comprehensive integration activities. Test Blanket Modules - Consortia of Associates (TBM-CA) applies the system engineering methods in all stages of the Test Blanket System (TBS) design integration. Completed so far integration engineering tasks cover among others status and initial set of TBS operating parameters; list of codes, standards and regulations related to TBS; planning of the TBS interfaces and baseline documentation. Most of the attention is devoted to the establishment the Helium-Cooled Lithium Lead (HCLL) and Helium-Cooled Pebble Bed Lead (HCPB) TBS configuration baseline, TBS break down into sub-systems, identification, definition and management of the internal and external interfaces, development of the TBS plant break down structure (PBS), establishment and management of the required TBS baseline documentation infrastructure. Break down of the TBS into sub-systems that is crucial for the further design and interfaces' management has been selected considering several options and using specific evaluation criteria. Process of the TBS interfaces management covers the planning, definition and description, verification and review, non-conformances and deviations, and modification and improvement processes. Process of interfaces review is developed, identifying the actors, input, activities and output of the review. Finally the relations and interactions of system engineering processes with TBM configuration management and TBM-CA Quality Management System are discussed.

  12. 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

  13. 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

  14. 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.)

  15. 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)

  16. Evaluation of electromagnetic forces on Chinese Dual Functional Lithium Lead Test Blanket Module in ITER

    International Nuclear Information System (INIS)

    The Dual Functional Lithium Lead (DFLL) TBM (Test Blanket Module) concept, using the RAFM steel as structural material, was proposed by Chinese Party as one alternative option of two main blanket concepts for testing in ITER. Because electromagnetic load is one of the main concerns for the ITER in-vessel components, the electromagnetic analysis of DFLL-TBM was implemented using ANSYS code. For transient electromagnetic analysis with the latest disruption scenarios, the complex FEM model was developed including the vacuum vessel, shielding blanket, equatorial port, and divertor of ITER. The goal of this analysis was principally to investigate the eddy current and quantify electromagnetic force on DFLL-TBM at plasma disruption; the peak value of Lorentz force reached 170 kN. Furthermore, the static electromagnetic analysis was carried out with the reference operation scenario II, the magnetization forces on DFLL-TBM due to magnetization were investigated. By the mechanical analysis, the results show the DFLL-TBM can accommodate the calculated loads.

  17. Remote-handling concept for target/blanket modules in the accelerator production of tritium

    International Nuclear Information System (INIS)

    The accelerator production of tritium (APT) has been proposed as the source of tritium for the United States in the next century. The APT will accelerate protons that will strike replaceable tungsten target modules. The tungsten target modules generate neutrons that pass through blanket modules and other modules where He gas is turned into tritium. The target and blanket modules are predicted to require replacement every 1 to 10 yr, depending on their location. The target modules may weigh as much as 78.8 tonnes (85 t) each. All of the modules will be contained in a target/blanket vessel, which is in a shielded facility. The spent modules will be very radioactive so that remote replacement is required. A proposed concept is to use a remotely operated bridge crane and a remotely operated, bridge-mounted manipulator to perform the entire replacement operation. This will require removing/replacing the vessel lid, installing/removing temporary water cooling, closing/opening valves on manifolds and modules, draining of jumpers, removing/replacing jumpers, removing/replacing shielding keys, and removing/replacing the modules. This application is unique because of the size and weight of the modules, the precision required, the type of connectors required, and the complexity of the entire operation. A three-dimensional simulation of the entire module replacement operation has been developed to help understand, communicate, and refine the concepts

  18. 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)

  19. 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

  20. Basic Concepts of DEMO and a Design of a Helium Cooled Molten Lithium Blanket

    International Nuclear Information System (INIS)

    Demonstration fusion power plant, DEMO is regarded as the last step before the development of a commercial fusion reactor in Korea National Basic Plan for the Development of Fusion Energy. The DEMO should demonstrate a net electric power generation, a tritium self sufficiency, and the safety aspect of a power plant. With a limited extension of the improved plasma physics and technology from the 2nd phase of the ITER operation (EPP phase), we developed the basic concepts of DEMO and identified the design parameters by considering the dependence of DEMO on performance objectives, design features and physical and technical constraints. Extensive system analyses have been performed to find device variables which optimize figures of merit such as major radius, ignition margin, neutron wall load, etc. The He Cooled Molten Lithium/FS (HCML) blanket is one of options for DEMO blanket and its tritium breeding capability and heat removal capability will be tested in ITER as a test blanket module (TBM). HCML blanket uses He as a coolant and Li as a tritium breeder. From a sensitivity study, 6Li enrichment was optimized in terms of tritium breeding ratio (TBR). An optimum was found for a natural enrichment in DEMO blanket but it was 12 wt% in TBM since the amount of Li is limited in ITER. Two layers of a graphite reflector were inserted as a reflector in the breeder zone to increase the TBR and the shielding performances. The graphite reflector thickness was optimized to maximize TBR without any special neutron multiplier and to minimize the neutron leakage. For TBM, a 3-D Monte Carlo neutronic analysis was performed with the MCCARD code and the total power was founded to be a 0.739 MW at normal heat flux 0.3 MW/m2 from plasma side. From the thermal-hydraulic analysis using CFX-10, the He cooling path was optimized and it was found that the maximum temperature of FW is below 550 oC at structural materials and the coolant velocities are 45 m/sec and 8.2 m/sec at FW and breeding

  1. Probabilistic Fracture Mechanics Analysis of APT Blanket Tubes

    International Nuclear Information System (INIS)

    A probabilistic fracture mechanics (PFM) model that is specific to the Accelerator Production of Tritium (APT) helium tubes was developed. The model performs Monte Carlo analyses of potential failure modes caused by cyclic stresses generated by beam trips and de-pressurizations from normal operation, coupled with material aging due to irradiation. Dominant failure probabilities are due to crack through-growth while brittle fracture and ductile tearing have lower probability. Failure mechanisms of global plastic collapse and buckling or crack initiation mechanisms of fatigue or local fracture (upon loss of ductility) have negligible probability. For the population of (7,311) tubes in the APT blanket, the worst-case, annual probability of one tube failing is 3 percent. The probability of 2 or more failures is substantially lower; therefore, unavailability impacts are driven by single failure. The average annual loss of production (unavailability) is below about 0.2 percent. Helium outflow and water inflow rates were characterized for the failures. (authors)

  2. Thermomechanics analysis and optimization for high power density blanket

    International Nuclear Information System (INIS)

    Thermomechanics analysis, i.e. steady thermal analysis and steady thermal stress analysis have been carried out for a high power density blanket. The Fusion Experimental Breeder (FEB) is adopted as the reference reactor. The parts for the blanket module in Pro/ENGINEER were created, then turn to Pro/MECHANICA functionality for thermomechanics analysis. During analysis, the distribution of the power density in the blanket was optimized to be more flat, the arched curvature and rounds of the cooling tube panels were optimized to less stiffness, and the boundary condition at the interface of helium cooling tube panel and manifold chamber was optimized, which is reasonable by using advanced welding processes with electron beam or laser beam in a single pass. To the end, a maximum temperature Tm 350 degree C and a maximum shear stress τm 80 MPa for the helium cooling panels have been shown in the calculations. (authors)

  3. 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.

  4. 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)

  5. 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)

  6. 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)

  7. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    International Nuclear Information System (INIS)

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials (1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF4 or ThF4 or some combination thereof. Future systems could look at using PuF3 or PuF4 as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory. Preliminary design studies

  8. 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

  9. 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 ...

  10. 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.

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

    International Nuclear Information System (INIS)

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

  12. 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

  13. Burnable Poison optimization for Seed-Blanket Cores

    International Nuclear Information System (INIS)

    The main objective of the Seed and Blanket Units (SBU) core designs is to reduce production of Pu and long-term toxicity of the spent PWR fuel. The SBU concept assumes a heterogeneous seed-blanket fuel assembly, with spatial separation of the U and Th parts of the fuel. In the SBU assembly the Seed fuel is an alloy of 20% enriched metallic Uranium and zircaloy, the Blanket fuel is ThO2 mixed with about 13% of 12.2% enriched UO2. The Uranium is included in the mix in order to increase the BOL power in the Th pins and dilute the bred U233 isotope to avoid proliferation concerns. The 108 Seed fuel rods are located in the central region of the assembly and surrounded by 156 blanket rods. The use of metallic fuel in the Seed enables high density of fissile material. The U-Zr alloy also has a higher thermal conductivity than UO2, although this advantage is partially offset by the low melting temperature of metallic fuels. More importantly, compared with oxide fuel, the radiation induced creep and swelling phenomena are more pronounced in metallic fuel at elevated temperatures due to the loss of its crystallographic state. This loss occurs at a rather low temperature of 6600 C, and in U-Zr alloys at 6160 C. For this reason, reduction of the local pin power peaking in the assembly is particularly important

  14. 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...

  15. LMFBR blanket assembly heat transfer and hydraulic test data evaluation

    International Nuclear Information System (INIS)

    The USA Test Program for characterization of breeder reactor blanket T and H performance is providing a data base for improved confidence in the design tools employed. Pressure drop tests with wire wrapped rod bundles having a 1.08 triangular pitch to diameter ratio and 4 inch (10 cm) wire wrap lead using water, sodium and air have defined a smooth, continuous, single-valued friction factor versus Reynolds number correlation. This eliminates a possible source of flow instability. The rod bundle temperature rise profiles measured in the heat transfer tests using a prototypic blanket rod bundle agrees in magnitude and shape with the predictions of the marching type sub-channel codes currently employed in blanket subchannel analysis. The low flow test data demonstrates increasing buoyancy induced flows in the lower Reynolds number flow regime. This and the remaining test data will supply a base for calibration of the mixing momentum exchange and conduction factors employed in the subchannel analysis codes; which will contribute to the confidence of the blanket design predictions and reduce the uncertainties which are commonly expressed as hot channel/spot factors

  16. Anchoring FRP Composite Armor in Flexible Offshore Riser Systems

    DEFF Research Database (Denmark)

    Costache, Andrei

    Unbonded flexible pipes find extensive use in the offshore oil industry. Although more expensive than rigid pipe, the total cost of flexible pipe installations are often less. This is because flexible pipes are easier to store and deploy, coupled with superior fatigue performance. Among other...... composite materials. One of the problems related to the substitution of tensile steel members is that anchoring in the metallic end fittings of the pipe is very challenging.The purpose of this thesis is to ensure the transfer of tensile loads between a unidirectional fiber reinforced polymer and a metallic...... the grip, and to measure the contact stresses distribution and evolution during pullout.The results of the experimental and numerical analysis show that it is possible to reliably anchor composite materials using a metallic grip. The models developed during the project show how to improve the...

  17. Fugitive emissions control on dry copper tailing with crushed rock armor

    International Nuclear Information System (INIS)

    Four inactive copper tailing impoundments totalling 1,900 acres near Ajo in southwestern Arizona were covered on horizontal surfaces with a 2 in. nominal thickness of crushed rock to control particulate emissions. The tailings are typically dominated by sand-sized particles but may also include significant PM10 fractions towards the centers of the impoundments. The technology was selected by Phelps Dodge Corporation, after investigation of several alternatives, as a permanent and practical cover that essentially eliminates fugitive emissions. It simulates the natural desert pavement that characterizes this arid area of the Sonoran Desert. Rocky overburden was crushed to minus 3 in. diameter and broadcast on dry surfaces of tailing impoundments with all-terrain, balloon-tired spreaders. Stony residues in the rock armor tend to cement together following rainfall, forming a crust that enhances surface stability and erosion control. Slopes with windblown tailing deposition were covered to a nominal 6 in. thickness by conventional dozer pushing and blading of minus 10 in. rock over the sides. Athel trees, planted extensively since 1970 on two of the four inactive impoundments, provided partial control of fugitives, but were subjected to harsh environmental conditions, including abrasion from saltating particles. The rock armor functions as a mulch which is expected to improve water relations for existing vegetation and areas seeded with native species. New surface microenvironments, and the virtual elimination of surface creep and saltation, are expected to support native plant growth under favorable climatic conditions

  18. Parasitoids and Predators of Armored Scales in Some Orchards in Serbia

    Directory of Open Access Journals (Sweden)

    Draga Graora

    2009-01-01

    Full Text Available Five parasitoid and five predator species were registered on four armored scale species in apple, pear, peach and plum orchards at 16 sites in Serbia. The armored scales found were: Diaspidiotus perniciosus (Comstock, Pseudaulacaspis pentagona (Targioni-Tozzetti, Epidiaspis leperii (Signoret and Lepidosaphes ulmi (L..The parasitoid species detected were: Encarsia (=Prospaltella perniciosi (Tow., Encarsia (=Prospaltella berlesei (How., Aphytis proclia (Wal., Aphytis mytilaspidis (Le Baron and Coccobius testaceus (Masi, all of the Aphelinidae family. The most important endoparasitoids were: Encarsia perniciosi on D. perniciosus, Encarsiaberlesei on P. pentagona and Coccobius testaceus on L. ulmi. Only one ectoparasitoid species, Aphytis mytilaspidis, was found on E. leperii, while Aphytis proclia was found as an ectoparasitoid on D. perniciosus, P. pentagona and L. ulmi. The predator species detected were: Chilocorus renipustulatus (Scriba, Chilocorus bipustulatus(L., Cybocephalus fodori Endrody – Younga, Chrysoperla carnea (Stephens and Deraeocoris ruber (L.. Cybocephalus fodori was for the first time found as a predator on E. leperii in Serbia.

  19. Structure and mechanical behaviors of protective armored pangolin scales and effects of hydration and orientation.

    Science.gov (United States)

    Liu, Z Q; Jiao, D; Weng, Z Y; Zhang, Z F

    2016-03-01

    As natural flexible dermal armor, pangolin scales provide effective protection against predatory threats and possess other notable properties such as anti-adhesion and wear-resistance. In this study, the structure, mechanical properties, deformation and damage behaviors of pangolin scales were systematically investigated with the effects of hydration and orientation evaluated. The scales are divided into three macro-layers constituted by overlapping keratin tiles with distinct lamellar arrangements which are further composed of lower-ordered lamellae. Both hardness and strength are significantly decreased by hydration; while the plasticity is markedly improved concomitantly, and as such, the mechanical damages are mitigated. The tensile strength invariably approximates to one third of hardness in value. The tensile deformation is dominated by lamellae stretching and pulling out under wet condition, which is distinct from the trans-lamellar fracture in dry samples. The compressive behaviors are featured by pronounced plasticity in both dry and wet scales; and notable strain-hardening capacity is introduced by hydration, especially along the thickness direction wherein kinking occurs. Inter-lamellar cracking is effectively alleviated in wet samples compared with the dry ones and both of them deform by macroscopic buckling. This study may help stimulate possible inspiration for the design of high-performance synthetic armor materials by mimicking pangolin scales. PMID:26703230

  20. Tritium permeation through helium-heated steam generators of ceramic breeder blankets for DEMO

    International Nuclear Information System (INIS)

    The potential sources of tritium contamination of the helium coolant of ceramic breeder blankets have previously been evaluated for the specific case of the European BIT DEMO blanket. This confirmed that the control of tritium losses to the steam circuit is a critical issue which demands development concerning (a) permeation barriers, (b) tritium recovery processes maintaining a very low tritium activity in the coolant, and (c) control of the coolant chemistry. The specifications of these developments required the evaluation of the tritium losses through the steam generators, and includes the definition of their operating conditions by thermodynamic cycle calculations, and their thermal-hydraulic design. For both tasks, specific computer tools were developed. The geometry obtained, the surface area and the temperature profiles along the heat-exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam-oxidized Incoloy 800 austenitic stainless steel was identified as the best-suited existing material. Our results indicate that in nominal steady-state operation the tritium escape into the steam cycle could be restricted to less than 10 Ci per day. The conditions for this are specified, but their feasibility demands, in particular, the resolution of certain gas chemistry problems, and their validation in the more stringent environment of an operating blanket. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) was identified as bearing a large tritium release potential. The problems associated with such transients are discussed and possible solutions are proposed. (orig.)

  1. 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

  2. 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)

  3. 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

  4. Improved thermal/MHD design of self-cooled blankets for high-power-density fusion reactors

    International Nuclear Information System (INIS)

    In this work, an improved self-cooled blanket design is conceived that seeks to minimize the induced current and pressure loss, while maintaining effective cooling and power output. Standard solutions for fully developed MHD flows in rectangular ducts are utilized to describe the magnetic pressure drop in rectangular ducts in terms of the duct aspects ratio. A newly available analytical result for developing and fully developed temperatures is utilized in determining the maximum wall temperature and outlet temperature. Based on results from rectangular ducts, improved annular-type duct designs are proposed and evaluated. The results from the rectangular duct analysis indicate reduced pressure drop and increased thermal performance for large aspect ratio (ratio of duct width in the toroidal B-field direction to width normal to B-field). An infinite aspect ratio occurs for the annular duct design and it is shown that this configuration has superior characteristics as a self-cooled blanket design concept

  5. Neutronics analysis for the test blanket modules proposed for EAST and ITER

    International Nuclear Information System (INIS)

    The Dual-Functional Lithium Lead - Test Blanket Module (DFLL-TBM) system, which is designated to demonstrate the integrated technologies of both He single coolant (SLL) blanket and He- LiPb dual coolant (DLL) blanket, is proposed for test in ITER to check and validate the feasibility of the Chinese LiPb blankets. So far, the construction and operation of ITER will still take a period of ten years, but EAST, the superconducting tokamak device, in China, has been in operation. In EAST D-D phase, the neutron yield is about 1015 ∼ 1017 n/s and about 1017 ∼ 1018 n/s in ITER D-D phase. Therefore, EAST is expected to serve as a valuable pre-testing platform for TBMs, which is not only for electro-magnetics (EM) and thermo-mechanics but also for neutronics. The neutronics analysis for the TBMs is performed by using the coupled three-dimensional (3D) Monte Carlo - Deterministic code MCSN and the nuclear data library FENDL2.1. The activation calculations will be carried out with the home-developed multi-functional neutronics analysis code system VisualBUS and multi-group data library HENDL. The real 3D neutronics calculation model of the middle-scale (1/3 size-reduced) TBM testing in the EAST super-conducting tokamak and full-scale consecutive TBM testing in the ITER machine have been developed with the Chinese home-developed CAD/MCNP interface code MCAM, which can be used as a converter of large complex 3D CAD models into MCNP models and vice versa as well as an analysis tool of MCNP models by the way of visualization to contribute the QA of neutronics analysis. Neutronics calculations, which include neutron spectra and flux distributions, tritium generation, nuclear energy deposition and D-D phase activation, of the TBMs in EAST are carried out and be made an analogy to those in ITER for the close extent of the neutron yield in D-D phase. Further, the foreseen D-D operations in ITER can be treated as an initial nuclear phase including D-T operation. So the presented

  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. 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

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

    International Nuclear Information System (INIS)

    The Vanadium/Lithium system has been the recent focus of ANL's Blanket Technology Pro-ram, and for the last several years, ANL's Liquid Metal Blanket activities have been carried out in direct support of the ITER (International Thermonuclear Experimental Reactor) breeding blanket task area. A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the Near the development of insulator coatings. Design calculations, Hua and Gohar, show that an electrically insulating layer is necessary to maintain an acceptably low magneto-hydrodynamic (MHD) pressure drop in the current ITER design. Consequently, the decision was made to convert Argonne's Liquid Metal EXperiment (ALEX) from a 200 degrees C NaK facility to a 350 degrees C lithium facility. The upgraded facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups at Hartmann numbers (M) and interaction parameters (N) in the range of 103 to 105 in lithium at 350 degrees C. Following completion of the upgrade work, a short performance test was conducted, followed by two longer multiple-hour, MHD tests, all at 230 degrees C. The modified ALEX facility performed up to expectations in the testing. MHD pressure drop and test section voltage distributions were collected at Hartmann numbers of 1000

  9. Conversion of a room temperature NaK loop to a high temperature MHD facility for Li/V blanket testing

    International Nuclear Information System (INIS)

    The Vanadium/Lithium system has been the recent focus of ANL's Blanket Technology Program, and for the last several years, ANL's Liquid Metal Blanket activities have been carried out in direct support of the ITER (International Thermonuclear Experimental Reactor) breeding blanket task area. A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the development of insulator coatings. Design calculations, Hua and Gohar, show that an electrically insulating layer is necessary to maintain an acceptably low magnetohydrodynamic (MHD) pressure drop in the current ITER design. Consequently, the decision was made to convert Argonne's Liquid Metal EXperiment (ALEX) from a 200 degree C NaK facility to a 350 degree C lithium facility. The upgraded facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups at Hartmann numbers (M) and interaction parameters (N) in the range of 103 to 105 in lithium at 350 degree C. Following completion of the upgrade work, a short performance test was conducted, followed by two longer, multiple-hour, MHD tests, all at 230 degree C. The modified ALEX facility performed up to expectations in the testing. MHD pressure drop and test section voltage distributions were collected at Hartmann numbers of 1000. 4 refs., 2 figs

  10. 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

  11. 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

  12. Understanding the spatial structure of peat permeability around natural pipes in blanket peatlands

    Science.gov (United States)

    Cunliffe, Andrew; Baird, Andy; Holden, Joseph

    2014-05-01

    Understanding the spatial structure of peat permeability around natural pipes in blanket peatlands We present the results of a detailed investigation of fine-scale variations in the permeability or hydraulic conductivity (K) of the peat around a natural pipe in a blanket peatland. Both vertical K and horizontal K ranged over seven orders of magnitude over scales of decimetres. K was found to be more variable than indicated by previous research. This finding has important implications for the approaches currently employed to investigate peatland hydrological processes, and the parameterisation of models used to simulate these complex ecohydrological systems. We also observed considerable spatial structuring in K. Lateral K parallel to the pipe was significantly greater than lateral K perpendicular to the pipe. Critically, a wedge of poorly-humified, high-permeability peat was present directly above the pipe, forming a hydrological connection between the peatland surface and the perennially-flowing pipe. These observations advance our mechanistic understanding of pipeflow generation in peatlands. We also attempted to investigate K across the pipe-peat interface to test for a hypothesised low-K skin; however, this was precluded by sample length dependency, which suggests that it is inappropriate to compare K measurements between peat samples of different lengths. Overall, we argue that high resolution work such as this is required for the development of more accurate perceptual models of peatland hydrological systems. Cunliffe, A. M., A. J. Baird, and J. Holden (2013), Hydrological hotspots in blanket peatlands: Spatial variation in peat permeability around a natural soil pipe, Water Resources Research, Vol.49, doi:10.1002/wrcr.20435.

  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. The EU advanced lead lithium blanket concept using SiCf/SiC flow channel inserts as electrical and thermal insulators

    International Nuclear Information System (INIS)

    Preparatory work on the EU advanced dual coolant (A-DC) blanket concept using SiCf/SiC flow channel inserts as electrical and thermal insulators has been carried out at the Forschungszentrum Karlsruhe in co-operation with CEA (SiCf/SiC composite-related issues) as a conceptual design proposal to the EU fusion power plant study planned to be launched in 2001 within the framework of the EU fusion programme having the main objective of specifying the characteristics of an attractive and viable commercial D-T fusion power plant. The basic principles and the study method for the A-DC blanket concept are presented in this report. The results of this study show that the A-DC blanket concept has a high potential for further development due to its high thermal efficiency and its simple concept solution

  16. APT 3He target/blanket. Topical report

    International Nuclear Information System (INIS)

    The 3He target/blanket (T/B) preconceptual design for the 3/8-Goal facility is based on a 1000-MeV, 200-mA accelerator to produce a high-intensity proton beam that is expanded and then strikes one of two T/B modules. Each module consists of a centralized neutron source made of tungsten and lead, a proton beam backstop region made of zirconium and lead, and a moderator made of D2O. Helium-3 gas is circulated through the neutron source region and the blanket to create tritium through neutron capture. The gas is continually processed to extract the tritium with an online separation process

  17. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

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

  18. Deposition and properties of blanket-W using silane reduction

    International Nuclear Information System (INIS)

    The authors have studied the deposition and properties of silane-reduced blanket tungsten films, and their suitability as interconnects. Their reflectance and surface morphology on Si substrates are higher than those of H2 - reduced films. This is suited for submicron lithography. Resistivity becomes 8 μΩ · cm, which is the same as that of H2-reduced films. Surface morphology and resistivity degrade when β-W is formed. Since step coverage of interconnects in half-micron contact holes is improved with selective deposition, a combination of silane-reduced selective and blanket tungsten deposition allows both contact and wiring metallization. This combination is also desirable in multilevel interconnects on ULSIs

  19. Nuclear performance analyses for HCPB test blanket modules in ITER

    International Nuclear Information System (INIS)

    Neutronic, shielding and activation analyses have been performed for recent design variants of the Helium Cooled Pebble Bed (HCPB) test blanket module (TBM) in ITER on the basis of 3D Monte Carlo calculations. The main objective has been to assess and optimise the nuclear performance of the HCPB test blanket modules in terms of the tritium generation, the nuclear heating and the radiation shielding and provide, among others, the data required for the engineering design of the test modules. The shielding efficiency of the TBM system was shown to be sufficient to allow access of work personnel to the port extension after a waiting time of 10 days after shut down as required by ITER. The activation analyses provided the afterheat and activation data for quality assured safety analyses assuming a representative irradiation scenario

  20. GCFR radial blanket and shield experiment: objectives, preanalysis, and specifications

    International Nuclear Information System (INIS)

    An integral experiment has been designed for the verification of radiation transport methods and nuclear data used for the design of the radial shield for the proposed 300 MW(e) gas-cooled fast breeder reactor (GCFR). The scope of the experiment was chosen to include a thorium oxide radial blanket mockup as well as several shield configurations in order to reduce the uncertainties in the calculated source terms for the radial shield, and to reduce the uncertainties in the calculated radiation damage to the prestressed concrete reactor vessel (PCRV). Additionally, the measurements are intended to bound the uncertainties in calculated gamma-ray heating rates within the blanket and shield. Although designed specifically for the GCFR, the experiment will provide generic data regarding deep penetration in ThO2 and common shield materials, which should also benefit LMFBR designers

  1. 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

  2. Damage structure in divertor armor materials exposed to multiple ITER relevant ELM loads

    International Nuclear Information System (INIS)

    The damage threshold and damage mechanisms of divertor armor materials, i.e. CFC and tungsten, were studied under the impact of ITER relevant ELM-like loads. These experiments were carried out in a Quasi-Stationary Plasma Accelerator applying repetitive pulses of 500 μs up to 100 cycles. CFC showed preferential erosion of the PAN fiber-bundles above 0.6 MJ/m2 and cracking of pitch fiber-bundles. Tungsten showed cracking already at 0.2 MJ/m2 and melting at flat surfaces above 1 MJ/m2. Cracks in tungsten were identified as primary and secondary cracks which all propagated in the vertical direction, which was considered to be less critical. At an energy density of 1.5 MJ/m2, the melt-layer completely covered the surface and bridged the castellation slots.

  3. 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

  4. ITER Blanket First Wall (WBS 1.6{sub 1}A)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Kim, H. G.; Kim, J. H. (and others)

    2008-03-15

    International Thermonuclear Experimental Reactor (ITER) project is the international collaboration one for the commercialization of nuclear fusion energy through the technical and engineering verification. In ITER project, we plan to procure the blanket systems which has the risk of technology and cost when it is newly developed. We are developing the manufacturing process and joining technology for the ITER blanket to complete the procurement with qualified blanket system. To evaluate the soundness of manufacturing process, specimen and mock-up tests are being prepared. Finally, we can obtain the key technology of nuclear fusion reactor especially on the blanket design, joining and manufacturing technology through the present project and these technologies will help the construction of Korea fusion DEMO reactor and the development of commercial nuclear fusion reactor in Korea. In 1st year, through the fabrication of the Cu/SS and Be/Cu joint specimen, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The optimized HIP conditions (1050 .deg. C, 150 MPa, 2 hr for Cu/SS and 580 - 620 .deg. C, 100-150 MPa, 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint and NDT such as UT (10 MHz, 0.25 inch D, flat type) and ECT. Several mock-ups were fabricated for confirming the joint integrity and NDT. specimens fabricated with these mock-ups were used in mechanical tests including microstructure observation. The mock-ups were used in the HHF test after the developed NDT. In 2nd year, PHHT of Cu was investigated in order to recover its mechanical properties, and the pre-qualification mock-up were fabricated against the Qualification Program and sent to RF for HHF testing in TSEFEY. FW fabrication and joining procedure were documented in the form of the TSD. Qualification mock

  5. 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.

  6. 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

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

    International Nuclear Information System (INIS)

    An 80-K Multi Layer Insulation (MLI) blanket in the interconnect region between magnets DD0019 and DD0027 in the Fermi National Accelerator Laboratory (FNAL) ER string was instrumented with temperature sensors to obtain the steady-state temperature gradient through the blanket after string cooldown. A thermal model of the 80-K blanket assembly was constructed to analyze the steady-state temperature gradient data. Estimates of the heat flux through the 80-K MLI blanket assembly and predicted temperature gradients were calculated. The thermal behavior of the heavy polyethylene terapthalate (PET) cover layers separating the shield and inner blanket and inner and outer blankets was derived empirically from the data. The results of the analysis predict a heat flux of 0.363--0.453 W/m2 based on the 11 sets of data. These flux values are 33--46% below the 80-K MLI blanket heat leak budget of 0.676 W/m2. The effective thermal resistance of the two heavy PET cover layers between the shield and inner blanket was found to be 2.1 times that of a single PET spacer layer, and the effective resistance of the two heavy PET cover layers between the inner blanket and outer blanket was found to be 7 times that of a single PET spacer layer. Based on these results, the 80-K MLI blanket assembly appears to be performing more than adequately to meet the 80-K static IR heat leak budget. However, these results should not be construed as a verification of the 80-K static IR heat leak, since no actual heat leak was measured. The results have been used to improve the empirically based model data in the 80-K MLI blanket thermal model, which has previously not included the effects of heavy PET cover layers on 80-K MLI blanket thermal performance

  8. Non-baseline damage detection from changes in strain energy mode shapes experiments on armored vehicle launched bridge

    International Nuclear Information System (INIS)

    There are several existing methods for damage detection based on identifying changes in strain energy mode shapes. Most of these methods require knowing strain energy mode shapes for a structure without damage in order to establish a baseline for damage detection. Usually, the mode shapes from the structure under test should be compared to the baseline mode shapes to identify and locate damage. However, these methods of damage detection are not very suitable for application on structures where baseline mode shapes cannot be readily obtained, for example, structures with preexisting damage. Conventional methods, like building a finite element model of a structure to be used as a baseline might be an expensive and time-consuming task that can be impossible for complex structures. A new (non-baseline) method for the extraction of localized changes (damage peaks) from strain energy mode shapes based on Fourier analysis of the strain energy mode shapes has been developed and analytically proved for the cases of a pinned-pinned and a free-free beam. The new method looks for characteristic changes in the power spectrum of the strain energy mode shapes in order to locate and identify damage. The analytical results have been confirmed both by the finite element model and impact testing experiments on a free-free aluminum beam, including single and multiple damage scenarios. This paper presents results of testing the non-baseline method on a complex structure -- Armored Vehicle Launched Bridge, which consists of loosely coupled hinged beams with variable cross-section. The results of testing confirm applicability of the non-baseline method to damage detection in complex structures and highlight certain particularities of its use

  9. 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

  10. Test apparatus for ITER blanket pebble packing behavior

    International Nuclear Information System (INIS)

    Current Japanese design for ITER Driver Blanket consists of three breeder layers, nine multiplier layers and five cooling panels. The breeder layers and the multiplier layers contain 1 mm diameter spheres of Li2O and Be, respectively. The heat transfer in such 'Pebble Layered Blanket' is largely affected by the packing fraction of the pebbles which can be easily changed by the vibration during the operation. The packing fraction of the pebbles are expected to be as high as possible on the view point of nuclear heat design to maintain the optimum temperature of the breeder layer. Thus, it is necessary to establish the stable packed bed of the breeder and multiplier. The present experimental apparatus was fabricated for the engineering tests with the partial model of Japanese blanket. Test apparatus consists of stainless steel test panels, transparent plastic test panels, vibrators and measurement instruments. The apparatus can examine various parameters of sphere packed beds such as packing fraction, panels deformation, loading weight at the bottom of the panels and so on under various vibrating conditions. (author)

  11. Enhanced plasma current collection from weakly conducting solar array blankets

    Science.gov (United States)

    Hillard, G. Barry

    1993-05-01

    Among the solar cell technologies to be tested in space as part of the Solar Array Module Plasma Interactions Experiment (SAMPIE) will be the Advanced Photovoltaic Solar Array (APSA). Several prototype twelve cell coupons were built for NASA using different blanket materials and mounting techniques. The first conforms to the baseline design for APSA which calls for the cells to be mounted on a carbon loaded Kapton blanket to control charging in GEO. When deployed, this design has a flexible blanket supported around the edges. A second coupon was built with the cells mounted on Kapton-H, which was in turn cemented to a solid aluminum substrate. A final coupon was identical to the latter but used germanium coated Kapton to control atomic oxygen attack in LEO. Ground testing of these coupons in a plasma chamber showed considerable differences in plasma current collection. The Kapton-H coupon demonstrated current collection consistent with exposed interconnects and some degree of cell snapover. The other two coupons experienced anomalously large collection currents. This behavior is believed to be a consequence of enhanced plasma sheaths supported by the weakly conducting carbon and germanium used in these coupons. The results reported here are the first experimental evidence that the use of such materials can result in power losses to high voltage space power systems.

  12. Test Blanket Module Pipe Forest integration in ITER equatorial port

    International Nuclear Information System (INIS)

    ITER Test Blanket Modules (TBMs) will allow testing Breeding Blanket concepts for a future application in DEMO. IRFM (Institut de Recherche sur la Fusion Magnetique) contribution to this test program consists in the integration of the 2 European TBMs (Helium Cooled Lithium Lead and Helium Cooled Pebble Bed) in a dedicated equatorial port. The two Breeding Blanket concepts use Helium gas as a coolant, liquid PbLi as breeder (for HCLL process) and Helium gas for Tritium extraction (for HCPB process). These materials are passing through the cryostat interspace forming a pipe network called the Pipe Forest. The main structural function of the Pipe Forest is to absorb the thermal expansion due to the Vacuum Vessel and due to the pipe system itself. The Pipe Forest has to cope with several design issues. In this study, the different key parameters of the Pipe Forest design are identified and their relative influence is analysed. Several design options were investigated and compared based on: -Thermo-mechanical finite element calculations -Pipe Forest integration within the cryostat interspace -Interface management -Assembly and maintenance scenarios -Complex pipe routing due to the expansion bends -RCC-MR 2007 requirements The chosen thermal compensation solution (thermal expansion loops) led to a Pipe Forest design. The CAE analysis of this Pipe Forest showed that it fulfills the requirements of the RCC-MR 2007, which is the reference design and construction code selected for the European TBM.

  13. 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

  14. 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.

  15. Mechanical analysis of a model of the breeding blanket of NET in faulted conditions

    International Nuclear Information System (INIS)

    This paper has been prepared in the framework of the safety analysis of a breeding blanket proposed for NET (Next European Torus). The basic features of the system are the following: - Li17Pb83 as breeder; - pressurized (5 MPa) water as coolant; - AISI 316 SS as structural material. The breeding blanket consists of 24 segments with an angular opening of 150 placed side by side in the toroidal direction and arranged in the inboard and outboard part of the plasma chamber. The outboard part of the segment is presently under development, and two different design options are proposed: - a modular concept in which the breeding units (arranged in five rows and four columns), named modulus, look like boxes; - a tubular concept in which the breeding units are tubes bent in the poloidal direction. In both concepts the vessel of the breeding unit must operate as the first barrier against the accident propagation in case of a pipe break in the unit's cooling system. The mechanical behaviour of the modular concept, loaded by the pressure transient due to such a pipe break, has been investigated and is presented in detail. The analysis of the result, taking into account material non-linearities, fluid-structure interactions and dynamic effects, shows that the structural reliability of the module vessel cannot be guaranteed, and suggests to continue the development of the tubular concept for which a much better mechanical behaviour is expected. (orig.)

  16. Design and techniques for fusion blanket neutronics experiments using an accelerator-based deuterium-tritium neutron source

    International Nuclear Information System (INIS)

    The experiments performed in the Japan Atomic Energy Research Institute/U.S. Department of Energy collaborative program on fusion blanket neutronics are designed with consideration of geometrical and material configurations. The general guide that is used to design the engineering-oriented neutronics experiment, which uses an accelerator-based 14-MeV neutron source, is discussed and compared with neutronics characteristics of the reactor models. Preparation of the experimental assembly, blanket materials, and the neutron source is described. A variety of techniques for measuring the nuclear parameters such as the tritium production rate are developed or introduced through the collaboration as a basis of the neutronics experiments. The features of these techniques are discussed with the experimental error and compared with each other. 25 refs., 15 figs., 4 tabs

  17. Theoretical investigations of a double ionization chamber for on-line monitoring of tritium production in fusion blankets

    International Nuclear Information System (INIS)

    This paper reports that a double ionization chamber employing a thin coating of enriched 6LiF radiating material offers an effective means of identifying a 6Li(n, α)t reaction. The concept is based on the detection of ionization caused by alpha particles and tritons. The charged particles emitted in opposite directions can be detected by a double parallel plate ionization chamber configuration. This method can therefore be employed to directly measure tritium breeding rates inside the fusion blankets. Complete details of the parameters that govern the response of such a detector system are described. A Monte Carlo scheme is developed to determine the direction and energy lost by the particles in traversing various media, and the detector response is calculated from the energy deposited in the ionization region of each chamber. The calculations are performed for the entire energy range of neutrons available in the fusion blankets

  18. Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

    Energy Technology Data Exchange (ETDEWEB)

    Madeleine, S. [CEA Cadarache, DSM/IRFM, F-13108 Saint Paul Lez Durance (France)], E-mail: sylvain.madeleine@cea.fr; Saille, A.; Martins, J.-P. [CEA Cadarache, DSM/IRFM, F-13108 Saint Paul Lez Durance (France); Salavy, J.-F.; Jonqueres, N.; Rampal, G. [CEA Saclay, DEN/DM2S, F-91191 Gif sur Yvette (France); Bede, O. [HAS - KFKI-RMKI, P.O. Box 49, H-1525, Budapest (Hungary); Neuberger, H.; Boccaccini, L. [FZK/Karlsruhe, IRS - Forschungszentrum Karlsruhe GmbH Karlsruhe (Germany); Doceul, L. [CEA Cadarache, DSM/IRFM, F-13108 Saint Paul Lez Durance (France)

    2009-06-15

    The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak. The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port. This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.

  19. Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

    International Nuclear Information System (INIS)

    The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak. The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port. This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.

  20. 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

  1. Studies of a self-cooled lead lithium blanket for HiPER reactor

    International Nuclear Information System (INIS)

    Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall - Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates. (authors)

  2. 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)

  3. 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.

  4. 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

  5. Numerical simulation of turbulent flow of coolant in a test blanket module of nuclear fusion reactor

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency has been performing the research, development and design of a test blanket module with a water-cooled solid breeder for ITER. For our design, the TBM is mainly composed of a first wall, two side walls, a back wall and membrane panels of bulkhead sections for pebbles. The temperature of a coolant pressurized up to 15 MPa is designed as 553 K and 598 K in an inlet and an outlet of the test blanket module, respectively. Establishment of estimation methods of the flow phenomena is important for designs of the channel network and predictions of the material corrosion and erosion. A purpose of our research is to establish and verify the method for the prediction of the flow phenomena. In this study, the Large-eddy simulation and Reynolds averaged Navier-Stokes simulation have been performed to predict the flow rates in the channels of the side wall. It results the inhomogeneous flow rates in each channel. At viewpoint of the heat removal capability, however, the smallest flow-rates near the first wall are evaluated with satisfying acceptance criteria. Moreover, the results of the numerical simulation correspond with those of experiment performed for the real size mockup. (author)

  6. ANL ITER high-heat-flux blanket-module heat transfer experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.E.

    1992-02-01

    An Argonne National Laboratory facility for conducting tests on multilayered slab models of fusion blanket designs is being developed; some of its features are described. This facility will allow testing under prototypic high heat fluxes, high temperatures, thermal gradients, and variable mechanical loadings in a helium gas environment. Steady and transient heat flux tests are possible. Electrical heating by a two-sided, thin stainless steel (SS) plate electrical resistance heater and SS water-cooled cold panels placed symmetrically on both sides of the heater allow achievement of global one-dimensional heat transfer across blanket specimen layers sandwiched between the hot and cold plates. The heat transfer characteristics at interfaces, as well as macroscale and microscale thermomechanical interactions between layers, can be studied in support of the ITER engineering design effort. The engineering design of the test apparatus has shown that it is important to use multidimensional thermomechanical analysis of sandwich-type composites to adequately analyze heat transfer. This fact will also be true for the engineering design of ITER.

  7. ANL ITER high-heat-flux blanket-module heat transfer experiments. Fusion Power Program

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.E.

    1992-02-01

    An Argonne National Laboratory facility for conducting tests on multilayered slab models of fusion blanket designs is being developed; some of its features are described. This facility will allow testing under prototypic high heat fluxes, high temperatures, thermal gradients, and variable mechanical loadings in a helium gas environment. Steady and transient heat flux tests are possible. Electrical heating by a two-sided, thin stainless steel (SS) plate electrical resistance heater and SS water-cooled cold panels placed symmetrically on both sides of the heater allow achievement of global one-dimensional heat transfer across blanket specimen layers sandwiched between the hot and cold plates. The heat transfer characteristics at interfaces, as well as macroscale and microscale thermomechanical interactions between layers, can be studied in support of the ITER engineering design effort. The engineering design of the test apparatus has shown that it is important to use multidimensional thermomechanical analysis of sandwich-type composites to adequately analyze heat transfer. This fact will also be true for the engineering design of ITER.

  8. 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

  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. 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)

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

    International Nuclear Information System (INIS)

    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

  12. MIT LMFBR blanket research project. Quarterly progress report, January 1, 1976--March 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J.

    1976-01-01

    Progress in the experimental and theoretical investigation of LMFBR breeding blanket design parameters is reported. State-of-the-art approaches for the calculation of gamma heating in the core, blanket, and reflector regions of LMFBR's have been evaluated, with particular emphasis on coupled neutron-gamma methods cross section sets. The effects of heterogeneity on resonance self-shielding were examined for the blanket region and the capture reaction in /sup 238/U. (DG)

  13. 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.

  14. 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.

  15. Structural optimization of the blanket first wall to reduce thermal stress using the Taguchi method

    International Nuclear Information System (INIS)

    The first wall of a fusion reactor blanket faces the core plasma directly. The first wall endures high heat loads that lead to high thermal stresses. To ensure the reliability of the first wall structure, it is desirable to reduce the thermal stress. In this study, structural optimization of the blanket first wall was carried out using the Taguchi method. The finite element method was used to conduct a numerical simulation to investigate the thermo-mechanical responses of the blanket first wall. The optimal configuration of the blanket first wall was derived. (author)

  16. 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

  17. 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)

  18. 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.)

  19. European reference design of the water-cooled lithium-lead blanket for a demonstration reactor

    International Nuclear Information System (INIS)

    The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European Union for DEMO-relevant design and R ampersand D activities. This paper gives a presentation of the reference conceptual design for water-cooled Pb-17Li DEMO blankets and an overview on the results of its performance assessments. Moreover, a critical discussion about the technical aspects requiring further improvements and/or modifications is performed taking into account the present status of the associated R ampersand D. This concept appears to be a very promising candidate for a DEMO reactor breeding blanket

  20. 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.