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Sample records for blanket structural materials

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. Waste management of first wall and blanket structural materials for tokamak fusion reactors

    International Nuclear Information System (INIS)

    A comparison has been made of the induced radioactivities in the first wall and structural materials of the breeder blanket in the high-flux region for two different fusion-reactor types. One system is the STARFIRE, a tokamak reactor with PCA, a modified stainless steel, as a first wall and a LiAlO2 breeder blanket; the other is a reactor based on the STARFIRE design with a vanadium alloy as the first wall and structural material, and circulating molten lithium as the breeder/coolant. The recycling or disposal of these structural materials is evaluated

  4. 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

  5. Tauro: a ceramic composite structural material self-cooled Pb-17Li breeder blanket concept

    International Nuclear Information System (INIS)

    The use of a low-activation (LA) ceramic composite (CC) as structural material appears essential to demonstrate the potential of fusion power reactors for being inherently or, at least, passively safe. Tauro is a self-cooled Pb-17Li breeder blanket with a SiC/SiC composite as structure. This study determines the required improvements for existing industrial LA composites (mainly SiC/SiC) in order to render them acceptable for blanket operating conditions. 3D SiC/SiC CC, recently launched on the market, is a promising candidate. A preliminary evaluation of a possible joining technique for SiC/SiC is also described. (orig.)

  6. 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

  7. 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.)

  8. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)] [and others

    1996-04-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 250{degrees}C. These specimens have been tested over a temperature range from 20 to 250{degrees}C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenic stainless steels, but the toughness remains quite high. The toughness decreases as the temperature increases. Irradiation at 250{degrees}C is more damaging that at 90{degrees}C, causing larger decreases in the fracture toughness. The ferritic-martensitic steels HT-9 and F82H show significantly greater reductions in fracture toughness that the austenitic stainless steels.

  9. STRUCTURAL DESIGN CRITERIA FOR TARGET/BLANKET SYSTEM COMPONENT MATERIALS FOR THE ACCELERATOR PRODUCTION OF TRITIUM PROJECT

    International Nuclear Information System (INIS)

    The design of target/blanket system components for the Accelerator Production of Tritium (APT) plant is dependent on the development of materials properties data specified by the designer. These data are needed to verify that component designs are adequate. The adequacy of the data will be related to safety, performance, and economic considerations, and to other requirements that may be deemed necessary by customers and regulatory bodies. The data required may already be in existence, as in the open technical literature, or may need to be generated, as is often the case for the design of new systems operating under relatively unique conditions. The designers' starting point for design data needs is generally some form of design criteria used in conjunction with a specified set of loading conditions and associated performance requirements. Most criteria are aimed at verifying the structural adequacy of the component, and often take the form of national or international standards such as the ASME Boiler and Pressure Vessel Code (ASME B and PV Code) or the French Nuclear Structural Requirements (RCC-MR). Whether or not there are specific design data needs associated with the use of these design criteria will largely depend on the uniqueness of the conditions of operation of the component. A component designed in accordance with the ASME B and PV Code, where no unusual environmental conditions exist, will utilize well-documented, statistically-evaluated developed in conjunction with the Code, and will not be likely to have any design data needs. On the other hand, a component to be designed to operate under unique APT conditions, is likely to have significant design data needs. Such a component is also likely to require special design criteria for verification of its structural adequacy, specifically accounting for changes in materials properties which may occur during exposure in the service environment. In such a situation it is common for the design criteria and

  10. STRUCTURAL DESIGN CRITERIA FOR TARGET/BLANKET SYSTEM COMPONENT MATERIALS FOR THE ACCELERATOR PRODUCTION OF TRITIUM PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    W. JOHNSON; R. RYDER; P. RITTENHOUSE

    2001-01-01

    The design of target/blanket system components for the Accelerator Production of Tritium (APT) plant is dependent on the development of materials properties data specified by the designer. These data are needed to verify that component designs are adequate. The adequacy of the data will be related to safety, performance, and economic considerations, and to other requirements that may be deemed necessary by customers and regulatory bodies. The data required may already be in existence, as in the open technical literature, or may need to be generated, as is often the case for the design of new systems operating under relatively unique conditions. The designers' starting point for design data needs is generally some form of design criteria used in conjunction with a specified set of loading conditions and associated performance requirements. Most criteria are aimed at verifying the structural adequacy of the component, and often take the form of national or international standards such as the ASME Boiler and Pressure Vessel Code (ASME B and PV Code) or the French Nuclear Structural Requirements (RCC-MR). Whether or not there are specific design data needs associated with the use of these design criteria will largely depend on the uniqueness of the conditions of operation of the component. A component designed in accordance with the ASME B and PV Code, where no unusual environmental conditions exist, will utilize well-documented, statistically-evaluated developed in conjunction with the Code, and will not be likely to have any design data needs. On the other hand, a component to be designed to operate under unique APT conditions, is likely to have significant design data needs. Such a component is also likely to require special design criteria for verification of its structural adequacy, specifically accounting for changes in materials properties which may occur during exposure in the service environment. In such a situation it is common for the design criteria

  11. Prospects and problems using vanadium alloys as a structural material of the first wall and blanket of fusion reactors

    International Nuclear Information System (INIS)

    Vanadium-based alloys are most promising as low activation structural materials for DEMO. It was previously established that high priority is to be given to V-alloys of the V-Ti-Cr system as structural materials of a tritium breeding blanket and the first wall of a fusion reactor. However, there is some uncertainty in selecting a specific element ratio between the alloy components in this system. This is primarily explained by the fact that the properties of V-alloys are dictated not only by the ratio between the main alloying elements (here Ti and Cr), but also by impurities, both metallic and oxygen interstitials. Based on a number of papers today one can say that V-Ti-Cr alloys with insignificant variations in the contents of the main constituents within 5-10 mass% Ti and 4-6 mass% Cr must be taken as a base for subsequent optimization of chemical composition and thermomechanical working. However, the database is obviously insufficient to assess the ecological acceptability (activation), physical and mechanical properties, corrosion and irradiation resistance and, particularly, the commercial production of alloys. Therefore, there is a need for comprehensive studies of promising V-alloys, namely V-4Ti-4Cr and V-10Ti-5Cr. (orig.)

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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.

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

    International Nuclear Information System (INIS)

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

  4. US/Japan collaborative program on fusion reactor materials: Summary of the tenth DOE/JAERI Annex I technical progress meeting on neutron irradiation effects in first wall and blanket structural materials

    International Nuclear Information System (INIS)

    This meeting was held at Oak Ridge National Laboratory on March 17, 1989, to review the technical progress on the collaborative DOE/JAERI program on fusion reactor materials. The purpose of the program is to determine the effects of neutron irradiation on the mechanical behavior and dimensional stability of US and Japanese austenitic stainless steels. Phase I of the program focused on the effects of high concentrations of helium on the tensile, fatigue, and swelling properties of both US and Japanese alloys. In Phase II of the program, spectral and isotropic tailoring techniques are fully utilized to reproduce the helium:dpa ratio typical of the fusion environment. The Phase II program hinges on a restart of the High Flux Isotope Reactor by mid-1989. Eight target position capsules and two RB* position capsules have been assembled. The target capsule experiments will address issues relating to the performance of austenitic steels at high damage levels including an assessment of the performance of a variety of weld materials. The RB* capsules will provide a unique and important set of data on the behavior of austenitic steels irradiated under conditions which reproduce the damage rate, dose, temperature, and helium generation rate expected in the first wall and blanket structure of the International Thermonuclear Experimental Reactor

  5. 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)

  6. 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.)

  7. Convertible liquid metal blankets for ITER with Pb-17Li as breeding material

    International Nuclear Information System (INIS)

    A convertible blanket concept is proposed for ITER, where, without replacement of the blanket structure, a non-breeding Pb alloy is used during the basic performance phase and the eutectic Pb-17Li during the enhanced performance phase. The concept is based on austenitic steel as structural material, an average neutron wall load of 1MWm-2 and either helium or water as coolant. The same design concept was used for both coolant options with respect to a stiff blanket segment box, direct cooling of the first wall using toroidal ducts, poloidal hairpin tubes to cool the quasi-stagnant liquid metal and tritium removal outside the vacuum vessel.Various design options were considered for the first-wall and pool cooling and corresponding headers. Owing to the different coolant properties, different combinations were selected for the two versions. The performance of the two versions was assessed among other things with respect to tritium breeding and control, reliability and R and D needs. (orig.)

  8. 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.

  9. Damage Identification of Continuum Structures Using Blanketing Effect

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. [Argonne National Lab., IL (United States); Wu, C.H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team; Koroda, T. [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan); Shatalov, G. [Kurchatov Inst. of Atomic Energy, Moscow (USSR)

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. (Argonne National Lab., IL (United States)); Wu, C.H. (Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team); Koroda, T. (Japan Atomic Energy Research Inst., Ibaraki-ken (Japan)); Shatalov, G. (Kurchatov Inst. of Atomic Energy, Moscow (USSR))

    1992-01-01

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

  13. 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)

  14. 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.)

  15. Transient electromagnetic and dynamic structural analyses of a blanket structure with coupling effects

    International Nuclear Information System (INIS)

    Transient electromagnetic and dynamic structural analyses of a blanket structure in the fusion experimental reactor (FER) under a plasma disruption event and a vertical displacement event (VDE) have been performed to investigate the dynamic structural characteristics and the feasibility of the structure. Coupling effects between eddy currents and dynamic deflections have also been taken into account in these analyses. In this study, the inboard blanket was employed because of our computer memory limitation. A 1/192 segment model of a full torus was analyzed using the analytical code, EDDYCUFF. In the plasma disruption event, the maximum magnetic pressure caused by eddy currents and poloidal fields was 1.2MPa. The maximum stress intensity by this magnetic pressure was 114MPa. In the VDE, the maximum magnetic pressure was 2.4MPa and the maximum stress intensity was 253MPa. This stress was somewhat beyond the allowable stress limit. Therefore, the blanket structure and support design should be reviewed to reduce the stress to a suitable value. In summary, the dynamic structural characteristics and design issues of the blanket structure have been identified. (orig.)

  16. Acoustic contributions of a sound absorbing blanket placed in a double panel structure: Absorption Versus Transmission

    CERN Document Server

    Doutres, Olivier; 10.1121/1.3458845

    2010-01-01

    The objective of this paper is to propose a simple tool to estimate the absorption vs. transmission loss contributions of a multilayered blanket unbounded in a double panel structure and thus guide its optimization. The normal incidence airborne sound transmission loss of the double panel structure, without structure-borne connections, is written in terms of three main contributions; (i) sound transmission loss of the panels, (ii) sound transmission loss of the blanket and (iii) sound absorption due to multiple reflections inside the cavity. The method is applied to four different blankets frequently used in automotive and aeronautic applications: a non-symmetric multilayer made of a screen in sandwich between two porous layers and three symmetric porous layers having different pore geometries. It is shown that the absorption behavior of the blanket controls the acoustic behavior of the treatment at low and medium frequencies and its transmission loss at high frequencies. Acoustic treatment having poor sound ...

  17. Gas permeability of SiC/SiC composite as blanket material of fusion reactor

    International Nuclear Information System (INIS)

    Gas permeability of SiC/SiC composite materials, which is one of the most important properties in application of SiC/SiC composite for first wall and blanket of fusion reactors, was measured by using a vacuum apparatus. The cylindrical SiC/SiC composite specimens were prepared by three different processes. The measurement on permeability for three materials was carried out with helium gas pressure ranging from 102 to 105 Pa at room temperature. The pressure in bottom chamber down stream of specimens increased with the helium gas pressure within the applied pressure range. The helium gas flow through the material is regarded as molecular flow. The material made by PIP method showed the highest permeability. The lowest permeability was observed in the one made by PIP followed by RS method. The material, SA-TyrannoHexTM made by hot pressing was in the second position. The difference of the permeability can be related with the macroscopic structure represented by pores and cracks. (author)

  18. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 3. Comprehending the blanket structure

    International Nuclear Information System (INIS)

    The functions and structure design of shield blanket of ITER and test blanket module (TBM) are stated. The design of shield blanket of ITER is finished and beginning its supply. ITER shield blanket consists of the first wall and shielding block made of SUS316LN-IG, of which design and joint method are explained in details. TBM is used for engineering test of nuclear fusion reactor. The fuel breeding function, power function and structure design of TBM are described. The cooling conditions of shield blanket, TBM and fusion reactor, structure of TBM, structure of the first wall, results of stress analysis of TBM first wall, heat and mechanical behavior measurement device of pebble packed layer, the effective thermal conductivity, and relation between stress and compressive strain of Li2TiO3 pebble packed layer, and distribution of Tresca stress on the first wall are illustrated. (S.Y.)

  19. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%

  20. 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

  1. Tritium self-sufficiency time and inventory evolution for solid-type breeding blanket materials for DEMO

    Science.gov (United States)

    Packer, L. W.; Pampin, R.; Zheng, S.

    2011-10-01

    One of the primary functions of a fusion blanket is to generate enough tritium to make a fusion power plant (FPP) self-sufficient. To ensure that there is satisfactory tritium production in a real plant the tritium breeding ratio (TBR) in the blanket must be greater than 1 + M, where M is the breeding margin. For solid-type blanket designs, the initial TBR must be significantly higher than 1 + M, since the blanket TBR will be reduced over time as the lithium fuel is consumed. The rate of TBR reduction will impact on the overall blanket self-sufficiency time, the time in which the net tritium inventory of the system is positive. DEMO relevant blanket materials, Li 4SiO 4 and Li 2TiO 3, are investigated by computational simulation using radiation transport tools coupled with time-dependent inventory calculations. The results include tritium inventory assessments and depletion of breeding materials over time, which enable self-sufficiency times and maximum surplus tritium inventories to be evaluated, which are essential quantities to determine to allow one to design a credible FPP using solid-type breeding material concepts. The blanket concepts investigated show self-sufficiency times of several years in some cases and maximum surplus inventories of up to a few tens of kg.

  2. Structural materials assessment

    International Nuclear Information System (INIS)

    The selection of first wall and structural materials is strongly dependent on the proposed design of breeding blanket components and the targets for a fusion reactor development. An envelope of parameters which have to be covered in future R and D activities and which have been adapted in different proposals has been compiled. A short description of interesting material groups like ferritic-martensitic steels, vanadium alloys and ceramic composites, major criteria for their selection and a survey on existing irradiation data is given. This is followed by a comparative assessment of relevant properties and an identification of major issues for each material group. A more detailed proposal for the future R and D activities is then developed for the ferritic-martensitic steels, the present reference material for the European Breeding Blankets. It describes different phases of development necessary for the qualification of this material for DEMO and gives time schedules which are compatible with parallel component developments. A more selective strategy is proposed for the development of vanadium alloys and the ceramic composite material SiC/SiC. For these alternatives work should be concentrated on identified high-risk issues, before a comprehensive development programme is started. The necessity of efficient irradiation facilities to study the irradiation behaviour of the materials under simulation - and realistic fusion conditions is discussed. The availability of high flux fission reactors and necessary extensions of irradiation rigs for the next decade is stressed. Finally it is shown that for the qualification of materials under realistic fusion conditions a high-energetic, high-flux neutron source is mandatory. An accelerator-driven d-Li neutron source (IFMIF) can fulfil essential users requirements as test bed for materials and can technically be made available in due time. In combination with ITER and DEMO, where a concept verification and full scale

  3. Assessment of titanium for use in the 1st wall/blanket structure of fusion power reactors

    International Nuclear Information System (INIS)

    This report describes a portion of the work that was performed as part of a First Wall/Blanket Systems Analysis Study. The objective of this part of the study was to assess the suitability of using titanium alloys in the first wall/blanket structure of commercial controlled thermonuclear reactors (CTR). While the purpose of this study was not to recommend a specific titanium alloy, but to examine titanium alloys, in general, two near-alpha titanium alloys were selected for an indepth examination. These alloys were Ti-6Al-4V and Ti-6Al-2Sn-4Zr-2Mo. Using properties important to the CTR first wall/blanket structures application, these titanium alloys were compared with five other candidate structural materials (2219 aluminum, 316 stainless steel, V-20 Ti, Nb-1Zr, and Mo-0.5 Ti-0.08 Zr (TZM)). The results of this study revealed that titanium offers potential for use in a CTR from strength, minimum radioactivity, and resources standpoints and should be considered in future fusion reactor studies

  4. 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

  5. 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)

  6. 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.)

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

    International Nuclear Information System (INIS)

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

  8. 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)

  9. Microstructure and hardness of HIP-bonded regions in F82H blanket structures

    International Nuclear Information System (INIS)

    Metallurgical examinations and hardness measurements were performed at hot isostatic pressing (HIP)-bonded regions in blanket structures made from F82H alloy in order to investigate the HIP-bondability and the influence on the microstructure due to the HIP and heat treatments which would correspond to the fabrication of an actual blanket. The metallurgical examination showed that the HIP-bonded interfaces were sufficiently diffusion-bonded without significant defects, i.e. voids and/or exfoliations, although grain coarsening was observed at a part of the HIP interfaces. Hardness was nearly equal in the coarsening region and a region without coarsening, but about a 10 Hv increase was found in a boundary in between the regions with and without coarsening. Microcrystallized grains were observed in a region about ∼6 μm from HIP interfaces, and the hardness increased by about 0.2 GPa in the region

  10. 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

  11. 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.

  12. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    International Nuclear Information System (INIS)

    Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF--BeF2, Pb--Li alloys, and solid ceramic compounds such as Li2O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies

  13. 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)

  14. ITER breeding blanket module design and analysis

    International Nuclear Information System (INIS)

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

  15. 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

  16. Stress and lifetime calculations for first wall and blanket structural components. Pt. 1

    International Nuclear Information System (INIS)

    In this report the lifetime of first wall and blanket structures of fusion reactors is investigated. The extension of small pre-existing cracks by the cyclic operation of a fusion reactor seems to be the most important failure mode. The special application of the present investigation are tubes acting directly as parts of the first wall and affected by various radiation effects. The outer surface is asymmetrically heated and by combination of thermal extension, swelling, irradiation creep and internal pressure a complex time dependent stress distribution results. Crack growth until failure caused by cyclic operation of the reactor is computed by application of fracture mechanical methods. (orig.)

  17. 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.

  18. A computational procedure for coupled electromagnetic-structural dynamic problems and its application to a fusion reactor blanket

    International Nuclear Information System (INIS)

    A method is presented in order to couple quasistationary electromagnetics and the dynamics of structure and fluid. This method allows to compute forces, strains and stresses in structures subjected to transient magnetic fields. An important application was to determine the dynamic loading of the self-cooled liquid metal blanket during a plasma distruption. (orig./HP)

  19. 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

  20. 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

  1. 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)

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

    International Nuclear Information System (INIS)

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

  3. Investigation of tritium inventory and permeation behaviour in the liquid breeder blanket concept of Demo as a function of design and material parameters

    International Nuclear Information System (INIS)

    A numerical code has been used to estimate the time dependence of tritium inventory and of tritium transport into the coolant, into the first wall boxes and through the liquid breeder in the Pb-17Li blanket concept of DEMO. Several issues in both design and material parameters have been considered and the effect on inventory and permeation of coatings with low surface recombination coefficient and/or low diffusivity at various surfaces of the structural material has been studied. TiC has been chosen as reference material for these calculations and a general database on coating efficiency as a function of its properties has also been produced on the basis of TiC data

  4. 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

  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. Structural design and preliminary analysis of liquid lead–lithium blanket for China Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    China Fusion Engineering Test Reactor (CFETR) has been proposed as an option in China to bridge the gaps between ITER and fusion power plant. Since one major goal of CFETR is to demonstrate long pulse or steady-state operation with duty cycle time ≥0.3–0.5, easier maintenance of the in-vessel components is emphasized in the design process. In this contribution, a kind of liquid lead–lithium tritium breeder blanket concept focus on the remote maintenance has been designed for CFETR. To make the pipes and mechanical connections at the rear of the blanket accessible from vacuum vessel, two kinds of guide tubes were adopted to provide passageways for remote handling tools. In order to evaluate the effects of the guide tube installation on the structural performance of the blanket, as a preliminary stage, thermal-hydraulic analysis of first wall was carried out based on the heat load obtained from 3D modeled neutronics calculations. In addition, thermal stress analysis of the first wall under normal condition was performed to evaluate the thermomechanical behavior. The preliminary analysis results validated the performance of current blanket design

  7. Structural design and preliminary analysis of liquid lead–lithium blanket for China Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ni, Muyi; Lian, Chao; Zhang, Shichao; Nie, Baojie; Jiang, Jieqiong, E-mail: jieqiong.jiang@fds.org.cn

    2015-05-15

    China Fusion Engineering Test Reactor (CFETR) has been proposed as an option in China to bridge the gaps between ITER and fusion power plant. Since one major goal of CFETR is to demonstrate long pulse or steady-state operation with duty cycle time ≥0.3–0.5, easier maintenance of the in-vessel components is emphasized in the design process. In this contribution, a kind of liquid lead–lithium tritium breeder blanket concept focus on the remote maintenance has been designed for CFETR. To make the pipes and mechanical connections at the rear of the blanket accessible from vacuum vessel, two kinds of guide tubes were adopted to provide passageways for remote handling tools. In order to evaluate the effects of the guide tube installation on the structural performance of the blanket, as a preliminary stage, thermal-hydraulic analysis of first wall was carried out based on the heat load obtained from 3D modeled neutronics calculations. In addition, thermal stress analysis of the first wall under normal condition was performed to evaluate the thermomechanical behavior. The preliminary analysis results validated the performance of current blanket design.

  8. Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

    International Nuclear Information System (INIS)

    The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.

  9. 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.

  10. 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.

  11. A method for measuring the corrosion rate of materials in spallation neutron source target/blanket cooling loops

    International Nuclear Information System (INIS)

    This paper summarizes the ongoing evaluation of the susceptibility of materials in accelerator target/blanket cooling loops to corrosion. To simulate the exposure environment in a target/blanket cooling loop, samples were irradiated by an 800 MeV proton beam at the A6 Target Station of the Los Alamos Neutron Science Center (LANSCE). To accomplish this, a cooling water loop capable of exposing corrosion samples to an 800 MeV proton beam at currents upwards of 1 mA was constructed. This loop allowed control and evaluation hydrogen water chemistry, water conductivity, and solution pH. Specially designed ceramic sealed samples were used to measure the real-time corrosion rates of materials placed directly in the proton beam using electrochemical impedance spectroscopy (EIS). EIS was also used to measure real-time corrosion rates of samples that were out of the proton beam and downstream from the in-beam samples. These out-of-beam probes primarily examined the effects of long lived water radiolysis products from proton irradiation on corrosion rates. An overview of the LANSCE corrosion loop, the corrosion probes, and data from an in-beam alloy 718 probe are presented

  12. 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)

  13. 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.)

  14. 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

  15. Current material and design studies for the utilization of Li2ZrO3 in the BIT blanket concept

    International Nuclear Information System (INIS)

    The DEMO-relevant, helium-cooled, BIT blanket concept jointly developed by CEA and ENEA has been originally designed on the basis of LiAlO2 properties data. Owing to its excellent tritium release behaviour Li2ZrO3 is being evaluated too. Thus, development of the Li2ZrO3 ceramic is in progress at CEA and conceptual work is partly devoted to the adaptation of the design using Li2ZrO3 characteristics. The material development work is focused on improving the thermomechanical behaviour of Li2ZrO3 with the aim to ensure ceramic pellet integrity during blanket operation. Such an improvement should be achieved without significant change of the tritium release characteristics. Among means being attempted is microstructural tailoring, i.e. density adjustment. Mechanical tests, thermal cycling tests and tritium release annealing tests allow to evidence any changes. A feasibility study of large scale production of BIT-type Li2ZrO3 pellets was undertaken with industry. In parallel, conceptual work is concentrated on redesigning the BIT concept with Li2ZrO3 properties data, on calculation of tritium breeding ratio, and on evaluation of the maximum lithium burnup in the breeder. ((orig.))

  16. 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

  17. An examination into weldability of irradiated material by a laser welding method for repair of ITER blanket

    International Nuclear Information System (INIS)

    SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of ITER (International Thermonuclear Experimental Reactor). This paper describes an examination into weldability of irradiated and un-irradiated SS316L(N)-IG for coolant piping and investigation of the mechanical properties of welding joints as well as the effect of helium generation on weldability. A YAG laser welding process was used for repair welding of the water cooling branch pipelines. It was clarified that welding of SS316L(N)-IG irradiated up to 0.3 dpa (He content: about 3 appm) can be carried out without a significant deterioration of tensile properties due to helium accumulation. Therefore, repair of coolant pipes for the ITER blanket can be performed by the laser welding process. (author)

  18. Durability evaluation of photovoltaic blanket materials exposed on LDEF tray S1003

    Science.gov (United States)

    Rutledge, Sharon K.; Olle, Raymond M.

    1991-06-01

    Several candidate protective coatings on Kapton and uncoated Kapton were exposed to the LEO environment on the LDEF in order to determine whether the coatings could be used to protect polymeric substrates from degradation in the LEO environment. These materials are used for flexible solar array panels in which the polymer is the structural member that supports the solar cell and current carriers. Arrays such as these are used on the Hubble Space Telescope and will be used on Space Station Freedom. The results of the experiments are presented.

  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. 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...

  2. 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)

  3. Effect of heat cycling on microstructure and thermal property of boron carbide sintered bulk as a shielding material for fusion blanket

    International Nuclear Information System (INIS)

    In the Force Free Helical Reactor (FFHR) design activity in NIFS, metallic carbides and hydrides are considered as candidate shielding materials for the fusion blankets. These materials are expected to have some advantages on neutronic and thermo-physical properties. In order to promote the blanket design, it is necessary to clarify thermal properties of the candidate materials. We studied microstructure and thermal property of boron carbide (B4C), which is one of the promising candidates shielding materials, including the effect of heat cycling. By the laser-flash method, thermal diffusivity, which is one of the properties necessary for evaluating thermal conductivity, was measured precisely for B4C samples. The thermal diffusivity of B4C around 200degC decreased to 1/3 (5 × 10-6 m2 S-1) compared with that at room temperature. The sintering density of B4C bulk was decreased slightly by the thermal cycling. It was suggested that the B4C bulk has high thermal stability and soundness of microstructure during the life-time of blanket system. (author)

  4. 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

  5. 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.)

  6. Auxetic materials and structures

    CERN Document Server

    Lim, Teik-Cheng

    2015-01-01

    This book describes the fundamentals of the mechanics and design of auxetic solids and structures, which possess a negative Poisson’s ratio. It will benefit two groups of readers: (a) industry practitioners, such as product and structural designers, who need to control mechanical stress distributions using auxetic materials, and (b) academic researchers and students who intend to produce structures with unique mechanical and other physical properties using auxetic materials.

  7. 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.)

  8. 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

  9. Mechanical properties of materials in fusion reactor first-wall and blanket systems

    International Nuclear Information System (INIS)

    With respect to the effects of irradiation on mechanical properties, the most significant difference between fast fission and fusion reactor spectra is the relatively large amount of helium produced by (n,α) transmutations in the latter. Relevant information on the effects of large amounts of helium (with concomitant displacement damage) comes from irradiation of alloys containing nickel in mixed spectrum reactors. At helium levels of interest for fusion reactor development, properties are degraded to unacceptable levels above Tm/2. Below this temperature, strength and ductility are retained and fractures remain transgranular. Importantly, the properties remain sensitive to composition and structure. A comparison of the response of bcc refractory alloys to that of stainless steel at equivalent damage levels shows the same general trends in properties with homologous temperature. The refractory alloys do offer potential for higher temperature applications because of their melting temperatures

  10. Analysis of the multiplication of D-T neutron sources in blanket materials of fusion reactors

    International Nuclear Information System (INIS)

    A D-T neutron source is amplified when emitted into a body of material with appreciable (n,2n), (n,3n) or (n,f) cross sections. This amplification is described by a simple theory, approximating the strict integral transport description of the process. The distribution of neutrons in energy, from 14 MeV down to the (n,2n) threshold, is approximated by effective one-group cross sections for amplifiers of high and medium mass numbers; two-group cross sections are needed for Be. The spatial character of the multiplication is described by average collision probabilities for non-flat collision sources. The probabilities are approximated for spherical shell geometry with a small number of geometrical parameters. The theory enables a very accurate determination of sigmasub(n,2n) + 2sigmasub(n,3n) + (vsub(f)-1)sigmasub(f) at the source energy from measurements of total multiplications. If total leakages above the (n,2n) threshold are also measured, then the hardness of the secondary neutron spectra can be estimated. The accuracy of the approximate theory was ascertained by energy-space detailed transport comparison calculations for Be, Cu, Zr, Fe, Pb and U238. (orig.)

  11. Durability evaluation of photovoltaic blanket materials exposed on LDEF tray S1003

    International Nuclear Information System (INIS)

    Several candidate protective coatings on Kapton and uncoated Kapton were exposed to the low Earth orbital (LEO) environment on the Long Duration Exposure Facility (LDEF) to determine if the coatings could be used to protect polymeric substrates from degradation in the LEO environment. The coatings that were evaluated were 700 A of aluminum oxide, 650 A of silicon dioxide, and 650 A of a 4 percent polytetrafluoroethylene-96 percent silicon dioxide mixed coating. All of the coatings evaluated were ion beam sputter deposited. These materials were exposed to a very low atomic oxygen fluence (4.8 x 10 exp 19 atoms/sq. cm) as a result of the experiment tray being located 98 degrees from the ram direction. As a result of the low atomic oxygen fluence, determination of a change in mass was not possible for any of the samples including the uncoated Kapton. There was no evidence of spalling of any of the coatings after the approximately 33,600 thermal cycles recorded for LDEF. The surface of the uncoated Kapton, however, did show evidence of grazing incidence texturing. There was a 7-8 percent increase in solar absorptance for the silicon dioxide and aluminum oxide coated Kapton and only a 4 percent increase for the mixed coating. It appears that the addition of a small amount of fluoropolymer may reduce the magnitude of absorptance increase due to environmental exposure. Thermal emittance did not change significantly for any of the exposed samples. Scanning electron microscopy revealed few micrometeoroid or debris impacts, but the impact sites found indicated that the extent of damage or cracking of the coating around the defect site did not extend beyond a factor of three of the impact crater diameter. This limiting of impact damage is of great significance for the durability of thin film coatings used for protection against the LEO environment

  12. Blanket concept of water-cooled lithium lead with beryllium for the SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    As an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. In SlimCS, the net tritium breeding ratio (TBR) supplied from WCLL blanket was not enough because the thickness of blanket in SlimCS was limited to about 0.5 m so as to allocate the conducting shell position near the plasma for high beta access and vertical stability of plasma. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier to reach a required TBR (≥ 1.05). Considering the operating temperature of blanket materials, a double pipe structure was adopted. The nuclear and thermal analysis were carried out by a nuclear-thermal-coupled code, ANIHEAT and DOHEAT so that blanket materials were appropriately arranged to satisfy the acceptable operation temperatures. The temperatures of materials were kept in appropriate range for the neutron wall load Pn = 5 MW/m2. It was found that the local TBR of WCLL with Be blanket was comparable with that of solid breeder blanket. (author)

  13. 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

  14. 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)

  15. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    International Nuclear Information System (INIS)

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers

  16. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Manly, W.D. [Oak Ridge National Lab., TN (United States); Dombrowski, D.E. [Brush Wellman, Inc., Cleveland, OH (United States)] [and others

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers.

  17. 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

  18. Structural and Material Instability

    DEFF Research Database (Denmark)

    Cifuentes, Gustavo Cifuentes

    This work is a small contribution to the general problem of structural and material instability. In this work, the main subject is the analysis of cracking and failure of structural elements made from quasi-brittle materials like concrete. The analysis is made using the finite element method. Three...... program based on the finite element method for the analysis of cracks in structural elements is presented; in this program the interface and elements with embedded discontinuities are implemented....... use of interface elements) is used successfully to model cases where the path of the discontinuity is known in advance, as is the case of the analysis of pull-out of fibers embedded in a concrete matrix. This method is applied to the case of non-straight fibers and fibers with forces that have...

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

    International Nuclear Information System (INIS)

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

  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. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. 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.)

  4. 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

  5. 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.)

  6. 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.

  7. 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

  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 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

  9. 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

  10. 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

  11. Towards a reduced activation structural materials database for fusion reactors

    International Nuclear Information System (INIS)

    Full text: The development of First Wall, Blanket and Divertor materials which are capable of withstanding many years the high neutron and heat fluxes, is a critical path to fusion power. Therefore, the timely availability of a sound materials database has become an indispensable element in international fusion road maps. In order to provide a related materials database for design, construction, licensing and safe operation of the ITER Test Blanket Modules and of a DEMO reactor, a wealth of R and D results on the European reduced activation ferritic-martensitic steel EUROFER, and on oxide dispersion strengthened (ODS) variants have become available, mainly in the temperature window 250-700 deg. C. Industrial EUROFER-batches of 3.5 and 8.0 tons have been produced with a variety of semi-finished, quality-assured product forms. Extensive chipless shaping and joining experience taking into account different welding procedures and powder technology product forms have demonstrated that EUROFER type steel complies with a wide range of established manufacturing processes. EUROFER is also resistant to high temperature aging, and the existing creep-rupture properties (∼30000 h) indicated long term stability and predictability. To increase the thermal efficiency of blankets beyond 45%, high temperature resistant SiCf/SiC channel inserts for liquid metal coolant tubes are developed. Mechanical and thermal properties of various SiCf/SiC composits have been measured after neutron radiation. Regarding radiation damage resistance of blanket structural materials, a broad based reactor irradiation programme counts several steps from 2 needs to be removed, the design is presently based on tiles made of W (∼2000 deg. C), as well as on structural materials like W-alloy (∼700-1300 deg. C) and RAF(M)-ODS steel (∼650 deg. C). Severe plastic deformation of pure W and W alloys improves ductility, but does not prevent from re-crystallisation between 850 and 1200 deg. C. For the

  12. Evaluation of Cortaderia selloana (Capim-dos-pampas) blankets as sorbent materials for oil spills in simulated hydro equipment; Estudo do desempenho de tecidos e mantas para utilizacao como sorventes para petroleo

    Energy Technology Data Exchange (ETDEWEB)

    Bonetti, T.F.; Sydenstricker, T.H.D. [Universidade Federal do Parana (UFPR), Curitiba, PR (Brazil)], e-mail: thais@demec.ufpr.br; Amico, S.C. [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil)

    2006-07-01

    Oil spills in aquatic environments may cause serious economy losses and severe environmental impact which both drive the development of commercial systems (e.g. sorbents) to control these accidents. One way of using sorbents is to encapsulate them with an involucre or cover, i.e. producing blankets. The focus of this research is to evaluate the key characteristics of interest (aerial density, water and oil sorption, mechanical strength and cost) of different materials to use as covers for blankets and to prepare blankets and compare their performance when made with various core materials, such as Cortaderia selloana fibers and different commercial sorbents. A simulated aqueous body with stream was used for the sorption experiments, where the oil and water phases were circulated and forced to pass under the blankets. On the sorption tests, the fibers of Cortaderia selloana reached a performance lower to that of commercial sorbents, mainly due to their low density and high volume (difficult packing), nevertheless a clear trend was noted, heavier blankets with higher sorption periods lead to higher sorption. (author)

  13. 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

  14. Lifetime predictions for the first wall and blanket structure of fusion reactors

    International Nuclear Information System (INIS)

    Lifetime analysis of the first wall including the divertor and limiter is an important subject for the design of fusion reactors. These components are exposed to severe mechanical, thermal, and irradiation effects that limit their useful structural life, and their design lifetime has a large influence on the selection of major reactor design parameters. Particular attention (at the meeting whose papers are included in this report) was given to different approaches and models for the prediction of component lifetimes. Topics covered include life-limiting mechanisms, stress analysis and lifetime evaluation, and erosion and deposition effects

  15. Nano structured Magnetic Materials

    International Nuclear Information System (INIS)

    The saga of nanostructured magnetic materials (NMMs) has prevailed since the discovery of the first giant magnetoresistance (GMR) effect in metals in 1988. NMMs represent a unique system that incorporates the interplay between the properties associated with spin degrees of freedom and the nanoscaled structures, which provide a very strong platform for exploring both basic science and technical applications in the fields of solid-state physics, chemistry, materials science, and engineering. In fact, an active research field called “spintronics,” which has a big overlap with NMMs, has emerged and prevailed very recently. Through manipulation of spin of electrons in solids, a wide variety of NMMs and devices have been playing a prominent role in information processing and transport in our modern life. A rich variety of materials, such as transition metals, manganite, wide bandgap semiconductors, and nanocomposites, have already been developed for generating well-controlled nanostructures with new functionalities. Many scientists believe that the 21st century will be a “Century of Spin.” Nanomaterials and nanotechnologies have provided historical opportunities for research and development of novel spintronics materials and devices. NMSs manifest fascinating properties compared to the bulks because of size effect and quantum effect. Nanotechnologies have been proven to be an effective way to fabricate devices with fine nanostructures. The combination of spintronics and nanomaterials will undisputedly open new pathways in solid-state physics. The present special issue focuses on the recent development in the understanding of the synthesis, the studies on magnetic properties of nanostructures, and their potential applications based on the multiple functionalities.

  16. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    International Nuclear Information System (INIS)

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits. (fusion engineering)

  17. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    Science.gov (United States)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

  18. Progress in the US program to develop low-activation structural materials for fusion

    International Nuclear Information System (INIS)

    It has long been recognized that attainment of the safety and environmental potential of fusion energy requires the successful development of low-activation materials for the first wall, blanket and other high heat flux structural components. Only a limited number of materials potentially possess the physical, mechanical and low-activation characteristics required for this application. The current US structural materials research effort is focused on three candidate materials: advanced ferritic steels, vanadium alloys, and silicon carbide composites. Recent progress has been made in understanding the response of these materials to neutron irradiation. (author)

  19. Progress in the U.S. program to develop low-activation structural materials for fusion

    International Nuclear Information System (INIS)

    It has long been recognized that attainment of the safety and environmental potential of fusion energy requires the successful development of low-activation materials for the first wall, blanket and other high heat flux structural components. Only a limited number of materials potentially possess the physical, mechanical and low-activation characteristics required for this application. The current U.S. structural materials research effort is focused on three candidate materials: advanced ferritic steels, vanadium alloys, and silicon carbide composites. Recent progress has been made in understanding the response of these materials to neutron irradiation. (author)

  20. 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.

  1. 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)

  2. 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

  3. 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)

  4. RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

    International Nuclear Information System (INIS)

    RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. Use of SiCf/SiC ceramic composites as structure material of a fusion reactor toroid internal components

    International Nuclear Information System (INIS)

    The use of low neutron-induced activation structural materials seems necessary in order to improve safety in future fusion power reactors. Among them, SiCf/SiC composites appear as a very promising solution because of their low activation characteristics coupled with excellent mechanical properties at high temperatures. With the main objective of evaluating the limit of present-day composites, a tritium breeding blanket using SiCf/SiC as structural material (the TAURO blanket) has been developed in the last years by the Commissariat a l'Energie Atomique (CEA). The purpose of this thesis was to modify the available design tools (computer codes, design criteria), normally used for the analyses of metallic structures, in order to better take into account the mechanical behaviour of SiCf/SiC. Alter a preliminary improvement of the calculation methods, two main topics of study could be identified: the modelling of the mechanical behaviour of the composite and the assessment of appropriate design criteria. The different behavioural models available in literature were analysed in order to find the one that was the best suited to the specific problems met in the field of fusion power. The selected model was then implemented in the finite elements code CASTEM 2000 used within the CEA for the thermo-mechanical analyses of the TAURO blanket. For the design of the blanket, we proposed a new resistance criterion whose main advantage, with respect to the other examined, lies in the easiness of identification. The suggested solutions were then applied in the design studies of the TAURO blanket. We then could show that the use of appropriate calculation methodologies is necessary in order to achieve a correct design of the blanket and a more realistic estimate of the limits of present day composites. The obtained results can also be extended to all nuclear components making use of SiCf/SiC structures. (author)

  10. 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

  11. Development, simulation and testing of structural materials for DEMO

    International Nuclear Information System (INIS)

    In DEMO the structural and functional materials of the in-vessel components will be exposed to a very intense flux of fusion neutrons with energies up to 14 MeV creating displacement cascades and gaseous transmutation products. Point defects and transmutations will induce new microstructures leading to changes in mechanical and physical properties such as hardening, swelling, loss of fracture toughness and creep strength. The kinetics of microstructural evolution depends on time, temperature and defect production rates. The structural materials to be used in DEMO should have very special properties: high radiation resistance up to the dose of 100 dpa, low residual activation, high creep strength and good compatibility with the cooling media in as wide a temperature operational window as possible for the achievement of high thermal efficiency. The most promising materials are: Reduced Activation Ferritic Martensitic (RAFM) steels (Eurofer and F82H), Oxide Dispersion Strengthened (ODS) RAFM and RAF steels, SiC fibres reinforced SiC matrix composites (SiCf/SiC), tungsten (W) and W-alloys. Each of these materials has its advantages and drawbacks and will be best used under certain conditions. Presently the best studied group of materials are the RAFM steels. They require the smallest extrapolation for use in DEMO but also offer the lowest upper temperature limit of operation (550 oC) and thus the lowest thermal efficiency. The other materials foreseen for more advanced breeder blanket and divertor concepts require intense fundamental R(and)D and testing before their acceptance, whereas the so-called Test Blanket Modules (TBMs) will be constructed using RAFM steel and tested in ITER. Validation of the DEMO structural materials will be done in IFMIF, the International Fusion Materials Irradiation Facility, which will produce neutron damage and transmutation products very similar to those characterising a fusion device and will allow accelerated testing with damage rates

  12. 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)

  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. 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

  15. 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

  16. Influence of LOTUS concrete structure, boron-loaded sheets, and B4C filter on the integral tritium production of a nature lithium graphite-reflected blanket and comparison with experiment

    International Nuclear Information System (INIS)

    Integral tritium production rate (TPR) measurements are important in comparisons of calculations to ascertain the suitability of computer codes and cross-section sets used in calculation. At the LOTUS facility, one of the objectives is to make measurements with different types of pure fusion and hybrid blankets and compare the results with calculations. Since the concrete cavity housing the blankets is small, it is of direct relevance to determine the influence of room-reflected neutrons on the integral TPR and, if possible, to reduce this effect by special absorbers. The effects on the TPR of a stainless steel-natural lithium-graphite-reflected blanket due to the concrete structure, B4C filter, and boron-loaded sheets covering the assembly are studied. Calculations are performed by the MCNP Monte Carlo code. Since the room-returned component depends strongly on the composition of the concrete and, more-over, does not correspond to a real blanket situation, it is advisable to compare measurements with calculations for the region where such interference is minimal. A central region is identified for the purpose of comparison. In addition to calculations for a fully homogenized blanket, the important central blanket region is considered in the form of rods, and the remaining blanket as a homogeneous region, to assess the effect of neutron streaming on the TPR of the assembly. An experiment is done by irradiating several Li2CO3 probes positioned in each tube so that the central region of interest is fully covered. The activity of the probes is measured by the standard liquid scintillation method, and the TPR for the entire region can be derived from the experimental reaction rate data. The complete details of the calculational model and the experimental procedure are provided. Good agreement is found between the calculated and experimental TPRs after accounting for various sources of errors. 14 refs., 6 figs., 4 tabs

  17. Infra-red material structures

    International Nuclear Information System (INIS)

    A process for modifying properties of infra-red materials in or for use in infra-red windows and other structures, in which the infra-red material is subjected to ion implantation whereby ions of an elemental dopant are implanted into the infra-red material under vacuum conditions. Heat diffusion may improve the depth of ion penetration. Specified materials are germanium, zinc sulphide, cadmium mercury telluride, calcium lanthanum sulphide and zinc selenide; specified dopants are boron, carbon and silicon. (author)

  18. Electrically insulating coatings for V-Li self-cooled blanket in a fusion system

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Reed, C. B.; Uz, M.; Park, J. H.; Smith, D. L.

    2000-05-17

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The liquid-metal blanket concept requires an electrically insulating coating on the first-wall structural material to minimize the magnetohydrodynamic pressure drop that occurs during the flow of liquid metal in a magnetic field. Based on the thermodynamics of interactions between the coating and the liquid lithium on one side and the structural V-base alloy on the other side, several coating candidates are being examined to perform the insulating function over a wide range of temperatures and lithium chemistries.

  19. Experimental lithium-lead module for neutronics studies of fusion blankets

    International Nuclear Information System (INIS)

    In a (D,T) type fusion machine, about 80% of the fusion energy is transported by neutrons outside the reactor's core and deposited in the blanket, an assembly of materials surrounding the machine. Tritium breeders, such as lithium and lithium-lead (LiPb) eutectic alloys, mainly dictate the design of fusion blankets. Neutronics studies, on blanket module assemblies, form an initial step towards real construction of one or another blanket. Within the framework of this dissertation, different blanket elements: first wall/structural material, tritium breeder etc., and leading fusion blanket concepts are briefly reviewed. Lithium-lead eutectic is of particular interest since the neutron multiplication takes place in the breeder, where tritium is produced. Therefore, one-dimensional optimization calculations were performed to study the use of Li17Pb83, with natural lithium abundance. Generally, this breeder is used with very high Li6 enrichment. It was found that it would be difficult to design compact blankets and to achieve reasonable tritium production, with Li17Pb83 eutectic having natural lithium isotopic composition. Either the breeder should be highly enriched in Li6 or another enriched lithium zone should follow. This study, however, led to the design and construction of the Experimental Lithium-Lead Module (EL2M). The EL2M was also used for International Comparison on Measuring Techniques of Tritium Production Rate (TPR) for Fusion Neutronics Experiments, a program initiated by JAERI (Japan Atomic Energy Research Institute) in which eight international research institutions joined. (author) figs., tabs., 124 refs

  20. 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.)

  1. Hypersonic Materials and Structures

    Science.gov (United States)

    Glass, David E.

    2016-01-01

    Thermal protection systems (TPS) and hot structures are required for a range of hypersonic vehicles ranging from ballistic reentry to hypersonic cruise vehicles, both within Earth's atmosphere and non-Earth atmospheres. The focus of this presentation is on air breathing hypersonic vehicles in the Earth's atmosphere. This includes single-stage to orbit (SSTO), two-stage to orbit (TSTO) accelerators, access to space vehicles, and hypersonic cruise vehicles. This paper will start out with a brief discussion of aerodynamic heating and thermal management techniques to address the high heating, followed by an overview of TPS for rocket-launched and air-breathing vehicles. The argument is presented that as we move from rocket-based vehicles to air-breathing vehicles, we need to move away from the insulated airplane approach used on the Space Shuttle Orbiter to a wide range of TPS and hot structure approaches. The primary portion of the paper will discuss issues and design options for CMC TPS and hot structure components, including leading edges, acreage TPS, and control surfaces. The current state-of-the-art will be briefly discussed for some of the components.

  2. 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

  3. 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

  4. Nuclear analysis of DEMO water-cooled blanket based on sub-critical water condition

    International Nuclear Information System (INIS)

    Highlights: ► For sub-critical water condition, the size of cooling loop would be more longer, for example, 2 m. ► Local TBR is related to the material fraction of breeders and multipliers, the beryllium is the dominant. ► Front area of blanket is dominant for blanket design and it would contribute the most of TBR comparing to the backside zones. - Abstract: For the water-cooled solid blanket of DEMO, the nuclear analysis was performed based on present cooling piping system. Especially, distributions of neutron load and temperature were calculated with Pn is 5 MW/m2. Furthermore, the local TBR was optimized by changing the material proportion for each Pn level (1–5 MW/m2). It was confirmed that the size of cooling loop for sub-critical water could be used as about 2000 × 450 mm and the cooling pipe diameter of D is 12 mm, d is 9 mm at v is 5.36 m/s. The pipe pitches would vary with Pn level which is related to the blanket structure design. Nuclear heat distribution is the base to decide the distribution of cooling pipe positions. It was found that the local TBR of blanket would be dropped down along with the Pn level rising which was mainly depended on the thickness of beryllium variation. Finally, the layout of cooling pipes for each level was obtained.

  5. Activation Analysis for a He/LiPb dual Coolant Blanket for DEMO Reactor

    OpenAIRE

    Catalán, J.P.; Ogando Serrano, Francisco; Sanz Gonzalo, Javier

    2010-01-01

    The objective of the Spanish national project TECNO_FUS is to generate a conceptual design of a DCLL (Dual-Coolant Lithium-Lead) blanket for the DEMO fusion reactor. The dually-cooled breeding zone is composed of He/Pb-15.7 6Li and SiC as liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer-97) for the blanket and austenitic steel (316LN) for the Vacuum Vessel (VV). The goal of this work is to analyze the radioactive waste production by the neutron-i...

  6. 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)

  7. Experimental investigations on the substantiation of the conception of tokamak blanket cooling with a lead-lithium eutectic

    International Nuclear Information System (INIS)

    One describes the experiments to look into possible application of Pb-Li eutectic to cool tokamak blanket. One investigated into effect of Pb-Li eutectic on characteristics of insulating coatings (IC) of structural materials. One investigated, as well, into MHD resistance of Pb-Li eutectic and into interaction within structural material - IC - Li(17)Pb(83) eutectic - H2O, O2, H2, Bi impurities system. The derived results are used to justify application of the mentioned coolant to cool tokamak blanket

  8. Structural materials for fusion reactors

    International Nuclear Information System (INIS)

    In order to preserve the condition of an environmentally safe machine, present selection of materials for structural components of a fusion reactor is made not only on the basis of adequate mechanical properties, behavior under irradiation and compatibility with other materials and cooling media, but also on their radiological properties, i.e. activity, decay heat, radiotoxicity. These conditions strongly limit the number of materials available to a few families of alloys, generically known as low activation materials. We discuss the criteria for deciding on such materials, the alloys resulting from the application of the concept and the main issues and problems of their use in a fusion environment. (author)

  9. Properties and Technology for Quasi-Composite Blanket Using Natural Reinforcement of the Metal by Strain Affected Areas

    Directory of Open Access Journals (Sweden)

    A. Kirichek

    2013-12-01

    Full Text Available Techniques for making materials with advanced performance attributes at the expense of blanket heterogeneous strengthening are considered. A new trend is defined in a multiple increase of performance attributes in metal materials by natural reinforcement with nanostructural and ultra-fine-grained fragments. The application of a wave strain hardening technique is substantiated for obtaining a heterogeneous structure in wide-area listed full-size products including bulky ones. A high carrying capacity of heavy-loaded material with a deep-strengthened blanket is determined.

  10. Behaviour of Structural Materials: China

    International Nuclear Information System (INIS)

    Since accelerator driven system (ADS) will be used as the facility for long life radioactive waste transmutation, some issues in structure materials of ADS are being paid attention, such as the compatibility of the materials and coolant, the development and selection of materials which are irradiation resistance to neutron and high energy proton, and corrosion resistance to Pb-Bi. In China, ADS material programme was proposed in 2000, started in 2001, and the compatibility project is focused on three topics: – The compatibility studies of tungsten with sodium and with water; – The development and investigation of the structure materials; – The primary tests of the complex technologies for W–S.S. and W–Zr cladding. Some primary results from these researches have been obtained, and some investigations are being performed. Since our investigation is in an early step, a lot of issues on materials have to be studied further in the near future

  11. Study of MHD Corrosion and Transport of Corrosion Products of Ferritic/Martensitic Steels in the Flowing PbLi and its Application to Fusion Blanket

    OpenAIRE

    Saeidi, Sheida

    2014-01-01

    Two important components of a liquid breeder blanket of a fusion power reactor are the liquid breeder/coolant and the steel structure that the liquid is enclosed in. One candidate combination for such components is Lead-Lithium (PbLi) eutectic alloy and advanced Reduced Activation Ferritic/Martensitic (RAFM) steel. Implementation of RAFM steel and PbLi in blanket applications still requires material compatibility studies as many questions related to physical/chemical interactions in the RAFM...

  12. Development of anti-corrosion coating on low activation materials against fluoridation and oxidation in Flibe blanket environment

    International Nuclear Information System (INIS)

    W coating by vacuum plasma spray process and Cr coating by chromizing process were performed on fusion low activation materials, JLF-1 ferritic steel and NIFS-HEAT-2 vanadium alloy. The present study discusses feasibility of the coatings as anti-corrosion coating against fluoridation in Flibe for fusion low activation materials. Coatings were characterized by microstructural analysis and examination on chemical stability by corrosion tests. The corrosion tests were conducted with H2O-47% HF solution at RT and He-1% HF-0.06 H2O gas mixture at 823 K to simulate fluoridation and oxidation in Flibe. The coatings presented suppression of fluoride formation compared with JLF-1 or NIFS-HEAT-2, however weight loss due to WF6 formation was induced, and much Cr2O3 was formed.

  13. Porous Materials - Structure and Properties

    DEFF Research Database (Denmark)

    Nielsen, Anders

    The paper presents some viewpoints on the description of the pore structure and the modelling of the properties of the porous building materials. Two examples are given , where it has been possible to connect the pore structure to the properties: Shrinkage of autoclaved aerated concrete and the...... properties of lime mortar....

  14. Porous Materials - Structure and Properties

    DEFF Research Database (Denmark)

    Nielsen, Anders

    1997-01-01

    The paper presents some viewpoints on the description of the pore structure and the modelling of the properties of the porous building materials. Two examples are given , where it has been possible to connect the pore structure to the properties: Shrinkage of autoclaved aerated concrete and the...

  15. Structural Chemistry of Functional Materials

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    @@ This innovative research group on structural chemistry of functional materials was approved by NSFC in 2005.Headed by Prof.HONG Maochun, the team consists of several young research scientists from the CAS Fujian Institute of Research on the Structures of Matter, including Profs CAO Rong, LU Canzhong, GUO Guocong, CHEN Zhongning, MAO Jianggao Mao and CHEN Ling.

  16. Development of fabrication technology for low activation vanadium alloys as fusion blanket structural materials

    International Nuclear Information System (INIS)

    High purity vanadium alloy products, such as plates, wires and tubes, were fabricated from reference high-purity V-4Cr-4Ti ingots designated as NIFS-HEAT, by using technologies applicable to industrial scale fabrication. Impurity behavior during breakdown, and its effect on mechanical properties were investigated. It was revealed that mechanical properties of the products were significantly improved by the control of Ti-C, N, O precipitation induced during the processes. (author)

  17. Structural materials for fusion magnets

    International Nuclear Information System (INIS)

    Of major technical and cost impact to Magnetic Fusion Energy development are the materials for the magnet structure. Likened to gas pressure, the magnetic field lines try to expand the structure with equivalent pressures up to 1000 atm. Not only are large tensile forces produced, but significant bending forces may also be present. To withstand these forces in the restricted spaces available, materials of exceptional strength and toughness are required. In this regard, the low-temperature environment of superconducting magnets can be an advantage because many materials exhibit enhanced properties at reduced temperatures. Those materials and fabrication techniques that are attractive to fusion magnets are discussed and relative comparisons made. Considerations such as strength, toughness, and joining techniques are balanced agains recommended design criteria to reach an optimum design. Several examples of material selection are cited for large fusion magnets such as Baseball II, the Mirror Fusion Test Facility, the Toroidal Fusion Test Facility, and the Large Coil Project. (orig.)

  18. Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

    OpenAIRE

    Stork, D.; Agostini, P.; Boutard, J.L.; Buckthorpe, D.; Diegele, E.; Dudarev, S. L.; English, C.; Federici, G.; Gilbert, M. R.; Gonzalez, S.; Ibarra, A.; Linsmeier, Ch.; LI PUMA, A; Marbach, G.; Morris, P. F.

    2014-01-01

    The findings of the EU 'Materials Assessment Group' (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision. A DEMO phase I with a 'Starter Blanket' and 'Starter Divertor' is foreseen: the blanket being capable of withstanding >= 2 MW yr m(-2) fusion neutron...

  19. Thermal-hydraulic performance and structural thermal stress analysis for ITER shield blanket module nearby NB rejoin

    International Nuclear Information System (INIS)

    Hydraulic and thermal analysis of the International Thermonuclear Experimental Reactor (ITER) standard neutral beam (NB) blanket module was carried out in order to check whether the latest design meets ITER requirements. Minor-loss coefficients were estimated with a CFD code, and friction factors of straight channels were obtained using existing formulas. The effects of different radial hole's diameter, length of the back of the radial hole, size of clearance, type of flow driver, branch velocity and flow direction on minor-loss coefficients for radial holes were investigated. Since total mass flow rate and dimensions of the cooling channels were given, when pressure drop due to intersection of the radial hole with back drilled collector was ignored, we can obtain pressure drop, flow rate, velocity and heat transfer coefficient in each radial hole. An improved calculation without neglecting the pressure drop caused by the intersection was also done to compare with the simplified one. Finally, maximum temperature, thermal stress and deformation were evaluated according to FEM thermal analysis. The results of the latest hydraulic and thermal analysis indicate that the current design meets ITER requirements well, except that flow distribution is not so uniform when different types of flow drivers are used, and temperature in the front head surface is a little high. Improved design is necessary in the further. (authors)

  20. 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

  1. The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

    2015-11-15

    In the present investigation, a hybrid reactor system was designed by using 100 % Flibe, 90 % Flibe-10 % ThF{sub 4}, 90 % Flibe-10 % UF{sub 4} fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

  2. The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system

    International Nuclear Information System (INIS)

    In the present investigation, a hybrid reactor system was designed by using 100 % Flibe, 90 % Flibe-10 % ThF4, 90 % Flibe-10 % UF4 fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

  3. 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.

  4. 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

  5. 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

  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. 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.

  8. Structural materials for fusion magnets

    International Nuclear Information System (INIS)

    Of major technical and cost impact to Magnetic Fusion Energy development are the materials for the magnet structure. Those materials and fabrication techniques that are attractive to fusion magnets are discussed and relative comparisons made. Considerations such as strength, toughness, and joining techniques are balanced against recommended design criteria to reach an optimum design. Several examples of material selection are cited for large fusion magnets such as Base II, the Mirror Fusion Test Facility, the Toroidal Fusion Test Facility, and the Large Coil Project

  9. Blanket and vacuum vessel design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li2O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

  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. 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

  12. 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.

  13. Behaviour of Structural Materials: Belgium

    International Nuclear Information System (INIS)

    For the development and qualification of structural materials for advanced and sustainable nuclear systems, the efforts are concentrated on the systems which are considered as priority for the Institute for Advanced Nuclear Systems. These are: accelerator driven systems where both the spallation source and the cooling medium is the liquid lead-bismuth eutecticum, lead cooled fast reactor (LFR) as a serious option for GEN IV systems. The material development and qualification activities are primarily concerned with the chemical and physical integrity of the selected materials when subjected to the expected but new harsh conditions. The issues that are being addressed and will be pursued in the future are: 1) the effects of irradiation on the material performance including spectrum, temperature and flux variations, 2) the specific compatibility issues related to application of these materials in the foreseen reactor environment especially under the synergetic effect of irradiation and 3) the modeling of these effects in order to be able to extrapolate the validity of the data to cover relevant operation conditions as no irradiation facility exist now a days that is able to reproduce the expected environment. The definition of these priority reactor systems focusses the research efforts on the high chromium ferritic–martensitic steels, austenitic stainless steels and oxide dispersion strengthened steels and alloys with emphasis on their fabricability, weldability and performance. This goal relies on the expertise, available from the activities along the research into radiation and ageing effects on structural materials for the currently operating nuclear power plants. This expertise covers databases on the behaviour of irradiated materials, validated test methods for evaluation of mechanical properties and environmental assisted cracking and models for prediction of radiation damage effects on the microstructure of the material and its effect on the material

  14. A strategic analysis of the development of structural materials for proto-type reactors for fusion

    International Nuclear Information System (INIS)

    Structural Materials Research and Development Subcommittee of Nuclear Materials Committee in Japan Atomic Energy Research Institute had made a study to propose a strategy how to expedite the research and development of structural materials for fusion reactors. This study was carried out along with the interim report of the Development of Structural Materials in Fusion Reactors proposed by Planning and Promotion Subcommittee of Fusion Council as well as with the Third Phase Basic Program of Fusion Research and Development settled by the Atomic Energy Commission. The present report was published to publicize the results of analyses of this study. In this report we focused mostly on the development of structural materials of blankets for tritium breeding because it is considered to be the most difficult task in the materials development due to severe conditions imposing on the blankets. We selected three candidate materials, namely, reduced low activation ferritic/ martensitic steel, SiC/SiC composites and Vanadium alloys, and elucidate the conditions in which these materials would be used as well as the design requirements for each material. Based on these conditions and requirements, we described the present status and the key issues of each material. For the development of the structural materials for the blankets, the keenest issue is the improvement and evaluation of radiation integrity and stability. Therefore, the necessity of radiation facilities, especially accelerator-type neutron sources with near fusion energy spectra was described. In addition the usage of fission reactors as irradiation facilities was also emphasized. In the processing of this reviewing we categorized reduced low activation ferritic/martensitic steel as advanced material, and SiC/SiC composites and Vanadium alloys as next-generation advanced material from the present status of developmental maturity. A periodical check and review in order to take the future progress in the development of

  15. 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 ...

  16. 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.)

  17. 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)

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

    International Nuclear Information System (INIS)

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li2TiO3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li2TiO3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328degC in

  19. 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

  20. Radiation damage of structural materials

    CERN Document Server

    Koutsky, Jaroslav

    1994-01-01

    Maintaining the integrity of nuclear power plants is critical in the prevention or control of severe accidents. This monograph deals with both basic groups of structural materials used in the design of light-water nuclear reactors, making the primary safety barriers of NPPs. Emphasis is placed on materials used in VVER-type nuclear reactors: Cr-Mo-V and Cr-Ni-Mo-V steel for RPV and Zr-Nb alloys for fuel element cladding. The book is divided into 7 main chapters, with the exception of the opening one and the chapter providing a phenomenological background for the subject of radiation damage. Ch

  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. 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. 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.

  4. 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

  5. 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

  6. Study on compact design of remote handling equipment for ITER blanket maintenance

    International Nuclear Information System (INIS)

    In the ITER, the neutrons created by D-T reactions activate structural materials, and thereby, the circumstance in the vacuum vessel is under intense gamma radiation field. Thus, the in-vessel components such as blanket are handled and replaced by remote handling equipment. The objective of this report is to study the compactness of the remote handling equipment (a vehicle/manipulator) for the ITER blanket maintenance. In order to avoid the interferences between the blanket and the equipment during blanket replacement in the restricted vacuum vessel, a compact design of the equipment is required. Therefore, the compact design is performed, including kinematic analyses aiming at the reduction of the sizes of the vehicle equipped with a manipulator handling the blanket and the rail for the vehicle traveling in the vacuum vessel. Major results are as follows: 1. The compact vehicle/manipulator is designed concentration on the reduction of the rail size and simplification of the guide roller mechanism as well as the reduction of the gear diameter for vehicle rotation around the rail. Height of the rail is reduced from 500 mm to 400 mm by a parameter survey for weight, stiffness and stress of the rail. The roller mechanism is divided into two simple functional mechanisms composed of rollers and a pad, that is, the rollers support relatively light loads during rail deployment and vehicle traveling while a pad supports heavy loads during blanket replacement. Regarding the rotation mechanism, the double helical gear is adopted, because it has higher contact ratio than the normal spur gear and consequently can transfer higher force. The smaller double helical gear, 996 mm in diameter, can achieve 26% higher output torque, 123.5 kN·m, than that of the original spur gear of 1,460 mm in diameter, 98 kN·m. As a result, the manipulator becomes about 30% lighter, 8 tons, than the original weight, 11.2 tons. 2. Based on the compact design of the vehicle/manipulator, the

  7. 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)

  8. Structural materials for fusion reactors

    International Nuclear Information System (INIS)

    Full text: A long term solution to problems of energy production, green house gas generation, and pollution control may rest with controlled nuclear fusion reactors. Candidate structural materials for such reactors include low activation ferritic steels. Understanding and eliminating deleterious irradiation effects in these materials is the goal of these experiments in this collaboration using the ANL facility. In recent experiments on ferritic alloys we have recently found a significant difference with alloy composition in the microstructural response to irradiation, which corresponds to a bulk mechanical property change at a similar composition. In a collaboration between the Department of Materials at the University of Oxford and the Materials Science Division at Argonne National Laboratory, experiments which employ the unique transmission electron microscope and in situ ion irradiation user facility at ANL were performed on a series of Fe-Cr alloys. Enhanced nanometer-sized defect formation with Cr concentrations up to 11 % have been found and correlated with a decrease in mechanical hardening and embrittlement in similar alloys. (author)

  9. Behaviour of Structural Materials: India

    International Nuclear Information System (INIS)

    Two buoyancy driven LBE loops have been set up in BARC to study corrosion effects of LBE on candidate structural materials for spallation target module; like SS-316 L, 316 LN, SS-304 L, 9Cr-1Mo etc. The larger loops getting ready would be ~ 7 metre high and its riser and downcomer legs would operate at 5500°C and 4500°C respectively. The buoyancy head is estimated to provide a LBE mass flow rate of 1.7kg/s. The flow velocity around the sample is ~0.6 m/s. Material samples exposed in an already operational smaller corrosion loop were being evaluated by Charpy and Tensile tests after exposing the samples for 2000 hours in the flow with oxygen control. Provision is also being made in this loop to enhance the LBE flow by gas injection

  10. 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

  11. Structural material requirements and related R and D for ITER plasma-facing components

    International Nuclear Information System (INIS)

    Solution-annealed 316 type austenitic steels have been selected as structural material for the frirst wall and blanket of ITER, with cold-worked 316 and MnCr austenitic steels as candidate alloys. The effects of irradiation on their properties: tensile, fracture toughness, low cycle fatigue, creep and rupture, swelling, stress corrosion cracking and hydrogen embrittlement, are discussed in terms of the proposed ITER operating conditions. Arguments for the selection of oxide dispersion strengthened Cu alloy for divertor plates are presented. Finally, the investigations which are being undertaken within the ITER Technology Programme to provide additional data are reviewed. (orig.)

  12. 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...

  13. 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.)

  14. 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.)

  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. Integral neutronics experiments in analytical mockups for blanket of a hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Rong, E-mail: liurongzy@163.com; Zhu, Tonghua; Lu, Xinxin; Wang, Xinhua; Yan, Xiaosong; Feng, Song; Yang, Yiwei; Wang, Mei; Jiang, Li

    2014-12-15

    Highlights: • For checking property of the hybrid blanket by integral experiments, three mockups are established. • In spherical mockup with depleted uranium and cubic mockup with natural uranium, the plutonium production rates and uranium fission rates are measured. • In spherical mockup with depleted uranium and LiPb, tritium production rates are measured. • The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data. - Abstract: The paper describes recent progress in integral neutronics experiments in the analytical mockups for the blanket in a fusion-fission hybrid energy reactor. A conceptual blanket of the hybrid reactor is mainly loaded with natural uranium and lithium material. In the fission fuel region, uranium material and light water are arranged alternately. The mockups of the conceptual blanket are designed and used for checking neutron property of the blanket by integral experiments. Based on materials available, the spherical fission mockup for fission research and plutonium production consists of three layers of depleted uranium shells and several layers of polyethylene and graphite shells. The spherical lithium mockup for tritium production consists of depleted uranium and LiPb alloy shells. The cubic mockup consists of natural uranium and polyethylene and its structure is basically consistent with one of the fuel region. In the mockups with the D-T neutron source, the plutonium production rates, uranium fission rates and tritium production rates are measured, separately. The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data.

  17. 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

  18. 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

  19. Application of martensitic stainless steels in long lifetime fusion first wall/blankets

    International Nuclear Information System (INIS)

    An assessment is made of the feasibility and design impact of using a ferromagnetic martensitic steel such as HT-9 in first wall/blanket structures. A preliminary analysis indicates that HT-9 may offer significantly greater wall lifetimes than 20CW316 for temperatures below about 5200C. A number of exploratory measurements and rough calculations suggest that the presence of ferromagnetic material does not significantly affect the operational characteristics of tokamak power reactors

  20. The application of martensitic stainless steels in long lifetime fusion first wall/blankets

    International Nuclear Information System (INIS)

    An assessment is made of the feasibility and design impact of using a ferromagnetic martensitic steel such as HT-9 in first wall/blanket structures. A preliminary analysis indicates that HT-9 may offer significantly greater wall lifetimes than 20% cold worked type 316 stainless steel for temperatures below about 5200C. A number of exploratory measurements and rough calculations suggest that the presence of ferromagnetic material does not significantly affect the operational characteristics of tokamak power reactors. (orig.)

  1. The rheology of structured materials

    Science.gov (United States)

    Sun, Ning

    2000-10-01

    In this work, the rheological properties of structured materials are studied via both theoretical (continuum mechanics and molecular theory) and experimental approaches. Through continuum mechanics, a structural model, involving shear-induced structural breakdown and buildup, is extended to model biofluids. In particular, we study the cases of steady shear flow, hysteresis, yield stress, small amplitude oscillatory flow as well as non-linear viscoelasticity. Model predictions are successfully compared with experimental data on complex materials such as blood and a penicillin suspension. Next, modifications are introduced into the network model. A new formulation involving non-affine motion is proposed and its applications are presented. The major improvement is that a finite elongational viscosity is predicted for finite elongational rate, contrary to infinite elongational viscosities existing at some elongational rates predicted by most previous network models. Comparisons with experimental data on shear viscosity, primary normal stress coefficient and elongational viscosity are given, in terms of the same set of model parameters. Model predictions for the stress growth are also shown. The model is successfully tested with data on a polyisobutylene solution (S1), on a polystyrene solution and on a poly-alpha-methylstyrene solution. A further extension of the network model is related to the prediction of the stress jump phenomenon which is defined as the instantaneous gain or loss of stress on startup or cessation of a deformation. It is not predicted by most existing models. In this work, the internal viscosity idea used in the dumbbell model is incorporated into the transient network model. Via appropriate approximations, a closed form constitutive equation, which predicts a stress jump, is obtained. Successful comparisons with the available stress jump measurements are given. In addition, the model yields good quantitative predictions of the standard steady

  2. Thermal cycle test of elemental mockups of ITER breeding blanket

    International Nuclear Information System (INIS)

    Thermal cycle tests for mockups of breeder pebble beds of ITER breeding blanket have been carried out to investigate their thermo-mechanical behavior with the interaction between a pebble bed and a breeder rod containing the breeder pebbles. The mockups have been designed to demonstrate a part of the Breeder Inside Tube (BIT)' structure of ITER breeding blanket. Candidate material pebbles of Li2TiO3 was applied as breeder specimen, and Al pebbles were applied for simulating the neutron multiplier of Be pebbles. These pebbles have been packed in test tubes by using a vibration machine. Tested configurations were single layer mockups with Li2TiO3 single diameter packing and binary packing beds, and double layer mockups with Li2TiO3/Al single diameter packing and binary packing beds. In order to clarify the deformation performance of breeder tube, two different thickness of the breeder rod were also tested: one for nominal condition and another for acceleration test. Pebble bed of Li2TiO3 is heated with an electric heater, which is equipped at the center of the breeder rod, simulating the temperature profile by volumetric heating of breeder pebbles. The outside of a breeder rod in a single layer mockup and the outside of the outer tube in case of double layer mockup is cooled by water. Temperature of the breeder beds has been controlled by a power input of the heater. After the thermal cycle tests, the internal dimensions and local packing fraction of mockups have been examined by using an X-ray CT device. As the result, no significant change of packing fraction was observed after five thermal cycles with maximum heater temperature of 600degC. Any bulging of the breeder rod or any cracking of the pebble has not been observed. A soundness of the typical structure and breeder pebble bed of ITER breeding blanket against thermal cycles was confirmed. (author)

  3. 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

  4. Conceptual design of solid breeder blanket system cooled by supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment] [and others

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

  5. Conceptual design of solid breeder blanket system cooled by supercritical water

    International Nuclear Information System (INIS)

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li2TiO3 or Li2O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for the energy

  6. 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.

  7. Low-Activation structural ceramic composites for fusion power reactors: materials development and main design issues

    International Nuclear Information System (INIS)

    This paper is devoted to the development of advanced Low-Activation Materials (LAMs) with favourable short-term activation characteristics for the use as structural materials in a fusion power reactor (in order to reduce the risk associated with a major accident, in particular those related with radio-isotopes release in the environment), and to try to approach the concept of an inherently safe reactor. LA Ceramics Composites (LACCs) are the most promising LAMs because of their relatively good thermo-mechanical properties. At present, SiC/SiC composite is the only LACC considered by the fusion community, and therefore is the one having the most complete data base. The preliminary design of a breeding blanket using SiC/SiC as structural material indicated that significant improvement of its thermal conductivity is required. (orig.)

  8. Low-activation structural ceramic composites for fusion power reactors: materials development and main design issues

    International Nuclear Information System (INIS)

    Development of advanced Low-Activation Materials (LAMs) with favourable short-term activation characteristics is discussed, for the use as structural materials in a fusion power reactor (in order to reduce the risk associated with a major accident, in particular those related with radio-isotopes release in the environment), and to try to approach the concept of an inherently safe reactor. LA Ceramics Composites (LACCs) are the most promising LAMs because of their relatively good thermo-mechanical properties. At present, SiC/SiC composite is the only LACC considered by the fusion community, and therefore is the one having the most complete data base. The preliminary design of a breeding blanket using SiC/SiC as structural material indicated that significant improvement of its thermal conductivity is required. (author) 11 refs.; 3 figs

  9. Cryogenic structural materials for superconducting magnets

    International Nuclear Information System (INIS)

    This paper reviews research in the United States and Japan on structural materials for high-field superconducting magnets. Superconducting magnets are used for magnetic fusion energy devices and for accelerators that are used in particle-physics research. The cryogenic structural materials that we review are used for magnet cases and support structures. We expect increased materials requirements in the future

  10. 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 ...

  11. 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

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

    International Nuclear Information System (INIS)

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

  13. IFMIF suitability for evaluation of fusion functional materials

    OpenAIRE

    Casal, N.; Sordo Balbín, Fernando; Mota, F.; Jordanova, J.; Garcia, A.; Ibarra, A.; Vila, R.; Rapisarda, D; Queral, V.; Perlado Martin, Jose Manuel

    2011-01-01

    The International FusionMaterials Irradiation Facility (IFMIF) is a future neutron source based on the D-Li stripping reaction, planned to test candidate fusionmaterials at relevant fusion irradiation conditions. During the design of IFMIF special attention was paid to the structural materials for the blanket and first wall, because they will be exposed to the most severe irradiation conditions in a fusion reactor. Also the irradiation of candidate materials for solid breeder blankets is plan...

  14. Radiation effects on structural materials

    International Nuclear Information System (INIS)

    This report discusses the following topics on the effect radiation has on thermonuclear reactor materials: Atomic Displacements; Microstructure Evolution; Materials Engineering, Mechanics, and Design; Research on Low-Activation Steels; and Research Motivated by Grant Support

  15. Materials research at Stanford University. [composite materials, crystal structure, acoustics

    Science.gov (United States)

    1975-01-01

    Research activity related to the science of materials is described. The following areas are included: elastic and thermal properties of composite materials, acoustic waves and devices, amorphous materials, crystal structure, synthesis of metal-metal bonds, interactions of solids with solutions, electrochemistry, fatigue damage, superconductivity and molecular physics and phase transition kinetics.

  16. Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

    International Nuclear Information System (INIS)

    Highlights: ► Dual-Coolant He/Pb15.7Li breeding blanket for a DEMO fusion reactor is studied. ► An iterative process optimizes neutronic responses minimizing reactor dimension. ► A 3D toroidally symmetric geometry has been generated from the CAD model. ► Overall TBR values support the feasibility of the conceptual model considered. ► Power density in TF coils is below load limit for quenching. - Abstract: The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

  17. Building Investigation: Material or Structural Performance

    OpenAIRE

    Yusof M.Z.

    2014-01-01

    Structures such as roof trusses will not suddenly collapse without ample warning such as significant deflection, tilting etc. if the designer manages to avoid the cause of structural failure at the material level and the structural level. This paper outlines some principles and procedures of PDCA circle and QC tools which can show some clues of structural problems in terms of material or structural performance

  18. Building Investigation: Material or Structural Performance

    Directory of Open Access Journals (Sweden)

    Yusof M.Z.

    2014-03-01

    Full Text Available Structures such as roof trusses will not suddenly collapse without ample warning such as significant deflection, tilting etc. if the designer manages to avoid the cause of structural failure at the material level and the structural level. This paper outlines some principles and procedures of PDCA circle and QC tools which can show some clues of structural problems in terms of material or structural performance

  19. The Karlsruhe solid breeder blanket and the test module to be irradiated in ITER/NET

    International Nuclear Information System (INIS)

    The blanket for the DEMO reactor should operate at an average neutron flux of 2.2 MW/m2 for 20000 h. This requires the use of a structural material which can withstand high neutron fluences without swelling. The ferritic steel Manet was chosen for this purpose. The breeder material is in the form of Li4SiO4 pebbles of 0.35 to 0.6 mm diameter. The 6 mm thick beds of pebbles are placed between beryllium plates which are cooled by high pressure helium flowing inside steel tubes. Breeder material and beryllium are contained in radial canisters, placed inside boxes. The coolant helium enters the blanket at 250deg C, cools first the box walls and then the breeder and multiplier, and leaves the blanket at 450deg C. The maximum temperature in the first wall steel is 550deg C, while the minimum and maximum temperatures in the breeder are 380 and 820deg C, respectively. The resulting total tritium inventory in the breeder is only 10 g, and the real tridimensional tritium breeding ratio is 1.11. The conceptual design of the test module, of its extraction system and of the required out-of-reactor ancillary systems has allowed an estimate of the time constants of the various components and thus allowed an assessment of the requirements given by the testing of the modules on the NET/ITER machine. (orig.)

  20. 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.)

  1. 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

  2. An analysis of electron beam welds in a dual coolant liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Numerical simulation of electron beam welding of blanket segments was performed using non-linear finite element code ABAQUS. The thermal and stress fields were assumed uncoupled, while preserving the temperature dependency of all material parameters. The martensite-austenite and austenite-martensite transformations were taken into account through volume shrinking/expansion effects, which is consistent with available data. The distributions of post welding residual stress in a complex geometry of the first wall are obtained. Also, the effects of preheating and post-welding heat treatment were addressed. Time dependent temperature and stress-strain fields obtained provide good insight into the welding process. They may be used directly to support reliability and life-time studies of blanket structures. On the other hand, they provide useful hints about the feasibility of the geometrical configurations as proposed by different design concepts. (orig.)

  3. 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

  4. 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

  5. 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)

  6. 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.

  7. Materials analogue of zero-stiffness structures

    Science.gov (United States)

    Kumar, Arun; Subramaniam, Anandh

    2011-04-01

    Anglepoise lamps and certain tensegrities are examples of zero-stiffness structures. These structures are in a state of neutral equilibrium with respect to changes in configuration of the system. Using Eshelby's example of an edge dislocation in a thin plate that can bend, we report the discovery of a non-trivial new class of material structures as an analogue to zero-stiffness structures. For extended positions of the edge dislocation in these structures, the dislocation experiences a zero image force. Salient features of these material structures along with the key differences from conventional zero-stiffness structures are pointed out.

  8. 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

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  9. 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.)

  10. Assessment of martensitic steels as structural materials in magnetic fusion devices

    International Nuclear Information System (INIS)

    This manuscript documents the results of preliminary experiments and analyses to assess the feasibility of incorporating ferromagnetic martensitic steels in fusion reactor designs and to evaluate the possible advantages of this class of material with respect to first wall/blanket lifetime. The general class of alloys under consideration are ferritic steels containing from about 9 to 13 percent Cr with some small additions of various strengthening elements such as Mo. These steels are conventionally used in the normalized and tempered condition for high temperature applications and can compete favorably with austenitic alloys up to about 6000C. Although the heat treatment can result in either a tempered martensite or bainite structure, depending on the alloy and thermal treatment parameters, this general class of materials will be referred to as martensitic stainless steels for simplicity

  11. Utilization of fusion neutrons very high energy effects in the design of a fusion reactor tritium breeding blanket

    International Nuclear Information System (INIS)

    Tritium needed for ITER fusion reactions can be regenerated in a blanket located in the tokamak. Such breeding has to be achieved through special interactions between very high energy (14.1 MeV) fusion neutrons and the blanket materials, like the lithium-7 tritium breeding effect and the thorium-232 neutron multiplication effect, all this while occupying the smallest possible space. Five blankets were designed and investigated; three of them were purely composed of lithium materials, while the two others were designed by adding a thorium layer before the lithium layer. A 3-D modeling was created using a Monte Carlo N-particle Code (MCNP) to simulate the fusion neutrons histories through the tritium breeding blankets, with a blanket thickness ranging from 1 cm to 200 cm. The minimum blanket thickness necessary to obtain a tritium breeding ratio (TBR) greater than one ranges from 20 cm to 45 cm. In particular, the Lithium Oxide Mono-Layer Blanket (LO-MLB) achieves a TBR greater than one while allowing blanket thickness to stay under 25 cm, thus making it the most efficient blanket in this sample. Second, the maximum TBR for thick blankets ranges from 1.5 to 2. In particular, the Natural Lithium Mono-Layer Blanket (NL-MLB) displays the highest maximum TBR thanks to its perfect combination of the lithium-7 and lithium-6 tritium breeding capacity. (author)

  12. Euro hybrid materials and structures. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Hausmann, Joachim M.; Siebert, Marc (eds.)

    2016-08-01

    In order to use the materials as best as possible, several different materials are usually mixed in one component, especially in the field of lightweight design. If these combinations of materials are joined inherently, they are called multi material design products or hybrid structures. These place special requirements on joining technology, design methods and manufacturing and are challenging in other aspects, too. The eight chapters with manuscripts of the presentations are: Chapter 1- Interface: What happens in the interface between the two materials? Chapter 2 - Corrosion and Residual Stresses: How about galvanic corrosion and thermal residual stresses in the contact zone of different materials? Chapter 3 - Characterization: How to characterize and test hybrid materials? Chapter 4 - Design: What is a suitable design and dimensioning method for hybrid structures? Chapter 5 - Machining and Processing: How to machine and process hybrid structures and materials? Chapter 6 - Component Manufacturing: What is a suitable manufacturing route for hybrid structures? Chapter 7 - Non-Destructive Testing and Quality Assurance: How to assure the quality of material and structures? Chapter 8 - Joining: How to join components of different materials?.

  13. Corrosion and compatibility of structural materials in liquid lead alloys

    International Nuclear Information System (INIS)

    The paper describes investigations of the behavior of structural materials in both eutectic lead alloys, Pb-17Li as the liquid breeder within the European WCLL-concept for Fusion Reactors and Pb-55Bi as a possible spallation target and coolant of future Accelerator Driven Systems (ADS) for the transmutation of minor actinides and long-lived fission products from nuclear waste. Ferritic-martensitic steels of the 8-10 wt.-% Cr type are considered as structural materials for the use in liquid Pb-17Li at temperatures up to 500degC. Steels like MANET I, Optifer, F82H-mod. and EUROFER are investigated in the flowing liquid metal over a long period of time. Due to the specific physical chemistry of Pb-17Li, dissolution corrosion is the major corrosion mechanism of iron-based alloys. No oxide formation on steel surfaces can occur due to the very low oxygen potential in Pb-17Li. Therefore, only coatings could be a solution to minimize corrosion effects, at medium (500degC) and higher operational temperatures (650degC) of advanced blanket concepts. Ferritic-martensitic and austenitic steels are considered as potential structural materials for ADS application, too. But the high nickel solubility in Pb-55Bi limits the use of unprotected austenitic steels to temperatures of about 400degC. Because of the much higher oxygen potential of Pb-55Bi, a totally different strategy of minimizing corrosion must be applied. By using active oxygen control, a desired oxygen activity (concentration) will be adjusted in the liquid metal via gas phase equilibrium. If the oxygen activity of the liquid metal is set in the right 'window' of the corresponding thermodynamic properties, a controlled formation of oxide layers on the steel surfaces will be found. These in-situ formed oxides can act as corrosion barriers and make the use of high-nickel alloys possible, even at temperatures of around 600degC. (author)

  14. Magnetism and Structure in Functional Materials

    CERN Document Server

    Planes, Antoni; Saxena, Avadh

    2005-01-01

    Magnetism and Structure in Functional Materials addresses three distinct but related topics: (i) magnetoelastic materials such as magnetic martensites and magnetic shape memory alloys, (ii) the magnetocaloric effect related to magnetostructural transitions, and (iii) colossal magnetoresistance (CMR) and related magnanites. The goal is to identify common underlying principles in these classes of materials that are relevant for optimizing various functionalities. The emergence of apparently different magnetic/structural phenomena in disparate classes of materials clearly points to a need for common concepts in order to achieve a broader understanding of the interplay between magnetism and structure in this general class of new functional materials exhibiting ever more complex microstructure and function. The topic is interdisciplinary in nature and the contributors correspondingly include physicists, materials scientists and engineers. Likewise the book will appeal to scientists from all these areas.

  15. Radiation Effects on Spacecraft Structural Materials

    International Nuclear Information System (INIS)

    Research is being conducted to develop an integrated technology for the prediction of aging behavior for space structural materials during service. This research will utilize state-of-the-art radiation experimental apparatus and analysis, updated codes and databases, and integrated mechanical and radiation testing techniques to investigate the suitability of numerous current and potential spacecraft structural materials. Also included are the effects on structural materials in surface modules and planetary landing craft, with or without fission power supplies. Spacecraft structural materials would also be in hostile radiation environments on the surface of the moon and planets without appreciable atmospheres and moons around planets with large intense magnetic and radiation fields (such as the Jovian moons). The effects of extreme temperature cycles in such locations compounds the effects of radiation on structural materials. This paper describes the integrated methodology in detail and shows that it will provide a significant technological advance for designing advanced spacecraft. This methodology will also allow for the development of advanced spacecraft materials through the understanding of the underlying mechanisms of material degradation in the space radiation environment. Thus, this technology holds a promise for revolutionary advances in material damage prediction and protection of space structural components as, for example, in the development of guidelines for managing surveillance programs regarding the integrity of spacecraft components, and the safety of the aging spacecraft. (authors)

  16. 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)

  17. 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.)

  18. The effect of self-shielding of resonance cross sections on the performance of some promising fusion blanket designs

    International Nuclear Information System (INIS)

    The effect of self-shielding of resonance cross sections on the tritium breeding ratio was investigated for three promising fusion blanket designs with liquid lithium, lithium oxide and lithium-lead breeders. Calculations were performed using ANISN and MCNP transport codes with the ENDF/B-V based nuclear data libraries. It is found that the self-shielding effect cannot be neglected in the blanket design if the blanket is neutron leaky in the case when the blanket is thin or with lower Li-6 enrichment in Li. This may result in an underestimate of the tritium breeding ratio if the cross sections are infinitely diluted. This is due to the resonances in the structure materials in which the absorption cross sections are enhanced in the infinitely diluted case. Thus the effect of self-shielding of resonance cross sections should be considered in neutronics calculations of fusion reactors. It is shown that the MCNP results are better reproduced by those from the transport code with the infinitely diluted library. This is probably due to the weight function used to generate the library and to the number of groups considered. Thus for fusion applications it is recommanded to collapse broad group cross sections with the spectrum obtained from an accurate calculation based on many fine groups. (author)

  19. Oxygen plasma damage to blanket and patterned ultralow-κ surfaces

    International Nuclear Information System (INIS)

    Oxygen plasma damage to blanket and patterned ultralow-κ (ULK) dielectric surfaces was investigated by examining the effect of plasma species and dielectric materials. Blanket ULK films and patterned structures were treated by O2 plasma in a remote plasma chamber where the ions and radicals from the plasma source can be separately controlled to study their respective roles in the damage process. The plasma damage was characterized by angle resolved x-ray photoelectron spectroscopy, x-ray reflectivity, and Fourier transform infrared spectroscopy. Studies of the angle dependence of oxygen plasma damage to blanket ULK films indicated that damage by ions was anisotropic while that by radicals was isotropic. Ions were found to play an important role in assisting carbon depletion by oxygen radicals on the blanket film surface. More plasma damage was observed with increasing porosity in ultralow-κ films. Probable reaction paths were proposed by analyzing the reaction by-products. Plasma damage to the sidewall of low-κ trenches was examined by electron energy loss (EELS) analysis. The depletion depth of carbon was found to be related to the penetration of radical species into the porous dielectric and the distribution at the sidewall and trench bottom was affected by the trench pattern geometry, i.e., the aspect ratio, which can be correlated with the electron potential distribution and subsequent trajectory of ions. Vapor silylation was applied for dielectric recovery of trench structure and the result was examined by EELS. The trimethylchlorosilane was found to be effective for recovery of the sidewall carbon loss. The recovery was better for loss induced by radical O2 than by hybrid O2 and the difference was attributed to the surface densification by ions limiting the mass transport of vapor chemicals.

  20. 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)

  1. 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.

  2. 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

  3. Steels from materials science to structural engineering

    CERN Document Server

    Sha, Wei

    2013-01-01

    Steels and computer-based modelling are fast growing fields in materials science as well as structural engineering, demonstrated by the large amount of recent literature. Steels: From Materials Science to Structural Engineering combines steels research and model development, including the application of modelling techniques in steels.  The latest research includes structural engineering modelling, and novel, prototype alloy steels such as heat-resistant steel, nitride-strengthened ferritic/martensitic steel and low nickel maraging steel.  Researchers studying steels will find the topics vital to their work.  Materials experts will be able to learn about steels used in structural engineering as well as modelling and apply this increasingly important technique in their steel materials research and development. 

  4. 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).

  5. Development of the structural materials information center

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission has initiated a Structural Aging Program at the Oak Ridge National Laboratory to identify potential structural safety issues related to continued service of nuclear power plants and to establish criteria for evaluating and resolving these issues. One of the tasks in this program focuses on the establishment of a Structural Materials Information Center where data and information on the time variation of concrete and other structural material properties under the influence of pertinent environmental stressors and aging factors are being collected and assembled into a data base. This data base will be used to assist in the prediction of potential long-term deterioration of critical structural components in nuclear power plants and to establish limits on hostile environmental exposure for these structures and materials. Two complementary data base formats have been developed. The Structural Materials Handbook is an expandable, hard-copy reference document that contains complete sets of data and information for selected portland cement concrete, metallic reinforcement prestressing tendon, and structural steel materials. Baseline data, reference properties and environmental information are presented in the hand book as tables, notes and graphs. The handbook, which will be published inn four volumes, serves as the information source for the electronic data base. The Structural Materials Electronic Data Base is accessible by an IBM-compatible personal computer and provides an efficient means for searching the various data base files to locate materials with similar properties. Properties are reported in the International System of Units (SI) and in customary units whenever possible

  6. Development of the Structural Materials Information Center

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission has initiated a Structural Aging Program at the Oak Ridge National Laboratory to identify potential structural safety issues related to continued service of nuclear power plants and to establish criteria for evaluating and resolving these issues. One of the tasks in this program focuses on the establishment of a Structural Materials Information Center where data and information on the time variation of concrete and other structural material properties under the influence of pertinent environmental stressors and aging factors are being collected and assembled into a data base. This data base will be used to assist in the prediction of potential long-term deterioration of critical structural components in nuclear power plants and to establish limits on hostile environmental exposure for these structures and materials. Two complementary data base formats have been developed. The Structural Materials Handbook is an expandable, hard-copy reference document that contains complete sets of data and information for selected portland cement concrete, metallic reinforcement, prestressing tendon, and structural steel materials. Baseline data, reference properties and environmental information are presented in the handbook as tables, notes and graphs. The handbook, which will be published in four volumes, serves as the information source for the electronic data base. The Structural Materials Electronic Data Base is accessible by an IBM-compatible personal computer and provides an efficient means for searching the various data base files to locate materials with similar properties. Properties will be reported in the International System of Units (SI) and in customary units whenever possible. 7 refs., 3 figs., 4 tabs

  7. Data base on structural materials aging properties

    Energy Technology Data Exchange (ETDEWEB)

    Oland, C.B.

    1992-03-01

    The US Nuclear Regulatory Commission has initiated a Structural Aging Program at the Oak Ridge National Laboratory to identify potential structural safety issues related to continued service of nuclear power plants and to establish criteria for evaluating and resolving these issues. One of the tasks in this program focuses on the establishment of a Structural Materials Information Center where long-term and environment-dependent properties of concretes and other structural materials are being collected and assembled into a data base. These properties will be used to evaluate the current condition of critical structural components in nuclear power plants and to estimate the future performance of these materials during the continued service period.

  8. Smart materials and structures: what are they?

    Science.gov (United States)

    Spillman, W. B., Jr.; Sirkis, J. S.; Gardiner, P. T.

    1996-06-01

    There has been considerable discussion in the technical community on a number of questions concerned with smart materials and structures, such as what they are, whether smart materials can be considered a subset of smart structures, whether a smart structure and an intelligent structure are the same thing, etc. This discussion is both fueled and confused by the technical community due to the truly multidisciplinary nature of this new field. Smart materials and structures research involves so many technically diverse fields that it is quite common for one field to completely misunderstand the terminology and start of the art in other fields. In order to ascertain whether a consensus is emerging on a number of questions, the technical community was surveyed in a variety of ways including via the internet and by direct contact. The purpose of this survey was to better define the smart materials and structures field, its current status and its potential benefits. Results of the survey are presented and discussed. Finally, a formal definition of the field of smart materials and structures is proposed.

  9. 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.)

  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. 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)

  12. Understanding structural conservation through materials science:

    DEFF Research Database (Denmark)

    Fuster-López, Laura; Krarup Andersen, Cecil

    2014-01-01

    Mechanical properties and the structure of materials are key elements in understanding how structural interventions in conservation treatments affect cultural heritage objects. In this context, engineering mechanics can help determine the strength and stability found in art objects as it can prov...

  13. Structural materials for fission & fusion energy

    Directory of Open Access Journals (Sweden)

    Steven J. Zinkle

    2009-11-01

    Full Text Available Structural materials represent the key for containment of nuclear fuel and fission products as well as reliable and thermodynamically efficient production of electrical energy from nuclear reactors. Similarly, high-performance structural materials will be critical for the future success of proposed fusion energy reactors, which will subject the structures to unprecedented fluxes of high-energy neutrons along with intense thermomechanical stresses. Advanced materials can enable improved reactor performance via increased safety margins and design flexibility, in particular by providing increased strength, thermal creep resistance and superior corrosion and neutron radiation damage resistance. In many cases, a key strategy for designing high-performance radiation-resistant materials is based on the introduction of a high, uniform density of nanoscale particles that simultaneously provide good high temperature strength and neutron radiation damage resistance.

  14. Preparation of Nickel Materials with Fractal Structure

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    A way of manufacturing nickel material with fractal structure has been studied. Some algae with natural fractalstructure were used as the basic substrates. The nickel was coated on the substrates by both electroless depositionand electrodeposition. After elimination of the foundational algae by erosion, dissolution etc, the pure nickel materialswith fractal structure were obtained. At last, the specific surface area was analyzed by BET analyses and the fractaldimension of the nickel material was calculated by means of box-counting technique. The comparison of fractaldimension between Ni structure and natural algae was also given.

  15. 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)

  16. Structure and thermal stability of nanocrystalline materials

    Indian Academy of Sciences (India)

    B S Murty; M K Datta; S K Pabi

    2003-02-01

    Nanocrystalline materials, which are expected to play a key role in the next generation of human civilization, are assembled with nanometre-sized “building blocks” consisting of the crystalline and large volume fractions of intercrystalline components. In order to predict the unique properties of nanocrystalline materials, which are a combination of the properties of the crystalline and intercrystalline regions, it is essential to understand precisely how the structures of crystalline and intercrystalline regions vary with decrease in crystallite size. In addition, study of the thermal stability of nanocrystalline materials against significant grain growth is both scientific and technological interest. A sharp increase in grain size (to micron levels) during consolidation of nanocrystalline powders to obtain fully dense materials may consequently result in the loss of some unique properties of nanocrystalline materials. Therefore, extensive interest has been generated in exploring the size effects on the structure of crystalline and intercrystalline region of nanocrystalline materials, and the thermal stability of nanocrystalline materials against significant grain growth. The present article is aimed at understanding the structure and stability of nanocrystalline materials.

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

    International Nuclear Information System (INIS)

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

  18. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

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

  19. Structural materials challenges for fusion power systems

    International Nuclear Information System (INIS)

    Full text: Structural materials in a fusion power system must function in an extraordinarily demanding environment that includes various combinations of high temperatures, reactive chemicals, time-dependent thermal and mechanical stresses, and intense damaging radiation. The fusion neutron environment produces displacement damage equivalent to displacing every atom in the material about 150 times during its expected service life, and changes in chemical composition by transmutation reactions, which includes creation of reactive and insoluble gases. Fundamental materials challenges that must be resolved to effectively harness fusion power include (1) understanding the relationships between material strength, ductility and resistance to cracking, (2) development of materials with extraordinary phase stability, high-temperature strength and resistance to radiation damage, (3) establishment of the means to control corrosion of materials exposed to aggressive environments, (4) development of technologies for large-scale fabrication and joining, and (5) design of structural materials that provide for an economically attractive fusion power system while simultaneously achieving safety and environmental acceptability goals. The most effective approach to solve these challenges is a science-based effort that couples development of physics-based, predictive models of materials behavior with key experiments to validate the models. The U.S. Fusion Materials Sciences program is engaged in an integrated effort of theory, modeling and experiments to develop structural materials that will enable fusion to reach its safety, environmental and economic competitiveness goals. In this presentation, an overview of recent progress on reduced activation ferritic/martensitic steels, nanocomposited ferritic alloys, and silicon carbide fiber reinforced composites for fusion applications will be given

  20. Blanket Cooling Concepts and Heat Conversion Cycles for Controlled Thermonuclear Reactors

    International Nuclear Information System (INIS)

    The neutron energy generated in fusion plasmas produces heat in solid and liquid blanket regions which confine the fusion plasma. This heat has to be removed by cooling the blanket, and is then converted into electric energy by thermodynamic processes. Blanket cooling can directly be achieved by cycling the liquid blanket material through an outside cooler, or indirectly by leading gaseous or liquid coolants through pipes in the blanket. For the conversion of heat into electricity, single Rankine or Brayton cycles can be applied as well as binary Rankine cycles using potassium and steam as cycle fluid. In this paper, the special features of different systems for blanket cooling and for heat conversion are described and discussed. The thermodynamic requirements for favourable operation of the different heat conversion cycles, and those for the heat removal system from the CTR ate pointed out. Also the pumping power in magnetically or unmagnetically coolant flows is considered. Selected solutions for combining the systems of blanket cooling and of heat conversion are compared with respect to plant efficiency, lost-volume fraction in the lithium blanket region, and the expected constructional and safety features. The comparisons are made for thermal powers of the reactor between 5000 and 10 000 MW and toroidal reactor configurations. Solutions to be preferred are pointed out. (author)

  1. Beryllium data base for in-pile mockup test on blanket of fusion reactor, (1)

    International Nuclear Information System (INIS)

    Beryllium has been used in the fusion blanket designs with ceramic breeder as a neutron multiplier to increase the net tritium breeding ratio (TBR). The properties of beryllium, that is physical properties, chemical properties, thermal properties, mechanical properties, nuclear properties, radiation effects, etc. are necessary for the fusion blanket design. However, the properties of beryllium have not been arranged for the fusion blanket design. Therefore, it is indispensable to check and examine the material data of beryllium reported previously. This paper is the first one of the series of papers on beryllium data base, which summarizes the reported material data of beryllium. (author)

  2. Dynamic and structural control utilizing smart materials and structures

    Science.gov (United States)

    Rogers, C. A.; Robertshaw, H. H.

    1989-01-01

    An account is given of several novel 'smart material' structural control concepts that are currently under development. The thrust of these investigations is the evolution of intelligent materials and structures superceding the recently defined variable-geometry trusses and shape memory alloy-reinforced composites; the substances envisioned will be able to autonomously evaluate emergent environmental conditions and adapt to them, and even change their operational objectives. While until now the primary objective of the developmental efforts presently discussed has been materials that mimic biological functions, entirely novel concepts may be formulated in due course.

  3. 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

  4. 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.

  5. Layer like porous materials with hierarchical structure.

    Science.gov (United States)

    Roth, Wieslaw J; Gil, Barbara; Makowski, Wacław; Marszalek, Bartosz; Eliášová, Pavla

    2016-06-13

    Many chemical compositions produce layered solids consisting of extended sheets with thickness not greater than a few nanometers. The layers are weakly bonded together in a crystal and can be modified into various nanoarchitectures including porous hierarchical structures. Several classes of 2-dimensional (2D) materials have been extensively studied and developed because of their potential usefulness as catalysts and sorbents. They are discussed in this review with focus on clays, layered transition metal oxides, silicates, layered double hydroxides, metal(iv) phosphates and phosphonates, especially zirconium, and zeolites. Pillaring and delamination are the primary methods for structural modification and pore tailoring. The reported approaches are described and compared for the different classes of materials. The methods of characterization include identification by X-ray diffraction and microscopy, pore size analysis and activity assessment by IR spectroscopy and catalytic testing. The discovery of layered zeolites was a fundamental breakthrough that created unprecedented opportunities because of (i) inherent strong acid sites that make them very active catalytically, (ii) porosity through the layers and (iii) bridging of 2D and 3D structures. Approximately 16 different types of layered zeolite structures and modifications have been identified as distinct forms. It is also expected that many among the over 200 recognized zeolite frameworks can produce layered precursors. Additional advances enabled by 2D zeolites include synthesis of layered materials by design, hierarchical structures obtained by direct synthesis and top-down preparation of layered materials from 3D frameworks. PMID:26489452

  6. 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

  7. 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

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

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

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

    International Nuclear Information System (INIS)

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

  10. Activation analyses for the different options considered in the US ITER blanket trade-off study

    International Nuclear Information System (INIS)

    Detailed activation analyses were performed for the different blanket design options considered in the ITER blanket option trade-off study. The options considered included a self-cooled Li/V option, a helium-cooled Li/V option and a water-cooled 316 SS non-breeding shield option. A vacuum vessel made of double-wall Inconel 625 and water-cooled 316 SS balls was used with all options. The He-cooled blanket activity is higher than that of the self-cooled blanket due to the larger structure content. Meanwhile, the vacuum vessel activity is lower for the He-cooled blanket option due to the larger neutron attenuation in the blanket. The shield activity and decay heat of the 316 SS/H2O option are higher than those of the Li/V blankets due to the large amount (80%) of 316 SS used. In both Li/V options the blanket qualifies as class C low-level waste. On the other hand, the 316 SS/H2O shield does not qualify for disposal as low- level waste. The 316 SS/H2O option produces the highest off-site doses in the case of accidental release of 100% of its radioactive inventory. Only remote maintenance would be allowed for all options. (orig.)

  11. 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.)

  12. 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

  13. Hierarchically structured materials for lithium batteries

    International Nuclear Information System (INIS)

    The lithium-ion battery (LIB) is one of the most promising power sources to be deployed in electric vehicles, including solely battery powered vehicles, plug-in hybrid electric vehicles, and hybrid electric vehicles. With the increasing demand for devices of high-energy densities (>500 Wh kg−1), new energy storage systems, such as lithium–oxygen (Li–O2) batteries and other emerging systems beyond the conventional LIB, have attracted worldwide interest for both transportation and grid energy storage applications in recent years. It is well known that the electrochemical performance of these energy storage systems depends not only on the composition of the materials, but also on the structure of the electrode materials used in the batteries. Although the desired performance characteristics of batteries often have conflicting requirements with the micro/nano-structure of electrodes, hierarchically designed electrodes can be tailored to satisfy these conflicting requirements. This work will review hierarchically structured materials that have been successfully used in LIB and Li–O2 batteries. Our goal is to elucidate (1) how to realize the full potential of energy materials through the manipulation of morphologies, and (2) how the hierarchical structure benefits the charge transport, promotes the interfacial properties and prolongs the electrode stability and battery lifetime. (paper)

  14. Hierarchically structured materials for lithium batteries

    Science.gov (United States)

    Xiao, Jie; Zheng, Jianming; Li, Xiaolin; Shao, Yuyan; Zhang, Ji-Guang

    2013-10-01

    The lithium-ion battery (LIB) is one of the most promising power sources to be deployed in electric vehicles, including solely battery powered vehicles, plug-in hybrid electric vehicles, and hybrid electric vehicles. With the increasing demand for devices of high-energy densities (>500 Wh kg-1), new energy storage systems, such as lithium-oxygen (Li-O2) batteries and other emerging systems beyond the conventional LIB, have attracted worldwide interest for both transportation and grid energy storage applications in recent years. It is well known that the electrochemical performance of these energy storage systems depends not only on the composition of the materials, but also on the structure of the electrode materials used in the batteries. Although the desired performance characteristics of batteries often have conflicting requirements with the micro/nano-structure of electrodes, hierarchically designed electrodes can be tailored to satisfy these conflicting requirements. This work will review hierarchically structured materials that have been successfully used in LIB and Li-O2 batteries. Our goal is to elucidate (1) how to realize the full potential of energy materials through the manipulation of morphologies, and (2) how the hierarchical structure benefits the charge transport, promotes the interfacial properties and prolongs the electrode stability and battery lifetime.

  15. Ordered mesoporous silica materials with complicated structures

    KAUST Repository

    Han, Yu

    2012-05-01

    Periodically ordered mesoporous silicas constitute one of the most important branches of porous materials that are extensively employed in various chemical engineering applications including adsorption, separation and catalysis. This short review gives an introduction to recently developed mesoporous silicas with emphasis on their complicated structures and synthesis mechanisms. In addition, two powerful techniques for solving complex mesoporous structures, electron crystallography and electron tomography, are compared to elucidate their respective strength and limitations. Some critical issues and challenges regarding the development of novel mesoporous structures as well as their applications are also discussed. © 2011 Elsevier Ltd.

  16. Safety analysis on the Korean He-Cooled Molten Lithium Test Blanket Module for the ITER

    International Nuclear Information System (INIS)

    Through a consideration of the requirements for a DEMO-relevant blanket concept, a He Cooled Molten Lithium (HCML) blanket with Ferritic Steel (FS) as a structural material is proposed to be tested in the International Thermonuclear Experimental Reactor (ITER). The HCML Test Blanket Module (TBM) uses He as a coolant at an inlet temperature of 300 deg C and an outlet temperature up to 376 deg C and Li is used as a tritium breeder by considering its potential advantages. Two layers of a graphite reflector are inserted as a reflector in the breeder zone to increase the Tritium Breeding Ratio (TBR) and the shielding performances. A 3-D Monte Carlo analysis is performed with the MCCARD code for the neutronics and the total TBM power is designed to be 0.739 MW at a normal heat flux from the plasma side. From the analysis results with CFX-10 for the thermal-hydraulics, the He cooling path is optimized and it shows that the maximum temperature of the first wall does not exceed 550 deg C at the structural materials and the coolant velocities are 45 m/sec and 8.2 m/sec in the first wall and breeding zone, respectively. The obtained temperature data is used in the thermal-mechanical analysis with ANSYS-10. The maximum von Mises equivalent stress of the first wall is 123 MPa and the maximum deformation of it is 3.73 mm, which is lower than the maximum allowable stress. The KO HCML is being designed and optimized from the current design. Since the safety analysis related to the postulated accident is essential for both licensing and acceptance for installation in ITER, the relatively severe cases were assumed for the safety assessment; (1) active plasma shut-down after delayed accident detection with disruption and (2) no active plasma shut-down. The safety analysis performed for both cases show the capability of decay heat removal in both cases. (author)

  17. A Helium Cooled Molten Lithium test blanket module for the ITER in Korea

    International Nuclear Information System (INIS)

    Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) blanket with Ferritic Steel (FS) as a structural material as part of the International Thermonuclear Experimental Reactor (ITER) program. The preliminary design and the performance of the KO HCML Test Blanket Module (TBM) are introduced in this paper. It uses He as a coolant at an inlet temperature of 300 C and an outlet temperature up to 400 C and Li is used as a tritium breeder by considering its potential advantages. Two layers of graphite are inserted as a reflector in the breeder zone to increase the Tritium Breeding Ratio (TBR) and the shielding performances. A 3-D Monte Carlo analysis is performed with the MCCARD code for the neutronics evaluation of the KO HCML and the total TBM power is designed to be 0.739 MW at a normal heat flux from the plasma side. From the analysis results with CFX-10 for the thermal-hydraulics evaluation, the He cooling path is determined and it shows that the maximum temperature of the first wall does not exceed 550 C for the structural materials and the coolant velocities are 45 m/sec and 8.2 m/sec for the first wall and breeding zone, respectively. The obtained temperature data was used in the thermal-mechanical analysis with ANSYS-10. The maximum von Mises equivalent stress of the first wall was 123 MPa and the maximum deformation of it was 3.73 mm, which is lower than the maximum allowable stress. And also, for the several accident scenarios such as a Loss of Coolant Accident (LOCA), a safety analysis is being performed. (orig.)

  18. Proceedings of the 2nd international workshop on electromagnetic forces and related effects on blankets and other structures surrounding the fusion plasma torus

    International Nuclear Information System (INIS)

    This publication is the collection of the papers presented at the title meeting. The subjects of the papers presented were categorized in six parts and are contained in this volume. In the first part, the valuable experiences are presented concerning electromagnetic phenomena in existing large devices or those under construction. In the 2nd part, the papers are mainly concerning on the evaluation of the electromagnetic fields and forces for the next experimental reactors. In the 3rd part, electromagnetomechanical coupling problems were treated by numerical and experimental approaches. In the part 4, numerical and experimental approaches for ferromagnetic structures are performed. In the 5th part, papers related to the structural integrity evaluation are presented. The part 6 is devoted to the proposal of the intelligent material system. A summary of the panel discussion held at the final session of the workshop is also included at the end of this volume. The 22 of the presented papers are indexed individually. (J.P.N.)

  19. 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

  20. 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

  1. US DCLL test blanket module design and relevance to DEMO

    International Nuclear Information System (INIS)

    Full text: In the design of Test Blanket Modules (TBMs) for ITER, it is required to provide a design concept that is demonstration power reactor (DEMO) relevant. It should be noted that in the US, DEMO is defined to be a good representation of the first generation fusion power reactor. In order to evaluate the potential of the US TBM design for DEMO, a system evaluation of DEMO design was performed with an improved GA system code, and the physics results were benchmarked to ITER. With the selection of ferritic steel as the structural material, the maximum neutron wall loading is limited to 3 MW/m2. When designed to a 3 GW fusion device the optimum aspect ratio is found to be in the range of 2.5 to 3. Results show that the US dual coolant lead-lithium (DCLL) blanket can satisfy all the DEMO design requirements. On the chamber wall material, for the ITER-TBM design, the design guidance is to apply a 2 mm Be layer onto the plasma facing surface. When extrapolated to the DEMO design, the Be layer will not be suitable due to radiation damage. Similarly, a carbon surface will not be suitable due to high physical and chemical sputtering rates, radiation damage of the material and potential large retention of tritium. Unfortunately, the remaining commonly proposed material, tungsten (W), would suffer radiation damage from alpha particle implantation and, with blistering, W transport to the plasma core could severely limit the core performance. To resolve this potential impasse, different innovative options were evaluated. All high performance tokamak experiments presently use boron or silicon to condition the first wall. To use boron in DEMO, it is found that in-situ boronization will be required in order to maintain a boronized layer on the chamber wall. This boronized layer could also protect the W substrate, while retaining low-Z wall characteristics. Further innovative ideas are being evaluated to handle transient events like ELMs and disruptions. TOPICS: (PPCA) Power

  2. 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

  3. Fullerenic structures and such structures tethered to carbon materials

    Science.gov (United States)

    Goel, Anish; Howard, Jack B.; Vander Sande, John B.

    2010-01-05

    The fullerenic structures include fullerenes having molecular weights less than that of C.sub.60 with the exception of C.sub.36 and fullerenes having molecular weights greater than C.sub.60. Examples include fullerenes C.sub.50, C.sub.58, C.sub.130, and C.sub.176. Fullerenic structure chemically bonded to a carbon surface is also disclosed along with a method for tethering fullerenes to a carbon material. The method includes adding functionalized fullerene to a liquid suspension containing carbon material, drying the suspension to produce a powder, and heat treating the powder.

  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. 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

  6. Materials and structures under shock and impact

    CERN Document Server

    Bailly, Patrice

    2013-01-01

    In risk studies, engineers often have to consider the consequences of an accident leading to a shock on a construction. This can concern the impact of a ground vehicle or aircraft, or the effects of an explosion on an industrial site.This book presents a didactic approach starting with the theoretical elements of the mechanics of materials and structures, in order to develop their applications in the cases of shocks and impacts. The latter are studied on a local scale at first. They lead to stresses and strains in the form of waves propagating through the material, this movement then extending

  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. Structural materials challenges for advanced reactor systems

    Science.gov (United States)

    Yvon, P.; Carré, F.

    2009-03-01

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

  9. The electronic structure of hard materials

    Science.gov (United States)

    Winarski, Robert Paul

    This research dissertation involves an experimental as well as a theoretical examination of the electronic structure of hard materials. The materials that are presented in this dissertation cover a wide class of materials, consisting of transition metal borides, irradiated polymer films, theoretically predicted superhard semiconductors, doped intermetallic alloys, and transition metal carbides. The borides are traditionally used in high temperature, hard coating applications, such as rocket nozzle linings, extreme wear surfaces, and corrosion coatings. Measurements of the borides appear to show that the bonding in these hard materials is primarily between the boron atoms in these systems. Also of note are the remarkably short interatomic distances between the boron atoms and between the boron and metal atoms in these materials. Irradiated polymer films are being developed for electronic applications, in the hopes that circuits can be developed that can benefit from the high thermal stability, dielectric properties, and mechanical properties provided by these materials. C3N4 is a theoretically predicted superhard material, and some of the first soft x-ray emission measurements of well-characterized samples of this compound are discussed in this work. Intermetallic alloys, in particular Ni3Al, are rather hard, but brittle metallic alloys. It has been found that the addition of boron atoms, in rather low concentrations, can increase the ductility of these alloys, allowing them to be utilized in a wider variety of applications. Measurements of this system have examined a question regarding the positioning of the boron atoms in the structures of this alloy. Finally, the transition metal carbides are used extensively as coatings in industrial applications such as cutting and grinding tools, and polishing compounds. Measurements of these materials suggest that the high degree of covalency between the metal and carbon atoms is primarily responsible for the hardness of

  10. Nondestructive Testing of Materials and Structures

    CERN Document Server

    Akkaya, Yılmaz

    2013-01-01

    Condition assessment and characterization of materials and structures by means of nondestructive testing (NDT) methods is a priority need around the world to meet the challenges associated with the durability, maintenance, rehabilitation, retrofitting, renewal and health monitoring of new and existing infrastructures including historic monuments. Numerous NDT methods that make use of certain components of the electromagnetic and acoustic spectra are currently in use to this effect with various levels of success and there is an intensive worldwide research effort aimed at improving the existing methods and developing new ones. The knowledge and information compiled in this book captures the current state-of-the-art in NDT methods and their application to civil and other engineering materials and structures. Critical reviews and advanced interdisciplinary discussions by world-renowned researchers point to the capabilities and limitations of the currently used NDT methods and shed light on current and future res...

  11. Nonlinearity in structural and electronic materials

    International Nuclear Information System (INIS)

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The project strengthens a nonlinear technology base relevant to a variety of problems arising in condensed matter and materials science, and applies this technology to those problems. In this way the controlled synthesis of, and experiments on, novel electronic and structural materials provide an important focus for nonlinear science, while nonlinear techniques help advance the understanding of the scientific principles underlying the control of microstructure and dynamics in complex materials. This research is primarily focused on four topics: (1) materials microstructure: growth and evolution, and porous media; (2) textures in elastic/martensitic materials; (3) electro- and photo-active polymers; and (4) ultrafast photophysics in complex electronic materials. Accomplishments included the following: organization of a ''Nonlinear Materials'' seminar series and international conferences including ''Fracture, Friction and Deformation,'' ''Nonequilibrium Phase Transitions,'' and ''Landscape Paradigms in Physics and Biology''; invited talks at international conference on ''Synthetic Metals,'' ''Quantum Phase Transitions,'' ''1996 CECAM Euroconference,'' and the 1995 Fall Meeting of the Materials Research Society; large-scale simulations and microscopic modeling of nonlinear coherent energy storage at crack tips and sliding interfaces; large-scale simulation and microscopic elasticity theory for precursor microstructure and dynamics at solid-solid diffusionless phase transformations; large-scale simulation of self-assembling organic thin films on inorganic substrates; analysis and simulation of smoothing of rough atomic surfaces; and modeling and analysis of flux pattern formation in equilibrium and nonequilibrium Josephson junction arrays and layered superconductors

  12. Nonlinearity in structural and electronic materials

    Energy Technology Data Exchange (ETDEWEB)

    Bishop, A.R.; Beardmore, K.M.; Ben-Naim, E. [and others

    1997-11-01

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The project strengthens a nonlinear technology base relevant to a variety of problems arising in condensed matter and materials science, and applies this technology to those problems. In this way the controlled synthesis of, and experiments on, novel electronic and structural materials provide an important focus for nonlinear science, while nonlinear techniques help advance the understanding of the scientific principles underlying the control of microstructure and dynamics in complex materials. This research is primarily focused on four topics: (1) materials microstructure: growth and evolution, and porous media; (2) textures in elastic/martensitic materials; (3) electro- and photo-active polymers; and (4) ultrafast photophysics in complex electronic materials. Accomplishments included the following: organization of a ``Nonlinear Materials`` seminar series and international conferences including ``Fracture, Friction and Deformation,`` ``Nonequilibrium Phase Transitions,`` and ``Landscape Paradigms in Physics and Biology``; invited talks at international conference on ``Synthetic Metals,`` ``Quantum Phase Transitions,`` ``1996 CECAM Euroconference,`` and the 1995 Fall Meeting of the Materials Research Society; large-scale simulations and microscopic modeling of nonlinear coherent energy storage at crack tips and sliding interfaces; large-scale simulation and microscopic elasticity theory for precursor microstructure and dynamics at solid-solid diffusionless phase transformations; large-scale simulation of self-assembling organic thin films on inorganic substrates; analysis and simulation of smoothing of rough atomic surfaces; and modeling and analysis of flux pattern formation in equilibrium and nonequilibrium Josephson junction arrays and layered superconductors.

  13. Strength of structural materials of nuclear reactors

    International Nuclear Information System (INIS)

    The monography reviews the phenomenon of stress-corrosion craving of zirconium-alloy fuel cans in the nuclear fuel fission products and ways of its prevention. Equations of creep and limiting state of FBR core materials were derived on the basis of the concept of deformation processes unity taking into account the degree of structural stability of alloys, temperature, nonstationary loading and aggressive media effects. Equations of durability under joint quasistatic and cyclic loading are developed. 146 refs.; 91 figs.; 16 tabs

  14. 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.)

  15. APT Blanket Thermal Analysis of Cavity Flood Cooling with a Beam Window Break; FINAL

    International Nuclear Information System (INIS)

    The cavity flood system is designed to be the primary safeguard for the integrity of the blanket modules and target assemblies during loss of coolant accidents, LOCA''s. In the unlikely event that the internal flow passages in a blanket module or a target assembly dryout, decay heat in the metal structures will be dissipated to the cavity flood system through the module or assembly walls. This study supplements the two previous studies by demonstrating that the cavity flood system can adequately cool the blanket modules when the cavity vessel beam window breaks

  16. 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.)

  17. Thermal-hydraulic design and analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and being designed aiming at 50-200 MW fusion power, 30-50% duty time factor, and tritium self-sustained. Three kinds of tritium breeding blanket concepts, including helium-cooled solid blanket, water-cooled solid blanket and liquid metal-cooled liquid blanket, have been considered for CFETR. Compared to ITER test blanket module, the blanket design for CFETR is facing much more challenges due to the compulsive requirements of tritium self-sufficiency, nuclear heat removal and the space limitation for blanket installation. In this paper, a kind of helium cooled solid tritium breeder blanket was designed for CFETR full superconducting tokamak. The thermal-hydraulic designs were carried out based on the blanket structure design and neutronics calculation. The performance evaluation was conducted using ANSYS, and three-dimensional fluid-solid coupled models were modeled for the accuracy results. The results showed that the FW and BU can satisfy the design requirements. (author)

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

    International Nuclear Information System (INIS)

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

  19. Structural materials: New challenges, manufacturing and performance

    International Nuclear Information System (INIS)

    Full texct: This paper reviews international perspectives on materials, manufacturing and performance on structural materials for fast reactors. It is recognized that fast reactors with closed fuel cycle shall play an eminent and major role to realize energy sustainability. Large scale exploitation of fast reactors shall require meeting of sustainability requirements such as economic competitiveness, safe and optimized waste management, increased proliferation resistance, improved use of uranium and thorium and enhanced efficiency. The designers of these reactors along with material specialists have a key role in meeting the above mentioned criteria of sustainability. The paper shall describe features of advanced sodium cooled fast reactors and the resulting requirements on the performance of the structural materials. The choice of the fuel governs the cladding material for these reactors. Considering the choice as oxide, carbide, metallic with or without minor actinides; cladding materials are chosen based on performance modeling, available experience and expertise. Improved varieties of 316 austenitic stainless steels and oxide dispersion ferritice-martensitic steels emerge as front runners to meet the requirements. The developments in Ti modified 316 austenitic stainless steels give enough confidence to take these fuels to burn-up of 120,000 MWd/t. It is inferred that the target burn-up of 250,000 MWd/t would only be achieved with oxide dispersion strengthened iron- chromium base steels. Research and development of modified austenitic stainless steels and ODS alloys in Japan, Europe and India would be described in the paper. For achieving high burn-up of 250,000 MWd/t, ferritic-martensitic steel has emerged as the choice for the wrapper material. The current status of development and the challenges shall be described in the paper. In particular, the European contribution to this paper addresses the development and performance assessment of the preferred

  20. 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

  1. 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....

  2. Protection and structural materials for fusion reactors: Main challenges, long term materials development needs, differences and commonalities with fission

    International Nuclear Information System (INIS)

    Developing plasma-facing and structural materials with Low nuclear Activation (LA), high heat and radiation resistance, is a challenge that the materials community has to address to enable reliable development and safe operation of fusion reactors in the future. This requires scientific innovation, new knowledge, and the exploration of a range of new materials. To address this challenge, the EU fusion programme has set up a Fusion Materials Topical Group to strengthen coordination of long-term fusion materials development for DEMO, and to undertake physically based modelling of radiation induced microstructure and degradation of mechanical properties required to guide the experimental developments. In this paper we describe main radiation effects induced by the intense flux of 14 MeV neutrons in the reference structural materials (i) for Tritium-Breeding Blanket (TBB) modules, LA Ferritic/Martensitic Steel EUROFER, Oxide Dispersion (ODS) EUROFER and ODS ferritic steels, and, (ii) for the divertor, W and W-alloys. Specific issues concerning the peculiar microstructure induced by the impact of α particles on the surface of tungsten, foreseen as a reference protection material, will also be discussed. Modelling radiation effects in EUROFER under fusion reactor relevant conditions is the first priority for the programme, in order to inter-correlate the data obtained with various neutron spectra, contribute to the definition of the irradiation matrix in the future intense source of 14 MeV neutrons, IFMIF, and bring comprehensive understanding and extrapolation capabilities towards the very large range of DEMO operating conditions. Formulating and developing predictive modelling tools is therefore a task of prime significance. Particular care has been taken (i) to focus the modelling effort on the scale where physics can be mastered, i.e. on the scale of the chemical bond, which is triggered by electronic correlation in the case of Fe-Cr system and body cubic centred

  3. 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

  4. 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.

  5. Basic concepts of DEMO and a design of a helium-cooled molten lithium blanket for a testing in ITER

    International Nuclear Information System (INIS)

    Basic concepts and the performance of DEMO for an early realization have been investigated with a limited extension of its plasma physics and technology from the second phase of the International Thermonuclear Experimental Reactor (ITER) operation (EPP phase). With the same plasma size as that of ITER, net electric power up to 600 MW is possible with βN > 4.0, H > 1.0 and a divertor heat load of Hdiv 2. Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea has proposed a He cooled molten lithium (HCML) blanket as an ITER TBM. It uses He as a coolant and Li is used as a tritium breeder by considering its potential advantages. Low activation Ferritic Steel (FS) is used as a structural material and two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. The design and the performance of the KO HCML test blanket module (TBM) are being modified in terms of its He coolant efficiency and its optimized path with a performance analysis; with a 3D Monte Carlo analysis (MCCARD code) for the neutronics; with the CFD code (CFX-10) for the thermal-hydraulics; with ANSYS-10 for the thermo-mechanical analysis

  6. Design and preliminary safety analysis of a helium cooled molten lithium test blanket module for the ITER in Korea

    International Nuclear Information System (INIS)

    Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) blanket with ferritic steel (FS) as a structural material in the International Thermonuclear Experimental Reactor (ITER) program. The design and the performance of the KO HCML test blanket module (TBM) and the preliminary results of the safety analyses such as activation, decay heat, and accident analysis by a loss of coolant are introduced briefly in this paper. KO HCML TBM uses He as a coolant and Li is used as a tritium breeder by considering its potential advantages. Two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. Performance analyses were performed with the MCCARD code for the neutronics, the CFX-10 code for the thermal-hydraulics, and with the ANSYS-10 code for the thermal-mechanical analysis. For the safety analyses, the activation and decay heat were obtained from the MCCARD and Origen codes. From the obtained decay heat, an accident analysis was performed

  7. Ageing in civil engineering materials and structures

    International Nuclear Information System (INIS)

    SETEC TPI will address the 'Aging' topic of the Dijon Symposium by talking about: aging in civil engineering materials and structures, prevention of aging phenomena, in-operation monitoring of degradations related to aging and compensatory measures required to maintain a good safety level. Works as the Millau viaduct, the EdF skyscraper at La Defense - Paris, the renovation of the Grand Palais of Paris and special structures with Monaco's floating dam as well as the 'number 10' shaped gateway boat at Marseilles are illustrations for the issues discussed. The durability of civil engineering structures has become a major concern for designers. The Millau viaduct is designed for a service life of 120 years, and the Monaco dam for 100 years. Calculation rules have been evolving toward the incorporation of the concept of life cycle, for example, the Eurocodes 2 rules (reinforced concrete). The talk will expose the factors which are being taken into account to delay aging versus structure types. This part will be focused towards materials and corresponding regulations: - Reinforced concrete (coating of reinforcements, opening of cracks, choice of reinforcement types), BAEL and Eurocodes 2 rules; - Frame steel (protection, sacrificial anode), CM66 and Eurocodes 3 rules. New materials will also be mentioned: - Ultra high-performance fiber/concrete, with the example of CERACEM applied at Millau for the covering of the toll area barrier; - Titanium, which is starting to appear in the building trades, as for instance for the Beijing China Opera House shell. The second part of the talk will be devoted to a specific case namely, the 'number 10' shaped gateway bridge, a prestressed concrete structure immersed in the Port of Marseilles, which will be used to illustrate the aging phenomenon in a corrosive environment. We will focus on the types of inspection series performed by the Autonomous Port Authority of Marseilles to check the behavior of its structure and the repair series

  8. Structural Materials: New Challenges, Manufacturing and Performance

    International Nuclear Information System (INIS)

    Important criteria of innovative fast reactors and advanced fuel cycle initiatives are improved efficiency, economic competitiveness and reduction of waste. To reach these goals and keep high safety standards, at least at the level of currently operating nuclear reactors, key issues are the availability of suitable structural materials and their performance assessment. The authors, on the basis of the wealth of experience gained in the European Union, India and Japan, aim to define the challenges and current status of material development and set the agenda for R and D in the coming years and decades. It is hoped that the joint perspective would enable realizing the expected criteria of sustainability envisaged through sodium cooled fast reactors and closed fuel cycles. (author)

  9. Adhesion of Dental Materials to Tooth Structure

    Science.gov (United States)

    Mitra, Sumita B.

    2000-03-01

    The understanding and proper application of the principles of adhesion has brought forth a new paradigm in the realm of esthetic dentistry. Modern restorative tooth procedures can now conserve the remaining tooth-structure and also provide for the strengthening of the tooth. Adhesive restorative techniques call for the application and curing of the dental adhesive at the interface between the tooth tissue and the filling material. Hence the success of the restoration depends largely on the integrity of this interface. The mechanism of adhesion of the bonding materials to the dental hard tissue will be discussed in this paper. There are four main steps that occur during the application of the dental adhesive to the oral hard tissues: 1) The first step is the creation of a microstructure in the tooth enamel or dentin by means of an acidic material. This can be through the application of a separate etchant or can be accomplished in situ by the adhesive/primer. This agent has to be effective in removing or modifying the proteinaceous “smear” layer, which would otherwise act as a weak boundary layer on the surface to be bonded. 2) The primer/adhesive must then be able to wet and penetrate the microstructure created in the tooth. Since the surface energies of etched enamel and that of etched dentin are different finding one material to prime both types of dental tissues can be quite challenging. 3) The ionomer types of materials, particularly those that are carboxylate ion-containing, can chemically bond with the calcium ions of the hydroxyapatite mineral. 4) Polymerization in situ allows for micromechanical interlocking of the adhesive. The importance of having the right mechanical properties of the cured adhesive layer and its role in absorbing and dissipating stresses encountered by a restored tooth will also be discussed.

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

    International Nuclear Information System (INIS)

    High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease magnetohydrodynamic resistance authors propose to form insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the insulating coatings characteristics ρδ is ∼ 10-5 Ohm·m2 for steels and 5,0x10-6 - 5,0x10-5 Ohm·m2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steamgenerators and equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem of technology of lead and lead-bismuth coolants for power high temperature radioactive facilities has been solved. Accidents, emergency situations such as leakage of steamgenerators or depressurization of gas system in facilities with lead and lead-bismuth coolants have been explored and suppressed. (author)

  11. LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-08

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  12. 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.)

  13. 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

  14. Producing and optimizing novel materials and structures

    Science.gov (United States)

    Ashrafi, Mahdi

    2011-12-01

    A series of detailed experimental and finite element investigations were carried out to study the response of selected objects which are currently utilized for load carrying. These investigations were later applied to optimize the mechanical performance of the studied structures and materials. First, a number of experiments and detailed finite element simulations were carried out to study the response and failure of single lap joints with non-flat interface under uniaxial tension. The adherents were made from fiber reinforced epoxy composite and the custom-made mold allowed the fibers to follow the profile of the bonded joint interface. The experiments showed that the interface shape has significant effect on the mechanical behavior and strength of the bonded joints. Finite element simulations were performed to estimate the distribution of shear and peeling stresses along the bonded joints and the results were linked to the experimental investigations. Additional parametric calculations were also carried out to highlight the role of interface shape on the distribution of stresses, and inherently the overall strength and behavior of the bonded joints. In addition, the role of a central void on the distribution of the stresses in a bonded joint with flat and non-flat sinusoidal interfaces was investigated. The second topic concerns Wood Plastic Composites (WPC) which are widely used in the industry due to its durability, low cost, and anti-moisture properties in comparison with the natural wood. In this research, we have produced flout shaped WPC samples using African black wood powder and Phenolic resin in a hot compression molding set-up. Initial WPC composites were produced by systematically changing the wood volume fraction. Based on these results the optimum temperature, pressure and wood volume fraction for developing WPC in a form of a flute is developed. A series of experimental procedures were performed to improve mechanical properties of WPC samples by

  15. 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

  16. 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.)

  17. 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

  18. Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

    Science.gov (United States)

    Stork, D.; Agostini, P.; Boutard, J. L.; Buckthorpe, D.; Diegele, E.; Dudarev, S. L.; English, C.; Federici, G.; Gilbert, M. R.; Gonzalez, S.; Ibarra, A.; Linsmeier, Ch.; Li Puma, A.; Marbach, G.; Morris, P. F.; Packer, L. W.; Raj, B.; Rieth, M.; Tran, M. Q.; Ward, D. J.; Zinkle, S. J.

    2014-12-01

    The findings of the EU 'Materials Assessment Group' (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision. A DEMO phase I with a 'Starter Blanket' and 'Starter Divertor' is foreseen: the blanket being capable of withstanding ⩾2 MW yr m-2 fusion neutron fluence (∼20 dpa in the front-wall steel). A second phase ensues for DEMO with ⩾5 MW yr m-2 first wall neutron fluence. Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction with functional materials/coolants; and a project-based risk analysis, with R&D to mitigate risks from material shortcomings including development of specific risk mitigation materials. The DEMO balance of plant constrains the blanket and divertor coolants to remain unchanged between the two phases. The blanket coolant choices (He gas or pressurised water) put technical constraints on the blanket steels, either to have high strength at higher temperatures than current baseline variants (above 650 °C for high thermodynamic efficiency from He-gas coolant), or superior radiation-embrittlement properties at lower temperatures (∼290-320 °C), for construction of water-cooled blankets. Risk mitigation proposed would develop these options in parallel, and computational and modelling techniques to shorten the cycle-time of new steel development will be important to achieve tight R&D timescales. The superior power handling of a water-cooled divertor target suggests a substructure temperature operating window (∼200-350 °C) that could be realised, as a baseline-concept, using tungsten on a copper

  19. Multiplex strengthening by aging and cold working of fusion structure material V-4Cr-4Ti

    International Nuclear Information System (INIS)

    V-4Cr-4Ti alloy is a kind of fusion blanket structure material. For strengthening, multiplex strengthening research has been developed by several combination of aging and cold working. Tensile test at room temperature and high temperature was used to evaluate the alloy's mechanical property with different combination of aging and cold working. The results show that, by comparison with traditional solid annealing and aging (SAA), cold working after solid annealing and aging (SAACW) greatly increased the yield and ultimate strength whether at room temperature or high temperature, and the elongation is also usable; However, while cold working was arranged between solid annealing and aging (SACWA), though the strength is a little declined compared to SAACW, elongation improved. This ensures the good combination of strength and plasticity; repeated cold working and aging (namely SACWACWA) strengthened the alloy significantly, which is in the similar level as SAACW, and the ductility is almost the same with SACWA. As stated above, SACWA and SACWACWA all have significant strengthening effect on the V-4Cr-4Ti alloy, and SACWACWA has a higher degree. This is possibly due to the dislocations created in the cold working are pinned by precipitations brought in the succeeding aging process. And by repeated cold working and aging, the reaction between dislocation and precipitation is much more complex. Fracture surface of all above-mentioned conditions shows dimple, that is to say, all the materials are ductile. (authors)

  20. 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.

  1. Methods of using structures including catalytic materials disposed within porous zeolite materials to synthesize hydrocarbons

    Science.gov (United States)

    Rollins, Harry W.; Petkovic, Lucia M.; Ginosar, Daniel M.

    2011-02-01

    Catalytic structures include a catalytic material disposed within a zeolite material. The catalytic material may be capable of catalyzing a formation of methanol from carbon monoxide and/or carbon dioxide, and the zeolite material may be capable of catalyzing a formation of hydrocarbon molecules from methanol. The catalytic material may include copper and zinc oxide. The zeolite material may include a first plurality of pores substantially defined by a crystal structure of the zeolite material and a second plurality of pores dispersed throughout the zeolite material. Systems for synthesizing hydrocarbon molecules also include catalytic structures. Methods for synthesizing hydrocarbon molecules include contacting hydrogen and at least one of carbon monoxide and carbon dioxide with such catalytic structures. Catalytic structures are fabricated by forming a zeolite material at least partially around a template structure, removing the template structure, and introducing a catalytic material into the zeolite material.

  2. 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

  3. The gas-cooled Li2O moderator/breeder canister blanket for fusion-synfuels

    International Nuclear Information System (INIS)

    A new integrated power and breeding blanket is described. The blanket incorporates features that make it suitable for synthetic fuel production. It is matched to the thermal and electrical requirements of the General Atomic water-splitting process for producing hydrogen. The fusion reaction is the Tandem Mirror Reactor (TMR) using Mirror Advanced Reactor Study (MARS) physics. The canister blanket is a high temperature, pressure balanced, crossflow heat exchanger contained within a low activity, independently cooled, moderate temperature, first wall structural envelope. The canister uses Li2O as the moderator/breeder and helium as the coolant. ''In situ'' tritium control, combined with slip stream processing and self-healing permeation barriers, assures a hydrogen product essentially free of tritium. The blanket is particularly adapted to synfuels production but is equally useful for electricity production or co-generation

  4. 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.)

  5. 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

  6. Power reactors and sub-critical blanket systems with lead and lead-bismuth as coolant and/or target material. Utilization and transmutation of actinides and long lived fission products

    International Nuclear Information System (INIS)

    collaborative R and D programmes are becoming of increasing importance. It is with this focus that the IAEA convened the Advisory Group Meeting on Design and Performance of Reactor and Subcritical Blanket Systems with Lead and Lead-Bismuth as Coolant and/or Target Material, in co-operation with the Research and Development Institute of Power Engineering (RDIPE). the Institute of Physics and Power Engineering (IPPE) and Minatom in the Russian Federation. This meeting, which assembled sixteen participants from eight countries, drew upon the vast experience of a group of international experts in order to review and discuss the recent R and D developments in critical and sub-critical concepts, coolant properties, and experimental and analytical validation work, as well as to exchange information on the experience accumulated. and to discuss the issues requiring further R and D. A total of twenty-four presentations and/or statements were made by the participants

  7. Atmospheric corrosion of structural materials in Poland

    International Nuclear Information System (INIS)

    The paper presents investigations of the effects of aggressive atmospheric corrosion agents on the progress of corrosion of metal structure materials: carbon steel, zinc, copper and aluminium. The tests were performed at the corrosion stations according standards: ISO 9223:1992; ISO 9224:1992; ISO 9225:1992; ISO 9226:1992; ISO 8565:1992. As the result of 8 years of experience were determined main atmospheric factors influencing corrosion rates of the several standard metals. It was shown that except factors discussed in the standard ISO 9223:1992 (sulfur dioxide, time of wetness and chloride ion deposition) ozone play a main part in corrosion rates. Ozone accelerates particularly the corrosion of copper, however steel and aluminium too. (author)

  8. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Zhongliang; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Chen, Chong; Li, Min; Zhou, Guangming

    2015-06-15

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements.

  9. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements

  10. Simplification of blanket system for SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    Simplification of blanket system is necessary for a fusion DEMO reactor. Although the conceptual design of a tritium-breeding blanket for SlimCS has been studied in the past several years, the structure of the previous blanket seems to be complex and difficult to manufacture from the viewpoint of engineering. In this paper, we proposed simplification of blanket structure without decreasing the net Tritium Breeding Ratio (TBR). In the proposed concept, the blanket is filled with the mixture of Li4SiO4 pebbles or Li2O pebbles for the tritium breeding and Be12Ti pebbles for the neutron multiplication. To confirm the effectiveness of this concept, an ANIHEAT code with the nuclear library FENDL-2.0 was used for calculations of the neutronic and thermal analyses. The result indicated that, under the constraint of the blanket thickness being less than 0.5 m, the mixture of Li2O pebbles and Be12Ti ones is the most effective and that the TBR is expected to be greater than 1.05.

  11. Towards a reduced activation structural materials database for fusion DEMO reactors

    International Nuclear Information System (INIS)

    The development of First Wall, Blanket and Divertor materials which are capable of withstanding many years the high neutron and heat fluxes, is a critical path to fusion power. Therefore, the timely availability of a sound materials database has become an indispensable element in international fusion road maps. In order to provide materials design data for short term needs of ITER Test Blanket Modules and for a DEMOnstration fusion reactor, a wealth of R and D results on the European reduced activation ferritic-martensitic steel EUROFER, and on oxide dispersion strengthened variants are being characterized, mainly in the temperature window 250-650 deg. C. The characterisation includes irradiations up to 15 dpa in the mixed spectrum reactor HFR and up to 75 dpa in the fast breeder reactor BOR60. Industrial EUROFER-batches of 3.5 and 7.5 tons have been produced with a variety of semi-finished, quality-assured product forms. To increase thermal efficiency of blankets, high temperature resistant SiCf/SiC channel inserts for liquid metal coolant tubes are also developed. Regarding radiation damage resistance, a broad based reactor irradiation programs counts several steps from ≤5dpa (ITER TBMs) up to 75 dpa (DEMO). For the European divertor designers, a materials data base is presently being set up for pure W and W alloys, and related reactor irradiations are foreseen with temperatures from 650-1000 deg. C. (author)

  12. 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

  13. Perspective: Role of structure prediction in materials discovery and design

    OpenAIRE

    Richard J. Needs; Pickard, Chris J.

    2016-01-01

    Materials informatics owes much to bioinformatics and the Materials Genome Initiative has been inspired by the Human Genome Project. But there is more to bioinformatics than genomes, and the same is true for materials informatics. Here we describe the rapidly expanding role of searching for structures of materials using first-principles electronic-structure methods. Structure searching has played an important part in unraveling structures of dense hydrogen and in identifying the record-high-t...

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

    Full text: One of the main requirements of advanced nuclear-power engineering is inherent safety of power installations. It initiates R and D of heavy liquid metals (lead, lead- bismuth eutectic) application in fission reactors as substitute of sodium. The same requirement makes advisable R and D of the lead and lead-bismuth eutectic application in blanket of fusion reactors as substitute of lithium. High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease MHD-resistance authors propose to form electro-insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the electro-insulating coatings characteristics rd (r - specific resistance of coatings, d - thickness) is ∼ 10-5Ω·m2 for steels and 5, 0x10-6 - 5, 0x10-5Ω·m2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there electro-insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steam generators and another equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. Mechanical and materials engineering of modern structure and component design

    CERN Document Server

    Altenbach, Holm

    2015-01-01

    This book presents the latest findings on mechanical and materials engineering as applied to the design of modern engineering materials and components. The contributions cover the classical fields of mechanical, civil and materials engineering, as well as bioengineering and advanced materials processing and optimization. The materials and structures discussed can be categorized into modern steels, aluminium and titanium alloys, polymers/composite materials, biological and natural materials, material hybrids and modern nano-based materials. Analytical modelling, numerical simulation, state-of-the-art design tools and advanced experimental techniques are applied to characterize the materials’ performance and to design and optimize structures in different fields of engineering applications.

  5. A comparative study of the performance and economics of advanced and conventional structural materials in fusion systems

    International Nuclear Information System (INIS)

    The impact of the neutron wall load as well as the lifetime and operating temperature of the structural material on tokamak reactor economics was investigated and a comparative study of stainless steel and vanadium alloys was performed. In order to limit the fractional increase in the cost of energy due to the plant downtime, t, for replacement of the structural material to delta, the structure lifetime, T, must be greater than t/delta where T and t are in years. Economically attractive tokamak reactors produce a neutron wall load of 3-4 MW/m2 for 3000 MW thermal power. The cost of energy is optimized by an operating temperature of the structural material in the wall/blanket in the range 475-5000C for stainless steel and 620-6600C for vanadium alloys. The gain in electric power due to higher operating temperatures is not sufficient to offset the penalty in the capital cost associated with the use of vanadium alloys as compared to stainless steel. Therefore, the vanadium alloy must exhibit a significant lifetime advantage over stainless steel to be economically competitive. The magnitude of this advantage is particularly sensitive to the plant downtime and the reference lifetime of stainless steel as well as the extent to which the refractory alloy has to be used in the heat transport system. (orig.)

  6. 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.)

  7. Design criteria and mitigation options for thermal fatigue effects in ATW blankets

    International Nuclear Information System (INIS)

    Thermal fatigue due to beam interruptions is an issue that must be addressed in the design of an ATW blanket. Two different approaches can be taken to address this issue. One approach is to analyze current ATW blanket designs in order to set interrupt frequency design limits for the accelerator. The other approach is to assume that accelerator reliability can not be guaranteed before design and construction of the blanket. In this approach the blanket must be designed so as to accommodate an accelerator with a beam interruption frequency significantly higher than current high power accelerators in order to provide a margin of error. Both approaches are considered in this paper. Both a sodium cooled blanket design and a lead-bismuth cooled blanket design are considered. Thermal hydraulic analysis of the blanket for beam interruption transients is carried out with the SASSYS-1 systems analysis code to obtain the time histories of the coolant temperatures in contact with structural components. These coolant temperatures are then used in a detailed structure temperature calculation to obtain structure surface and structure average temperatures. The difference between the average temperature and the surface temperature is used to obtain thermal strains. Low cycle fatigue curves from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code are used to determine the number of cycles that the structural components can endure, based on these strains. Calculations are made for base case designs and for a number of mitigation options. The mitigation options include using two separate accelerators to provide the beam, reducing the thickness of the above core load pads in the subassemblies, increasing the coolant flow rate or reducing power in order to reduce the core temperature rise, and reducing the superheat in the once-through steam generator. (author)

  8. A New Light Weight Structural Material for Nuclear Structures

    Energy Technology Data Exchange (ETDEWEB)

    Rabiei, Afsaneh [North Carolina State Univ., Raleigh, NC (United States)

    2016-01-14

    Radiation shielding materials are commonly used in nuclear facilities to attenuate the background ionization radiations to a minimum level for creating a safer workplace, meeting regulatory requirements and maintaining high quality performance. The conventional radiation shielding materials have a number of drawbacks: heavy concrete contains a high amount of elements that are not desirable for an effective shielding such as oxygen, silicon, and calcium; a well known limitation of lead is its low machinability and toxicity, which is causing a major environmental concern. Therefore, an effective and environmentally friendly shielding material with increased attenuation and low mass density is desirable. Close-cell composite metal foams (CMFs) and open-cell Al foam with fillers are light-weight candidate materials that we have studied in this project. Close-cell CMFs possess several suitable properties that are unattainable by conventional radiation shielding materials such as low density and high strength for structural applications, high surface area to volume ratio for excellent thermal isolation with an extraordinary energy absorption capability. Open-cell foam is made up of a network of interconnected solid struts, which allows gas or fluid media to pass through it. This unique structure provided a further motive to investigate its application as radiation shields by infiltrating original empty pores with high hydrogen or boron compounds, which are well known for their excellent neutron shielding capability. The resulting open-cell foam with fillers will not only exhibit light weight and high specific surface area, but also possess excellent radiation shielding capability and good processability. In this study, all the foams were investigated for their radiation shielding efficiency in terms of X-ray, gamma ray and neutron. X-ray transmission measurements were carried out on a high-resolution microcomputed tomography (microCT) system. Gamma-emitting sources: 3.0m

  9. A New Light Weight Structural Material for Nuclear Structures

    International Nuclear Information System (INIS)

    Radiation shielding materials are commonly used in nuclear facilities to attenuate the background ionization radiations to a minimum level for creating a safer workplace, meeting regulatory requirements and maintaining high quality performance. The conventional radiation shielding materials have a number of drawbacks: heavy concrete contains a high amount of elements that are not desirable for an effective shielding such as oxygen, silicon, and calcium; a well known limitation of lead is its low machinability and toxicity, which is causing a major environmental concern. Therefore, an effective and environmentally friendly shielding material with increased attenuation and low mass density is desirable. Close-cell composite metal foams (CMFs) and open-cell Al foam with fillers are light-weight candidate materials that we have studied in this project. Close-cell CMFs possess several suitable properties that are unattainable by conventional radiation shielding materials such as low density and high strength for structural applications, high surface area to volume ratio for excellent thermal isolation with an extraordinary energy absorption capability. Open-cell foam is made up of a network of interconnected solid struts, which allows gas or fluid media to pass through it. This unique structure provided a further motive to investigate its application as radiation shields by infiltrating original empty pores with high hydrogen or boron compounds, which are well known for their excellent neutron shielding capability. The resulting open-cell foam with fillers will not only exhibit light weight and high specific surface area, but also possess excellent radiation shielding capability and good processability. In this study, all the foams were investigated for their radiation shielding efficiency in terms of X-ray, gamma ray and neutron. X-ray transmission measurements were carried out on a high-resolution microcomputed tomography (microCT) system. Gamma-emitting sources: 3.0m

  10. 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

  11. Preliminary design of a helium cooled molten lithium test blanket module for the ITER test in Korea

    International Nuclear Information System (INIS)

    Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) blanket with ferritic steel (FS) as a structural material in the International Thermonuclear Experimental Reactor (ITER) program. The preliminary design and its performance of KO HCML test blanket module (TBM) are introduced in this paper. It uses He as a coolant at an inlet temperature of 300 deg. C and an outlet temperature up to 400 deg. C and Li is used as a tritium breeder by considering its potential advantages. Two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. A 3-D Monte Carlo analysis is performed with the MCCARD code for the neutronics and the total TBM power is designed to be 0.739 MW at a normal heat flux from the plasma side. From the analysis results with CFX-10 for the thermal-hydraulics, the He cooling path is determined and it shows that the maximum temperature of the first wall does not exceed 550 deg. C at the structural materials and the coolant velocities are 45 and 11.5 m/s in the first wall and breeding zone, respectively. The obtained temperature data is used in the thermal-mechanical analysis with ANSYS-10. The maximum von Mises equivalent stress of the first wall is 123 MPa and the maximum deformation of it is 3.73 mm, which is lower than the maximum allowable stress

  12. Status of LWR primary pressure boundary structural materials

    International Nuclear Information System (INIS)

    The integrity of major systems, structures and components is a prerequisite to the economy and safety of an existing light water reactor and also for the next generation reactors. As few reactor structural materials are being manufactured by domestic companies, based on economic and safety reasons, a new demand to improve the quality of domestic reactor structural materials and to develop reactor structural steels has arisen. Investigations on the state-of-the-art of the materials specifications, performance and current state of structural materials development were performed as a first step to domestic reactor structural steel development and summarized the result in the present report. (Author) 10 refs., 10 figs., 21 tabs

  13. Reversible Hydrogen Storage MaterialsStructure, Chemistry, and Electronic Structure

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, Ian M. [University of Wisconsin-Madison; Johnson, Duane D. [Ames Lab., Iowa

    2014-06-21

    To understand the processes involved in the uptake and release of hydrogen from candidate light-weight metal hydride storage systems, a combination of materials characterization techniques and first principle calculation methods have been employed. In addition to conventional microstructural characterization in the transmission electron microscope, which provides projected information about the through thickness microstructure, electron tomography methods were employed to determine the three-dimensional spatial distribution of catalyst species for select systems both before and after dehydrogenation. Catalyst species identification as well as compositional analysis of the storage material before and after hydrogen charging and discharging was performed using a combination of energy dispersive spectroscopy, EDS, and electron energy loss spectroscopy, EELS. The characterization effort was coupled with first-principles, electronic-structure and thermodynamic techniques to predict and assess meta-stable and stable phases, reaction pathways, and thermodynamic and kinetic barriers. Systems studied included:NaAlH4, CaH2/CaB6 and Ca(BH4)2, MgH2/MgB2, Ni-Catalyzed Magnesium Hydride, TiH2-Catalyzed Magnesium Hydride, LiBH4, Aluminum-based systems and Aluminum

  14. Dynamic response of structures constructed from smart materials

    OpenAIRE

    Caughey, T. K.

    1995-01-01

    The dynamic analysis of structures constructed of homogeneous smart materials is greatly simplified by the observation that the eigenfunctions of such structures are identical to those of the same structures constructed entirely of purely elastic materials. The dynamic analysis of such structures is thus reduced to the analysis of the temporal behaviour of the eigenmodes of the structure. The theory is illustrated for both continuous and discrete structures using the generalization of 'positi...

  15. 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

  16. Controlled structuration and functionalization of organic-inorganic hybrid materials

    International Nuclear Information System (INIS)

    Organic-inorganic hybrid materials obtained by sol-gel process are very interesting materials. Indeed, these materials have both the properties of the organic phase and those of the inorganic matrix. Two functionalization and structuration ways are presented here: 1)the meso-porous hybrid materials formed in presence of structuring surfactant and the lamellar hybrid materials obtained by auto-assembling. (O.M.)

  17. Neutronics and thermal design analyses of US solid breeder blanket for ITER

    International Nuclear Information System (INIS)

    The US Solid Breeder Blanket is designed to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Safety, low tritium inventory, reliability, flexibility cost, and minimum R ampersand D requirements are the other design criteria. To satisfy these criteria, the produced tritium is recovered continuously during operation and the blanket coolant operates at low pressure. Beryllium multiplier material is used to control the solid-breeder temperature. Neutronics and thermal design analyses were performed in an integrated manner to define the blanket configuration. The reference parameters of ITER including the operating scenarios, the neutron wall loading distribution and the copper stabilizer are included in the design analyses. Several analyses were performed to study the impact of the reactor parameters, blanket dimensions, material characteristics, and heat transfer coefficient at the material interfaces on the blanket performance. The design analyses and the results from the different studies are summarized. 6 refs., 3 figs., 3 tabs

  18. Two dimensional distribution of tritium breeding ratio and induced activity in Japanese water cooled and helium cooled test blanket modules

    International Nuclear Information System (INIS)

    not take into account the common frame. In the verification of the tritium breeding ability of TBMs, this effect should be accounted. The result of induced activation analysis showed that induced activation level of the TBM is one order of magnitude lower than the fusion power plant. However, the activation level of the test blanket module seems to be two orders of magnitude higher than the acceptable activation level for recycle of the activated material. Analysis results obtained in this work will be used for the detailed estimation of tritium production performance and thermo-mechanical performance of TBM structures. The results of induced activation analysis will be used for the planning of the post-irradiation examination of TBM's. (author)

  19. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

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

  20. 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.

  1. 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

  2. A granulation model using diosgenin wastewater in an upflow anaerobic sludge blanket reactor

    Institute of Scientific and Technical Information of China (English)

    Jianguo BAO; Hui LIU; Yanxin WANG; Lijun ZHANG

    2009-01-01

    An enhanced start-up of an upfiow anaerobic sludge blanket (UASB) reactor for diosgenin wastewater treatment was designed and experimentally tested. Gran-ular sludge was formed on day 35 in the reactor with high concentrations of chloride (4000-7000 mg/L) and COD (5000-13000mg/L) as substrate. A new model for the granulation was proposed which divides the formation of anaerobic granules into six consecutive stages; they include semi-embryonic granule formation, embryonic granule formation, single-nucleus granule formation, multi-nuclei granule formation, granule growth and granule maturation. A model of the granule structure was also proposed based on scanning electron microscope observation. The microspores occurring on the surface and further leading into the interior of the granules were considered as the channels and the passage of the materials and the products of the microorganisms' metabolism inside the granules.

  3. 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.)

  4. Properties of structural materials in liquid metal environment

    International Nuclear Information System (INIS)

    The proceedings contain 16 contributions to the following topics: 1. Creep-Rupture Behaviour of Structural Materials in Liquid Metal Environment; 2. Behaviour of Materials in Liquid Metal Environment under Off-Normal Conditions; 3. Fatigue and Creep-Fatigue of Structural Materials in Liquid Metal Environment; and 4. Crack Propagation in Liquid Sodium. (MM)

  5. Neutronic Analysis of Flux Dispersion in a Multi-Layered, (D-T) Driven Hybrid Blanket

    OpenAIRE

    İPEK, Osman

    2000-01-01

    The concept of `hybrid blanket' is based on the placement of the nuclear fuel layer, which is a fertile material and fissionable by the fusion neutrons, at the front or the rear sides of the tritium breeding zone so that, in addition to gaining fission energy, a fissile fuel is produced. The neutronic flux distribution (neutron spectrum) along the radial direction varies with the type of material and the geometry used in the blanket, and also according to whether it is multi-layered or s...

  6. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    International Nuclear Information System (INIS)

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 ∼ -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author)

  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. Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

    Energy Technology Data Exchange (ETDEWEB)

    Stork, D., E-mail: derek.stork@btinternet.com [Euratom – CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Agostini, P. [ENEA, Brasimone Research Centre, 40032 Cumugnano, Bologna (Italy); Boutard, J.L. [CEA, cab HC, Saclay, F-91191 Gif-sur-Yvette (France); Buckthorpe, D. [AMEC, Booths Park, Chelford Road, Knutsford, Cheshire WA16 8QZ (United Kingdom); Diegele, E. [Karlsruhe Institute for Technology, IMF-I, D-7602 Karlsruhe (Germany); Dudarev, S.L. [Euratom – CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); English, C. [National Nuclear Laboratory, Chadwick House, Warrington Road, Birchwood Park WA3 6AE (United Kingdom); Federici, G. [EFDA Power Plant Physics and Technology, Boltzmannstr. 2, Garching 85748 (Germany); Gilbert, M.R. [Euratom – CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Gonzalez, S. [EFDA Power Plant Physics and Technology, Boltzmannstr. 2, Garching 85748 (Germany); Ibarra, A. [CIEMAT, Avda. Complutense 40, Madrid (Spain); Linsmeier, Ch. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, EURATOM Association, 52425 Jülich (Germany); Li Puma, A. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Marbach, G. [CEA, cab HC, Saclay, F-91191 Gif-sur-Yvette (France); Morris, P.F. [Formerly of TATA Steel Europe, Swinden Technology Centre, Moorgate, Rotherham S60 3AR (United Kingdom); Packer, L.W. [Euratom – CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Raj, B. [Indian National Academy of Engineering, Shaheed Jeet Singh Marg, New Delhi 110016 (India); Rieth, M. [Karlsruhe Institute for Technology, IMF-I, D-7602 Karlsruhe (Germany); and others

    2014-12-15

    The findings of the EU ‘Materials Assessment Group’ (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R and D up to a DEMO construction decision. A DEMO phase I with a ‘Starter Blanket’ and ‘Starter Divertor’ is foreseen: the blanket being capable of withstanding ⩾2 MW yr m{sup −2} fusion neutron fluence (∼20 dpa in the front-wall steel). A second phase ensues for DEMO with ⩾5 MW yr m{sup −2} first wall neutron fluence. Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction with functional materials/coolants; and a project-based risk analysis, with R and D to mitigate risks from material shortcomings including development of specific risk mitigation materials. The DEMO balance of plant constrains the blanket and divertor coolants to remain unchanged between the two phases. The blanket coolant choices (He gas or pressurised water) put technical constraints on the blanket steels, either to have high strength at higher temperatures than current baseline variants (above 650 °C for high thermodynamic efficiency from He-gas coolant), or superior radiation-embrittlement properties at lower temperatures (∼290–320 °C), for construction of water-cooled blankets. Risk mitigation proposed would develop these options in parallel, and computational and modelling techniques to shorten the cycle-time of new steel development will be important to achieve tight R and D timescales. The superior power handling of a water-cooled divertor target suggests a substructure temperature operating window (∼200–350 °C) that could be realised, as a

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-22

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

  10. Structured Piezoelectric Composites: Materials and Applications

    NARCIS (Netherlands)

    Van den Ende, D.A.

    2012-01-01

    The piezoelectric effect, which causes a material to generate a voltage when it deforms, is very suitable for making integrated sensors, and (micro-) generators. However, conventional piezoelectric materials are either brittle ceramics or certain polymers with a low thermal stability, which limits t

  11. Material, Structural Design of Armour Units

    DEFF Research Database (Denmark)

    Burcharth, Hans F.

    Stone and concrete are two materials generally used for the construction of rubble mound breakwaters. This paper deals with concrete only.......Stone and concrete are two materials generally used for the construction of rubble mound breakwaters. This paper deals with concrete only....

  12. Hierarchical Optimization of Material and Structure

    DEFF Research Database (Denmark)

    Rodrigues, Helder C.; Guedes, Jose M.; Bendsøe, Martin P.

    2002-01-01

    This paper describes a hierarchical computational procedure for optimizing material distribution as well as the local material properties of mechanical elements. The local properties are designed using a topology design approach, leading to single scale microstructures, which may be restricted in...

  13. Perspective: Role of structure prediction in materials discovery and design

    Science.gov (United States)

    Needs, Richard J.; Pickard, Chris J.

    2016-05-01

    Materials informatics owes much to bioinformatics and the Materials Genome Initiative has been inspired by the Human Genome Project. But there is more to bioinformatics than genomes, and the same is true for materials informatics. Here we describe the rapidly expanding role of searching for structures of materials using first-principles electronic-structure methods. Structure searching has played an important part in unraveling structures of dense hydrogen and in identifying the record-high-temperature superconducting component in hydrogen sulfide at high pressures. We suggest that first-principles structure searching has already demonstrated its ability to determine structures of a wide range of materials and that it will play a central and increasing part in materials discovery and design.

  14. Structured materials for catalytic and sensing applications

    Science.gov (United States)

    Hokenek, Selma

    The optical and chemical properties of the materials used in catalytic and sensing applications directly determine the characteristics of the resultant catalyst or sensor. It is well known that a catalyst needs to have high activity, selectivity, and stability to be viable in an industrial setting. The hydrogenation activity of palladium catalysts is known to be excellent, but the industrial applications are limited by the cost of obtaining catalyst in amounts large enough to make their use economical. As a result, alloying palladium with a cheaper, more widely available metal while maintaining the high catalytic activity seen in monometallic catalysts is, therefore, an attractive option. Similarly, the optical properties of nanoscale materials used for sensing must be attuned to their application. By adjusting the shape and composition of nanoparticles used in such applications, very fine changes can be made to the frequency of light that they absorb most efficiently. The design, synthesis, and characterization of (i) size controlled monometallic palladium nanoparticles for catalytic applications, (ii) nickel-palladium bimetallic nanoparticles and (iii) silver-palladium nanoparticles with applications in drug detection and biosensing through surface plasmon resonance, respectively, will be discussed. The composition, size, and shape of the nanoparticles formed were controlled through the use of wet chemistry techniques. After synthesis, the nanoparticles were analyzed using physical and chemical characterization techniques such as X-Ray Diffraction (XRD), Transmission Electron Microscopy (TEM), and Scanning Transmission Electron Microscopy- Energy-Dispersive Spectrometry (STEM-EDX). The Pd and Ni-Pd nanoparticles were then supported on silica for catalytic testing using mass spectrometry. The optical properties of the Ag-Pd nanoparticles in suspension were further investigated using ultraviolet-visible spectrometry (UV-Vis). Monometallic palladium particles have

  15. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

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

  16. Innovated Building Material's Interactions with Structural Form in Architectural Projects

    OpenAIRE

    Mozaikci, Begüm

    2009-01-01

    ABSTRACT: Interpretation of building materials into architectural form, is gaining greater interest and attention due to the conservation of architectural heritage. This study highlight, the influences of technological developments of building materials and their interactions with structural form in architectural design projects. Architectural form and decisions can potentially effects by introduced new materials and this study focus on the interaction of new material and structural form...

  17. Multi-Material Design Optimization of Composite Structures

    OpenAIRE

    Hvejsel, Christian Frier

    2011-01-01

    This PhD thesis entitled “Multi-Material Design Optimization of Composite Structures” addresses the design problem of choosing materials in an optimal manner under a resource constraint so as to maximize the integral stiffness of a structure under static loading conditions. In particular stiffness design of laminated composite structures is studied including the problem of orienting orthotropic material optimally. The approach taken in this work is to consider this multi-material design probl...

  18. Optimization of Structure and Material Properties for Solids Composed of Softening Material

    DEFF Research Database (Denmark)

    Bendsøe, Martin P.; Guedes, J.M.; J.M., Plaxton; Taylor, J.E.

    1996-01-01

    Recent results on the design of material properties in the context of global structural optimization provide, in analytical form, a prediction of the optimal material tensor distributions for two or three dimensional continuum structures. The model developed for that purpose is extended here to...... cover the design of a structure and associated material properties for a system composed of a generic form of nonlinear softening material. As was established in the earlier study on design with linear materials, the formulation for combined 'material and structure' design with softening materials can...... be expressed as a convex problem. However, the optimal distribution of material properties predicted in the nonlinear problem depends on the magnitude of load, in contrast to the case with linear material. Computational solutions are presented for several example problems, showing how the optimal...

  19. Neutronic evaluation of fissile fuel breeding blankets for the fission-suppressed Tandem-Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    A computational study was performed on the blanket design of the Lawrence Livermore National Laboratory (LLNL) fission-suppressed Tandem Mirror Hybrid Reactor (TMHR) to qualify the methods and data bases available at Oak Ridge National Laboratory (ORNL) for use in analyzing the neutronic performance of fissile fuel breeding blankets. The eventual goal of the study was to establish the capability for analysis and optimization of advanced fissile fuel production blanket designs. Discrete ordinates radiation transport calculations were performed in one-dimensional cylindrical geometry to obtain the blanket spatial distribution and energy spectra of the neutron and gamma-ray fluxes resulting from the monoenergetic (14.1 MeV) fusion first wall source. Key macroscopic cross sections of the blanket materials were then folded with the flux spectra to obtain reaction rates critical to evaluating blanket feasibility. Finally, a time-dependent depletion analysis was performed to evaluate the blanket performance during equilibrium cycle conditions. The results of the study are presented both as graphs and tables

  20. A Thermal/Hydraulic Safety Assessment of the Blanket Conceptual Design for the Accelerator Production of Tritium Facility

    International Nuclear Information System (INIS)

    In support of the Accelerator Production of Tritium (APT) project, safety analyses for the blanket system have been performed based on the conceptual design for the Target/Blanket (T/B) Facility. During mitigated event sequences safety engineered features, such as the residual heat removal (RHR) and cavity flood systems, provide sufficient protection for maintaining the structural integrity of the blanket system and its components. During unmitigated (with beam shutdown only) event sequences, passive features such as natural circulation, thermal inertia, and boil-off provide significant time for corrective measures to be taken

  1. Physical investigation for neutron consumption and multiplication in fusion–fission hybrid test blanket module

    International Nuclear Information System (INIS)

    Highlights: • Preliminary design of hybrid test blanket module was done for ITER. • Blanket performance was analyzed for five fuel types with Li and LiPb coolants. • Detailed neutronic analysis was performed by computing neutron production and loss factors. • Inelastic neutron source factor of LiPb caused major change in blanket performance. • TiC reflector improved performance in TRU transmutation and tritium breeding. - Abstract: Inelastic scattering of high energy fusion neutrons does affect the performance of fusion blanket based on the choice of different materials. It will also affect the behavior of source neutrons in a subcritical fusion fission hybrid blanket and consequently the transmutation and tritium breeding performance. A fusion fission hybrid test blanket module (HTBM) is designed which is presumed to be tested in a large sized tokamak and plasma neutron source is similar to ITER. In this preliminary design of HTBM the neutron source and loss factors are computed for the detailed neutronic performance analysis. The neutronic analysis of hybrid blanket module is performed for five different TRU fuel types: TRU-Zr, TRU-Mo, TRU-Oxide, TRU-Carbide and TRU-Nitride. In this module design, it is aimed to burn and transmute the TRU nuclides from high-level radioactive waste of PWR spent fuel. The effect of TiC reflector on transmutation and tritium breeding performance of HTBM is also quantified. MCNPX is used for neutronic computations. Neutron spectrum, capture to fission ratio and waste transmutation ratio of each fuel type are compared to evaluate their waste transmutation performance. Tritium breeding ratio is also compared for two coolant options: Li and LiPb eutectic

  2. Optimisation of hot isostatic pressing bonded SS/SS joints conditions for ITER blanket shield

    International Nuclear Information System (INIS)

    In the engineering design activity of international thermonuclear experimental reactor (ITER), stainless steels are being considered as candidates materials for several module type structures. Hot isostatic pressing (HIP) technique is expected for the fabrication of these modules. Stainless steel powders are simultaneously consolidated as mono-material block or/and joined in bi-material module. This paper reviews the manufacturing stages, non-destructive examination and the developments of the HIP bonded joints of 316L SS (powder and solid) for application to the ITER shield blanket. It is well known that the powder surface oxidation negatively influences the impact toughness of raw material and joints consolidated by this way. In order to get acceptable mechanical properties of materials, a study on the effect of reducing the powder oxygen content has been launched. To evaluate susceptibility to the oxygen content of HIPed joint specimens, tensile and toughness tests have been performed. From this study, optimal conditions of HIP were fitted and the influence of oxygen was mastered to obtain good mechanical properties of the consolidated powder material as well as for HIPed junction.

  3. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

  4. Computational Design of Ageless Structural Materials Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Crack initiation and propagation is a dominant failure mode for many materials and applications – usually managed via damage tolerance approaches." ...

  5. Damage evolution during fatigue in structural materials

    Czech Academy of Sciences Publication Activity Database

    Polák, Jaroslav; Man, Jiří; Petrenec, Martin

    2012-01-01

    Roč. 1, August (2012), s. 3-12. ISSN 2211-8128. [International Congress on Metallurgy and Materials - SAM/CONAMET 2011 /11./. Rosario, 18.10.2011-21.10.2011] R&D Projects: GA ČR GA106/09/1954 Institutional support: RVO:68081723 Keywords : cyclic plasticity * crack nucleation * crack growth * fatigue damage Subject RIV: JL - Materials Fatigue, Friction Mechanics

  6. 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

  7. Lightweight IMM PV Flexible Blanket Assembly

    Science.gov (United States)

    Spence, Brian

    2015-01-01

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

  8. Structure of random porous materials: Silica aerogel

    International Nuclear Information System (INIS)

    Using small-angle x-ray scattering, we show that porous silica aerogel has a fractal backbone structure. The observed structure is traced to the underlying chemical (polymerization) and physical (colloid aggregation) growth processes. Comparison of scattering curves for aerogel with silica aggregates confirms this interpretation

  9. Optical Spectroscopy of Nano Materials and Structures

    Science.gov (United States)

    Guo, Wenhao

    In this thesis, nanostructures and nanomaterials ranging from 3D to OD will be studied compresively, by using optical methods. Firstly, for 3D and 2D nanomaterials, nanoporous zeolite crystals, such as AFI and AEL are introduced as host materials to accommodate diatomic iodine molecules. Polarized Raman spectroscopy is utilized to identify the two configurations of iodine molecules to stay in the channels of AEL: the lying mode (the bond of the two atoms is parallel to the direction of the channels) and the standing mode (the bond is perpendicular to the direction of the channels). The lying mode and standing mode are switchable and can be well controlled by the amount of water molecules inside the crystal, revealed by both molecule dynamics simulation and experiment observation. With more water molecules inside, iodine molecules choose to stay in the standing mode, while with less water molecules, iodine molecules prefer to lie along the channel. Therefore, the configurations of molecules could be precisely controlled, globally by the surrounding pressure and temperature, and locally by the laser light. Ii is believed that this easy and reversible control of single molecule will be valuable in nanostructured devices, such as molecular sieving or molecular detection. Secondly, for 1D case, the PL spectrum of ZnO nanowire under uniaxial strain is studied. When a ZnO nanowire is bent, besides the lattice constant induced bandgap change on the tensile and compressive sides, there is a piezoelectric field generated along the cross section. This piezoelectric potential, together with the bandgap changes induced by the deformation, will redistribute the electrons excited by incident photons from valence band to conduction band. As a result, the electrons occupying the states at the tensile side will largely outnumbered the ones at the compressive side. Therefore, the PL spectrum we collected at the whole cross section will manifest a redshift, other than the peak

  10. Challenges in structural materials for thermal reactors

    International Nuclear Information System (INIS)

    The safe, reliable and economic operation of nuclear power plants is critically dependent on good performance of materials. Material selection is one of the most important steps in the design of the engineering components. Functional requirements and operating conditions such as stress, temperature and environment dictate the choice of materials and their properties. Criteria for selection of materials for non-nuclear components include mechanical properties (strength-ductility-toughness), corrosion properties (resistance to uniform and localized corrosion), fabricability (forming-welding-heat treatment) and cost. Additional considerations for nuclear components include neutron absorption, induced radioactivity and resistance to radiation damage. Nuclear core components are continuously subjected to bombardment by fast neutrons. Important effects of fast neutron irradiation on properties of materials are : (i) change in mechanical properties by irradiation hardening and irradiation embrittlement (ii) dimensional changes by irradiation growth, irradiation creep and void swelling and (iii) change in corrosion properties like increase in corrosion rate and introduction of new corrosion mode. There are 2 types of thermal power reactors viz., Pressure Vessel Type (BWRs and PWRs) and Pressure Tube Type (PHWRs). A very wide range of materials are used for nuclear reactor components. These include carbon and low alloy steels, stainless steels (austenitic, martensitic, precipitation hardenable), nickel-base alloys (inconels, incoloys), zirconium base alloys, etc. Three most important classes of materials are : (1) Low alloy steels for Reactor Pressure Vessel : the main concerns are increase in Ductile-Brittle transition temperature (DBTT) and decrease in toughness due to neutron irradiation. (2) Zirconium Alloys for Reactor Core Components : main concerns are Delayed Hydride cracking (DHC) and change in dimensions due to Irradiation Creep and Growth. (3) Stainless Steels

  11. Development of Steel Foam Materials and Structures

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth Kremer; Anthony Liszkiewicz; James Adkins

    2004-10-20

    In the past few years there has been a growing interest in lightweight metal foams. Demands for weight reduction, improved fuel efficiency, and increased passenger safety in automobiles now has manufacturers seriously considering the use of metal foams, in contrast to a few years ago, when the same materials would have been ruled out for technical or economical reasons. The objective of this program was to advance the development and use of steel foam materials, by demonstrating the advantages of these novel lightweight materials in selected generic applications. Progress was made in defining materials and process parameters; characterization of physical and mechanical properties; and fabrication and testing of generic steel foam-filled shapes with compositions from 2.5 wt.% to 0.7 wt.% carbon. A means of producing steel foam shapes with uniform long range porosity levels of 50 to 60 percent was demonstrated and verified with NDE methods. Steel foam integrated beams, cylinders and plates were mechanically tested and demonstrated advantages in bend stiffness, bend resistance, and crush energy absorption. Methods of joining by welding, adhesive bonding, and mechanical fastening were investigated. It is important to keep in mind that steel foam is a conventional material in an unconventional form. A substantial amount of physical and mechanical properties are presented throughout the report and in a properties database at the end of the report to support designer's in applying steel foam in unconventional ways.

  12. Course Modules on Structural Health Monitoring with Smart Materials

    Science.gov (United States)

    Shih, Hui-Ru; Walters, Wilbur L.; Zheng, Wei; Everett, Jessica

    2009-01-01

    Structural Health Monitoring (SHM) is an emerging technology that has multiple applications. SHM emerged from the wide field of smart structures, and it also encompasses disciplines such as structural dynamics, materials and structures, nondestructive testing, sensors and actuators, data acquisition, signal processing, and possibly much more. To…

  13. Lithium ceramic of blankets intend for Russian fusion reactors and an influence of the ceramic properties on parameters of reactor tritium systems

    International Nuclear Information System (INIS)

    Russian Controlled Fusion Program involves development of a DEMO design and participation in ITER Project. A solid breeder blanket in DEMO contains a ceramic orthosilicate lithium breeder and a beryllium multiplier. Test Modules of the blanket are developed in a frame of ITER activities. Experimental models of tritium breeding zones (TBZ) for the Modules, materials and technology fabrication of the TBZ, tritium reactor systems to control and treat of gases released from lithium ceramic being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ were designed, manufactured and have been tested already in IVV-2M nuclear reactor. The first model consists of lithium orthosilicate ceramic sphere pebbles (1-1.5 mm diameter) and beryllium sphere (0.1 and 1.0 mm diameter). Ceramic cylindrical pellets (11 mm diameter and 10 mm height) and porous beryllium (20% porosity) are in the second model. Some properties and microstructure of the ceramic elements are performed. Initial results of some changes of ceramic structure and in-pile experimental ratio of hydrogen and oxygen form of tritium release in helium/neon purge gas are presented. These results and outcome of irradiated LiAlO2, Li4SiO4 and Li2SiO3 ceramics in a water-graphite nuclear reactor are considered to be a DATE BASE for development of the Test Modules and the DEMO blanket and influence of the kinetic tritium release parameters on DEMO tritium systems are discussed. (author)

  14. Lightweight Materials and Structures (LMS): Inflatable Structures Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Inflatable Structures (InSTAR) project goal is to demonstrate long term durability of inflatable habitat structures for potential utilization as either in-space...

  15. Influence of nuclear data uncertainties on thorium fusion-fission hybrid blanket nucleonic performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, E.T.; Mathews, D.R.

    1979-09-01

    The fusion-fission hybrid blanket proposed for the Tandem Mirror Hybrid Reactor employs thorium metal as the fertile material. Based on the ENDF/B-IV nuclear data, the /sup 233/U and tritium production rate and blanket energy multiplication averaged over the blanket lifetime of about 9 MW-yr/m/sup 2/ are 0.76 and 1.12 per D-T neutron and 4.8, respectively. At the time of the blanket discharge, the /sup 233/U enrichment in the thorium metal is about 3%. The thorium cross sections given by the ENDF/B-IV and V were reviewed, and the important partial cross sections such as (n,2n), (n,3n), and (n,..gamma..) were found to be known to +-10 to 20% in the respective energy range of interest. A sensitivity study showed that the /sup 233/U and tritium production rate and blanket energy multiplication are relatively sensitive to the thorium capture and fission cross section uncertainties. In order to predict the above parameters within +-1%, the Th(n,..gamma..) and Th(n,..nu..f) cross sections must be measured within about +-2% in the energy range 3 to 3000 keV and 13.5 to 15 MeV, respectively.

  16. 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.

  17. Influence of nuclear data uncertainties on thorium fusion-fission hybrid blanket nucleonic performance

    International Nuclear Information System (INIS)

    The fusion-fission hybrid blanket proposed for the Tandem Mirror Hybrid Reactor employs thorium metal as the fertile material. Based on the ENDF/B-IV nuclear data, the 233U and tritium production rate and blanket energy multiplication averaged over the blanket lifetime of about 9 MW-yr/m2 are 0.76 and 1.12 per D-T neutron and 4.8, respectively. At the time of the blanket discharge, the 233U enrichment in the thorium metal is about 3%. The thorium cross sections given by the ENDF/B-IV and V were reviewed, and the important partial cross sections such as (n,2n), (n,3n), and (n,γ) were found to be known to +-10 to 20% in the respective energy range of interest. A sensitivity study showed that the 233U and tritium production rate and blanket energy multiplication are relatively sensitive to the thorium capture and fission cross section uncertainties. In order to predict the above parameters within +-1%, the Th(n,γ) and Th(n,νf) cross sections must be measured within about +-2% in the energy range 3 to 3000 keV and 13.5 to 15 MeV, respectively

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

    International Nuclear Information System (INIS)

    Neutronics experiments on two types of heterogeneous blankets have been performed as the Phase-IIC experiment of JAERI/USDOE collaborative program on fusion blanket neutronics. The experimental system was used in the same geometry as the previous Phase-IIA series which was a closed geometry using neutron source enclosure of lithium carbonate. The heterogeneous blankets selected here are the beryllium edge-on and the water coolant channel assemblies. In the former the beryllium and lithium-oxide layers are piled up alternately in the front part of test blanket. In the latter, the three simulated water cooling channels are settled in the Li2O blanket. These are producing steep gradient of neutron flux around material boundary. The calculation accuracy and measurement method for these features is a key of interest in the experiments. The measurements were performed for tritium production rate and the other nuclear parameters as well as the previous experiments. This report describes the experimental detail and the results enough to use for the benchmark data for testing the data and method of design calculation of fusion reactors. (author)

  19. Layer like porous materials with hierarchical structure

    Czech Academy of Sciences Publication Activity Database

    Roth, W. J.; Gil, B.; Makowski, W.; Marszalek, B.; Eliášová, Pavla

    -, - (2016). ISSN 0306-0012 R&D Projects: GA ČR GBP106/12/G015 Institutional support: RVO:61388955 Keywords : porous materials * physical chemistry Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 33.383, year: 2014

  20. Friction stir method for forming structures and materials

    Science.gov (United States)

    Feng, Zhili; David, Stan A.; Frederick, David Alan

    2011-11-22

    Processes for forming an enhanced material or structure are disclosed. The structure typically includes a preform that has a first common surface and a recess below the first common surface. A filler is added to the recess and seams are friction stir welded, and materials may be stir mixed.

  1. ACEE Composite Structures Technology: Review of selected NASA research on composite materials and structures

    Science.gov (United States)

    1984-01-01

    The NASA Aircraft Energy Efficiency (ACEE) Composite Primary Aircraft Structures Program was designed to develop technology for advanced composites in commercial aircraft. Research on composite materials, aircraft structures, and aircraft design is presented herein. The following parameters of composite materials were addressed: residual strength, damage tolerance, toughness, tensile strength, impact resistance, buckling, and noise transmission within composite materials structures.

  2. Band Structure Characteristics of Nacreous Composite Materials with Various Defects

    Science.gov (United States)

    Yin, J.; Zhang, S.; Zhang, H. W.; Chen, B. S.

    2016-06-01

    Nacreous composite materials have excellent mechanical properties, such as high strength, high toughness, and wide phononic band gap. In order to research band structure characteristics of nacreous composite materials with various defects, supercell models with the Brick-and-Mortar microstructure are considered. An efficient multi-level substructure algorithm is employed to discuss the band structure. Furthermore, two common systems with point and line defects and varied material parameters are discussed. In addition, band structures concerning straight and deflected crack defects are calculated by changing the shear modulus of the mortar. Finally, the sensitivity of band structures to the random material distribution is presented by considering different volume ratios of the brick. The results reveal that the first band gap of a nacreous composite material is insensitive to defects under certain conditions. It will be of great value to the design and synthesis of new nacreous composite materials for better dynamic properties.

  3. Reactor structural materials and corrosion protection method therefor

    International Nuclear Information System (INIS)

    In the inside of a pressure vessel of a BWR-type reactor, structural materials, for example, a shroud disposed at the periphery of a reactor core comprise a material such as SUS 304, stainless steel, and inconel 600. A titanium oxide semiconductor layer having a predetermined thickness is formed on the surface of the structural materials, for example, by powder plasma flame-coating. The titanium oxide semiconductor layer is determined to have an oxygen-lacking structure. A slight amount of trivalent metals is preferably doped to the titanium oxide semiconductor layer. Since radiation rays are irradiated to the titanium oxide semiconductor layer, corrosion potential on the surface of the structural materials is lowered in the vicinity of the semiconductor layer by an anode reaction. This can improve corrosion resistance of the reactor structural materials in radiation and underwater circumstance for a long period of time. (I.N.)

  4. 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.

  5. 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.)

  6. Graphene materials having randomly distributed two-dimensional structural defects

    Science.gov (United States)

    Kung, Harold H; Zhao, Xin; Hayner, Cary M; Kung, Mayfair C

    2013-10-08

    Graphene-based storage materials for high-power battery applications are provided. The storage materials are composed of vertical stacks of graphene sheets and have reduced resistance for Li ion transport. This reduced resistance is achieved by incorporating a random distribution of structural defects into the stacked graphene sheets, whereby the structural defects facilitate the diffusion of Li ions into the interior of the storage materials.

  7. Graphene materials having randomly distributed two-dimensional structural defects

    Energy Technology Data Exchange (ETDEWEB)

    Kung, Harold H.; Zhao, Xin; Hayner, Cary M.; Kung, Mayfair C.

    2016-05-31

    Graphene-based storage materials for high-power battery applications are provided. The storage materials are composed of vertical stacks of graphene sheets and have reduced resistance for Li ion transport. This reduced resistance is achieved by incorporating a random distribution of structural defects into the stacked graphene sheets, whereby the structural defects facilitate the diffusion of Li ions into the interior of the storage materials.

  8. Polymer composite material structures comprising carbon based conductive loads

    OpenAIRE

    Jérôme, Robert; Pagnoulle, Christophe; Detrembleur, Christophe; Thomassin, Jean-Michel; Huynen, Isabelle; Bailly, Christian; Bednarz, Luikasz; Daussin, Raphaël; Saib, Aimad; Baudouin, Anne-Christine; Laloyaux, Xavier

    2007-01-01

    The present invention provides a polymer composite material structure comprising at least one layer of a foamed polymer composite material comprising a foamed polymer matrix and 0.1 wt % to 6 wt % carbon based conductive loads, such as e.g. carbon nanotubes, dispersed in the foamed polymer matrix. The polymer composite material structure according to embodiments of the present invention shows good shielding and absorbing properties notwithstanding the low amount of carbon based conductive loa...

  9. Polymer composite material structures comprising carbon based conductive loads

    OpenAIRE

    Jérôme, Robert; Pagnoulle, Christophe; Detrembleur, Christophe; Thomassin, Jean-Michel; Huynen, Isabelle; Bailly, Christian; Bednarz, Lucasz; Daussin, Raphaël; Saib, Aimad

    2006-01-01

    The present invention provides a polymer composite material structure comprising at least one layer of a foamed polymer composite material comprising a foamed polymer matrix and 0.1 to 6 wt% carbon based conductive loads, such as e.g. carbon nanotubes, dispersed in the foamed polymer matrix. The polymer composite material structure according to embodiments of the present invention shows good shielding and absorbing properties notwithstanding the low amount of carbon based conductive loads. Th...

  10. Structural Description of Powder Metallurgy Prepared Materials

    Czech Academy of Sciences Publication Activity Database

    Michalcová, A.; Vojtěch, D.; Kubatík, Tomáš František; Novák, P.; Dvořák, P.

    2014-01-01

    Roč. 14, č. 3 (2014), s. 359-362. ISSN 1213-2489. [Mezinárodní konference „Mikroskopie a nedestruktivní zkoušení materiálů/3./. Litoměřice, 22.10.2014-24.10.2014] Institutional support: RVO:61389021 Keywords : SPS * Intermetallics * powder metallurgy Subject RIV: JG - Metallurgy http://journal.strojirenskatechnologie.cz/templates/obalky_casopis/XIV_2014-3.pdf

  11. Beryllium usage in fusion blankets and beryllium data needs

    International Nuclear Information System (INIS)

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

  12. 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

  13. 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

  14. 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.

  15. Smart Materials in Structural Health Monitoring, Control and Biomechanics

    CERN Document Server

    Soh, Chee-Kiong; Bhalla, Suresh

    2012-01-01

    "Smart Materials in Structural Health Monitoring, Control and Biomechanics" presents the latest developments in structural health monitoring, vibration control and biomechanics using smart materials. The book mainly focuses on piezoelectric, fibre optic and ionic polymer metal composite materials. It introduces concepts from the very basics and leads to advanced modelling (analytical/ numerical), practical aspects (including software/ hardware issues) and case studies spanning civil, mechanical and aerospace structures, including bridges, rocks and underground structures. This book is intended for practicing engineers, researchers from academic and R&D institutions and postgraduate students in the fields of smart materials and structures, structural health monitoring, vibration control and biomedical engineering. Professor Chee-Kiong Soh and Associate Professor Yaowen Yang both work at the School of Civil and Environmental Engineering, Nanyang Technological University, Singapore. Dr. Suresh Bhalla is an A...

  16. A data base for aging of structural materials

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (USNRC) initiated a Structural Aging (SAG) Program at the Oak Ridge National Laboratory (ORNL). The objective of the program is to provide assistance in identifying potential structural safety issues and to establish acceptance criteria for use in nuclear power plant evaluations for continued service. One of the main parts of the program focuses on the development of a Structural Materials Information Center where long-term and environment-dependent material properties are being collected and assembled into a data base. This data base is presented in two complementary formats. The Structural Materials Handbook is an expandable, hard-copy reference document that contains the complete data base for each material. The Structural Materials Electronic Data Base is accessible using an IBM-compatible personal computer. This paper presents an overview of the Structural Materials Information Center and briefly describes the features of the handbook and the electronic data base. In addition, a proposed method for using the data base to establish current property values for materials in existing concrete structures and to estimate the future performance of these materials is also presented. (author)

  17. A data base for aging of structural materials

    International Nuclear Information System (INIS)

    USNRC initiated a Structural Aging (SAG) Program ORNL. The objective of the program is to provide assistance in identifying potential structural safety issues and to establish acceptance criteria for use in nuclear power plant evaluations for continued service. One main part focuses on the development of a Structural Materials Information Center where long-term and environment-dependent material properties are being collected and assembled into a data base. This data base is presented in two complementary formats. The Structural Materials Handbook is an expandable, hard-copy reference document that contains the complete data base for each material. The Structural Materials Electronic Data Base is accessible using an IBM-compatible personal computer. This paper presents an overview of the Structural Materials Information Center and briefly describes the features of the handbook and the electronic data base. In addition, a proposed method for using the data base to establish current property values for materials in existing concrete structures and to estimate the future performance of these materials is also presented

  18. Design of Spintronic Materials with Simple Structures

    Energy Technology Data Exchange (ETDEWEB)

    Fong, C Y; Qian, M C; Liu, K; Yang, L H; Pask, J E

    2007-05-03

    A brief comparison of conventional electronics and spintronics is given. The key features of half metallic binary compounds with the zincblende structure are presented, using MnAs as an example. We discuss the interactions responsible for the half metallic properties. Special properties of superlattices and a digital ferromagnetic heterostructure incorporating zincblende half metals are also discussed.

  19. Design of Spintronic Materials with Simple Structures

    International Nuclear Information System (INIS)

    A brief comparison of conventional electronics and spintronics is given. The key features of half metallic binary compounds with the zincblende structure are presented, using MnAs as an example. We discuss the interactions responsible for the half metallic properties. Special properties of superlattices and a digital ferromagnetic heterostructure incorporating zincblende half metals are also discussed

  20. Structural materials for the next generation of technologies

    CERN Document Server

    Van de Voorde, Marcel Hubert

    1996-01-01

    1. Overview of advanced technologies; i.e. aerospace-aeronautics; automobile; energy technology; accelerator engineering etc. and the need for new structural materials. 2. Familiarisation with polymers, metals and alloys, structural ceramics, composites and surface engineering. The study of modern materials processing, generation of a materials data base, engineering properties includind NDE, radiation damage etc. 3. Development of new materials for the next generation of technologies; including the spin-off of materials developed for space and military purposes to industrial applications. 4. Materials selection for modern accelerator engineering. 5. Materials research in Europe, USA and Japan. Material R & D programmes sponsored by the European Union and the collaboration of CERN in EU sponsored programmes.

  1. Micro-and nano-structured conducting polymeric materials

    Institute of Scientific and Technical Information of China (English)

    LU Gewu; CHEN Feng'en; WU Xufeng; QU Liangti; ZHANG Jiaxin; SHI Gaoquan

    2005-01-01

    Conducting polymeric materials with micro-/nano-structures have potential applications in fabrication of various optical, electronic, sensing and electrochemical devices. This is mainly because these materials not only possess the characteristics of conducting polymers, but also have special functions based on their micro- or nano-structures. In this review, we summarize the recent work on "soft" and "hard" template-guided syntheses of micro-/nano-structured conducting polymers and open up the prospects of the main trends in this field.

  2. Deuterium desorption behavior of solid tritium breeding material, lithium titanate

    International Nuclear Information System (INIS)

    Several types of blanket module, solid breeder/water or helium cooling, LiPb breeder/helium cooling, liquid LiN and Flibe, have been developed toward both ITER and a demonstration reactor in Japan. In the solid breeder blanket cooled by water, pellets of Li2TiO3 will be employed as tritium breeding material. Structure material in this blanket is low activation ferritic steel, F82H. The operation temperature is limited below approximately 820 K owing to swelling caused by neutron irradiation. Tritium produced by fusion neutrons in this breeding material has to be desorbed under a blanket operation for tritium recovery to be easy. The blanket module, however, has a spatial distribution of temperature. Thus, the tritium desorption behavior has to be clarified in order to make a scheme for tritium recovery. In the present study, a solid breeding material, Li2TiO3, was irradiated by 1.7 keV deuterium ions, and an amount of retained deuterium and deuterium desorption behavior were investigated using a thermal desorption. Dependence of deuterium fluence on amount of retained deuterium was also obtained. In order to examine trapping mechanisms of deuterium in Li2TiO3, similar experiments were conducted for Li and Ti. Deuterium implanted to Li2TiO3 desorbed in forms of HD, D2, HDO and D2O. The amount of deuterium desorbed in form of HD was a few times larger than those of other gas species. The desorption peak appeared at 600 K, but significant desorption up to 900 K was observed. The range of temperature in the lithium titanate of the blanket module is assumed from 550 K to 1200 K. These results suggest that the tritium produced in the blanket is partly not desorbed. Thus, the spatial distribution of temperature in the blanket has to be controlled for the tritium to be desorbed during the operation. The desorption spectra of deuterium in Li2TiO3 were similar to those of Li. This suggests that most of implanted deuterium is trapped in forms of Li-D and Li-OD. Based upon the

  3. Production enhancement and quality degradation of Pu produced in FBR blankets

    International Nuclear Information System (INIS)

    Full text: For smooth deployment of FBR in near future, securing economy and non-proliferation are pivotal factors. This study has two main objectives related to these two key factors. One is to enhance Pu production efficiency in blanket region of fast breeder reactor core. The other is Pu composition degradation to improve the proliferation resistance of FBR fuel cycle. The former contributes to reduce amount of blanket fuels to be fabricated and reprocessed to gain the same quantity of Pu, consequently it is expected to improve the fuel cycle economy. The composition of Pu generated in FBR blanket generally has very high quality, namely, equal to weapon or super grade. The latter objective is to deteriorate the quality of generated Pu by adjusting neutron spectrum in the blanket region and the initial composition of blanket fuels. A conceptual core design of MOX fueled, sodium cooled FBR with 1500MWe rating is performed by using SLAROM and CIRATION code with nuclear data set prepared for fast reactors of JFS-3-J3.2. The initial fuels in the active core region contains 5% of minor actinides produced in UO2-fueled PWR with enrichment of 4.1% and burnup of 43GWd/t after 10 years cooling. A index FIR (fissile inventory ratio at EOC and BOC) is used to measure the net fissile balance that is suited for fuel cycle mass balance analysis in addition to the BR (breeding ratio) in the standard definition. The moderator material used to tailor the neutron spectrum in the blanket region is ZrH. The depleted uranium pins in the radial blanket are replaced by ZrH pins in the volume fraction range from 0% to 30%. By increasing ZrH pins, the Pu production per unit mass of UO2 in blanket increased from 0.08kg-Pu/kg-UO2 to more than 0.1kg-Pu/kg-UO2. The FIR also slightly improved by replacing small fraction of UO2 by ZrH pins however it turned for the worse in the higher range more than 5% of ZrH. Loading ZrH into blanket fuel assembly drastically affects the Pu composition at

  4. Family Structure, Materialism, and Compulsive Consumption.

    OpenAIRE

    Rindfleisch, Aric; Burroughs, James E; Denton, Frank

    1997-01-01

    Despite the rapid and dramatic changes in the structure of the American family over the past thirty years (e.g., divorce, single parenting), consumer researchers have largely neglected the issue of how alternative family forms influence consumer behavior. The authors' initial inquiry into this area finds that young adults reared in disrupted families are more materialistic and exhibit higher levels of compulsive consumption than young adults reared in intact families. Furthermore, they show t...

  5. Reactor structural materials study by PGNAA

    International Nuclear Information System (INIS)

    Prompt gamma neutron activation analysis (PGNAA) facility at Pakistan Research Reactor (PARR-I) has been used for the study of different kinds of materials. High efficiency, high-resolution gamma spectrometer based on high purity germanium detector is being employed. The cadmium ratio of the incident neutron beam is about 60 and the thermal neutron flux at the target positron is ∼2·106 n/cm2/s. The system is usually operated in Compton suppression mode to identify lower intensity gamma ray peaks in the lower energy region. The detection system is calibrated with standard radioactive sources from up to 1.5 MeV. In the higher energy region (up to 10 MeV), the energy calibration is done with prompt gamma rays from thermal neutron capture in chlorine and nitrogen. The aim of present study is to determine the quantity of boron in reactor grade steels as well as in some other materials containing boron because it is difficult to measure the concentration of boron by other analytical techniques. The results of boron assessment in SS-304L and ribbon will also be presented

  6. Detailed mechanical design and manufacturing study for the ITER reference breeding blanket

    International Nuclear Information System (INIS)

    This papers relates on the detailed mechanical design, manufacturing feasibility and assembly analysis of a water-cooled solid breeding blanket concept, selected as the ITER reference design. This breeding blanket design is characterised by: i) pressurised water flowing inside flat steel panels for cooling of the internals; each panel is welded along its contour onto the first wall structure and to the rear shield plate after closure of the module (last assembly step). ii) Beryllium (neutronic multiplier) in the form of micro-spheres filling the volume between parallel flat coolant panels. iii) Breeder pebbles enclosed in rods, which form bundles and are themselves embedded inside the Beryllium micro-spheres. (authors)

  7. NONLINEAR BUCKLING CHARACTERISTIC OF GRADED MULTIWEB STRUCTURE OF HETEROGENEOUS MATERIALS

    Institute of Scientific and Technical Information of China (English)

    LI Yong; ZHANG Zhi-min

    2005-01-01

    The graded multiweb structure of heterogeneous anisotropic materials, which makes full use of the continuous, gradual and changing physical mechanical performance of material properties, has a widespread application in aeroplane aerofoil structure and automobile lightweight structure. On the basis of laminate buckling theory,the equivalent rigidity method is adopted to establish the corresponding constitutive relation and the non-linear buckling governing equation for the graded multiweb structure. In finding the solution, the critical load of buckling under different complicated boundary conditions together with combined loads were obtained and testification of the experimental analysis shows that the calculation results can satisfy the requirements of engineering design in a satisfactory way. Results obtained from the research say that: graded materials can reduce the concentrated stress on the interface in an effective way and weaken the effect of initial defect in materials and thereby improve the strength and toughness of materials.

  8. 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)

  9. Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

    International Nuclear Information System (INIS)

    A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket

  10. Structural disorder in ion irradiated carbon materials

    International Nuclear Information System (INIS)

    The effects of ion irradiation on carbon based materials are reviewed laying emphasis on the well known ability of carbon to have different kinds of bonding configuration with the surrounding atoms. It was found that two kinds of bonding configuration of the carbon atoms are allowed in solid amorphous carbon phases. These rearrange the four valence electrons of carbon into sp2 (trigonal bond) and sp3 (tetrahedral bond) hybridizations. Driving the trigonal carbon fraction (x), the physical and chemical nature of solid carbon materials can change in a dramatic way ranging from metallic (x∼100%) to insulating (x∼0%) through semiconductor properties. The amount of the tetrahedral (or trigonal) carbon atoms can be controlled by ion beam irradiation, using suitable conditions and/or introducing foreign species such as hydrogen or silicon by the implantation technique. In hydrogenated amorphous carbon (a-C:H) and hydrogenated amorphous silicon-carbon alloys (a-Si1-xCx:H), the ion beam effects are able to produce stable and reproducible compounds, achieved by tuning the hydrogen (silicon) concentration with well defined equilibrium curves between the trigonal carbon fraction and hydrogen (silicon) content. Raman spectroscopy and temperature dependent conductivity experiments performed on these alloys suggest clustering effects in samples with high carbon content (x∼0.5) due to the strong binding energy of the C-C double bond with respect to C-Si and Si-Si. Several models and theoretical studies such as the 'random covalent network' (RCN) and molecular dynamics calculations have been used to fit the experimental results. It is shown that, while RCN models are highly inaccurate because of the clustering effects, molecular dynamics calculation data are very close to the experimental measured physical properties and confirm the ability of the trigonal carbon to cluster in graphite-like aggregate

  11. Method for fabricating high aspect ratio structures in perovskite material

    Science.gov (United States)

    Karapetrov, Goran T.; Kwok, Wai-Kwong; Crabtree, George W.; Iavarone, Maria

    2003-10-28

    A method of fabricating high aspect ratio ceramic structures in which a selected portion of perovskite or perovskite-like crystalline material is exposed to a high energy ion beam for a time sufficient to cause the crystalline material contacted by the ion beam to have substantially parallel columnar defects. Then selected portions of the material having substantially parallel columnar defects are etched leaving material with and without substantially parallel columnar defects in a predetermined shape having high aspect ratios of not less than 2 to 1. Etching is accomplished by optical or PMMA lithography. There is also disclosed a structure of a ceramic which is superconducting at a temperature in the range of from about 10.degree. K. to about 90.degree. K. with substantially parallel columnar defects in which the smallest lateral dimension of the structure is less than about 5 microns, and the thickness of the structure is greater than 2 times the smallest lateral dimension of the structure.

  12. Flight service environmental effects on composite materials and structures

    Science.gov (United States)

    Dexter, H. Benson; Baker, Donald J.

    1992-01-01

    NASA Langley and the U.S. Army have jointly sponsored programs to assess the effects of realistic flight environments and ground-based exposure on advanced composite materials and structures. Composite secondary structural components were initially installed on commercial transport aircraft in 1973; secondary and primary structural components were installed on commercial helicopters in 1979; and primary structural components were installed on commercial aircraft in the mid-to-late 1980's. Service performance, maintenance characteristics, and residual strength of numerous components are reported. In addition to data on flight components, 10 year ground exposure test results on material coupons are reported. Comparison between ground and flight environmental effects for several composite material systems are also presented. Test results indicate excellent in-service performance with the composite components during the 15 year period. Good correlation between ground-based material performance and operational structural performance has been achieved.

  13. Advanced structural integrity assessment procedures. Working material

    International Nuclear Information System (INIS)

    The purpose of the meeting was to provide an international forum for discussion on recent results in research and utility practice in the field of methodology for the structural integrity assessment of components including relevant non-codified procedures. The scope of the meeting included deterministic and probabilistic approaches. The papers covered the following topics: Leak-before-break concepts; non-destructive examination (NDE) and surveillance results; statistical evaluation of non-destructive examination data; pressurized thermal shock evaluation; fatigue effects (including vibration); and verification qualification. The meeting was attended by 32 specialists from 8 countries. Refs, figs and tabs

  14. Overview of European Community (Activity 3) work on materials properties of fast reactor structural materials

    International Nuclear Information System (INIS)

    The Fast Reactor Coordinating Committee set up in 1974 the Working Group Codes and Standards, and organized its work into four main activities: Manufacturing standards, Structural analysis, Materials and Classification of components. The main purpose of materials activity is to compare and contrast existing national specifications and associated properties relevant to structural materials in fast reactors. Funds are available on a yearly basis for tasks to be carried out through Study Contracts. At present about four Study Contract Reports are prepared each year

  15. Development of Structural Materials for JSFR - Overview and Current Status

    International Nuclear Information System (INIS)

    This paper summarized the ongoing efforts regarding new core and structural materials that are to applied to the Japanese sodium cooled fast breeder reactors (JSFR) of which demonstration plant's operation is envisioned in around 2025. For core materials, oxide dispersed strengthened steel (ODS) and precipitation hardened (PH) ferritic steels will be applied. For structural materials, 316FR stainless steel and Modified 9Cr-1Mo steel are applied. The current status alloy design, acquisition of data necessary to establish material strength standard, fabrication techniques to meet the requirements of the design of JSFR both in terms of quality and quantity, and codification of the material strength standards regarding the new materials are overviewed. Further described is path forward to the application of the materials to the JSFR. (author)

  16. From materials development to their test in IFMIF: an overview

    OpenAIRE

    Baluc, Nadine; Schäublin, Robin; Spaetig, Philippe; Ilchuk, Nazar; Veleva, Lyubomira; Oksiuta, Zbigniew; Theile, Jürgen; Tran, Minh Quang

    2011-01-01

    R&D activities on fusion reactor materials in Switzerland focus on (1) the development of advanced metallic materials for structural applications in plasma-facing (first wall, divertor) and breeding blanket components of the future fusion power reactors, in particular oxide dispersion strengthened reduced activation ferritic steels and tungsten-base materials, (2) the modelling of radiation damage and radiation effects and (3) small specimen test technology for the future International Fusion...

  17. Neutron Scattering Studies of Nanomagnetism and Artificially Structured Materials

    Energy Technology Data Exchange (ETDEWEB)

    Fitzsimmons, M.R.; Bader, S.D.; Borchers, J.A.; Felcher, G.P.; Furdyna, J.K.; Hoffmann, A.; Kortright, J.B.; Schuller, Ivan K.; Schulthess, T.C.; Sinha, S.K.; Toney, M.F.; Weller, D.; Wolf, S.

    2003-02-01

    Nanostructured magnetic materials are intensively studied due to their unusual properties and promise for possible applications. The key issues in these materials relate to the connection between their physical properties (transport, magnetism, mechanical, etc.) and their chemical-physical structure. In principle, a detailed knowledge of the chemical and physical structure allows calculation of their physical properties. Theoretical and computational methods are rapidly evolving so that magnetic properties of nanostructured materials might soon be predicted. Success in this endeavor requires detailed quantitative understanding of the magnetic structure and properties.

  18. Space Transportation Materials and Structures Technology Workshop. Volume 2; Proceedings

    Science.gov (United States)

    Cazier, Frank W., Jr. (Compiler); Gardner, James E. (Compiler)

    1993-01-01

    The Space Transportation Materials and Structures Technology Workshop was held on September 23-26, 1991, in Newport News, Virginia. The workshop, sponsored by the NASA Office of Space Flight and the NASA Office of Aeronautics and Space Technology, was held to provide a forum for communication within the space materials and structures technology developer and user communities. Workshop participants were organized into a Vehicle Technology Requirements session and three working panels: Materials and Structures Technologies for Vehicle Systems, Propulsion Systems, and Entry Systems.

  19. Effects of material disintegration on structural performance

    International Nuclear Information System (INIS)

    Using a state variable approach it will be shown how the results of appropriately chosen tests can be used to develop constitutive equations which describe the effects of creep deterioration on the deformation of metals operating at elevated temperatures. In this way a systematic phenomenological approach can be developed. Using this approach in conjunction with the results of multiaxial stress tests a set of constitutive equations are proposed which define the creep deformation rates in terms of applied stress and a material state variable, referred to as damage. In spite of the fact that the equations have been comstructed using purely phenomenological procedures, it is possible, nevertheless to demonstrate the physical interpretation of the damage state. Using the constitutive equations in conjunction with the virtual work relationships it is possible to obtain bounds on the rupture life of components subjected to both constant and cyclic loading. It is found convenient to express the results in terms of a so-called representative rupture stress, so that the uniaxial creep rupture curves can be used directly, thereby avoiding complex curve fitting procedures. The bounds are used to predict the rupture life of a number of components subjected to constant and variable load and comparison is made with experimental results. The components tested are: plane stress plates with holes and slots, subjected to uniaxial stress; plane stress plates with holes and slots subjected to biaxial tension; cylindrical bars with round notches; reinforced shell intersections subjected to internal pressure

  20. Improved structural materials for fast breeder reactors

    International Nuclear Information System (INIS)

    Electricity plays a crucial role in the economic development of our country. Coal is the primary fuel for generation of electricity in India as in many other countries. In India, generation of power by nuclear reactors is very important because of (i) availability of large thorium resource, (ii) constraints on setting up of fossil fuel based power plants and (iii) the negligibly small green house gas emissions by nuclear energy. The nuclear programme of the country is being implemented in three stages: (i) pressurized heavy water reactors of the CANDU type, (ii) sodium-cooled fast reactors and (iii) thorium-based reactors. Sodium-cooled fast reactor (SFR) technology is envisioned to make use of the large thorium reserves available. India has undertaken and made rapid strides in developing SFR technology and building of fast reactors for energy generation. A Fast Breeder Test Reactor (FBTR) of 40 MWt is operating successfully for over 25 years at Indira Gandhi Centre for Atomic Research. Based on the design, construction and operational experience, a 500 MWe Prototype Fast Breeder Reactor (PFBR) has been designed indigenously and is in an advanced stage of construction. Its design is being further optimised for enhanced economy with respect to cost of electricity production, for use in commercial reactors. Currently, several R and D programmes are under implementation for the development of new materials required for improved economy of commercial fast reactors

  1. Microfabrication of hierarchical structures for engineered mechanical materials

    Science.gov (United States)

    Vera Canudas, Marc

    Materials found in nature present, in some cases, unique properties from their constituents that are of great interest in engineered materials for applications ranging from structural materials for the construction of bridges, canals and buildings to the fabrication of new lightweight composites for airplane and automotive bodies, to protective thin film coatings, amongst other fields. Research in the growing field of biomimetic materials indicates that the micro-architectures present in natural materials are critical to their macroscopic mechanical properties. A better understanding of the effect that structure and hierarchy across scales have on the material properties will enable engineered materials with enhanced properties. At the moment, very few theoretical models predict mechanical properties of simple materials based on their microstructures. Moreover these models are based on observations from complex biological systems. One way to overcome this challenge is through the use of microfabrication techniques to design and fabricate simple materials, more appropriate for the study of hierarchical organizations and microstructured materials. Arrays of structures with controlled geometry and dimension can be designed and fabricated at different length scales, ranging from a few hundred nanometers to centimeters, in order to mimic similar systems found in nature. In this thesis, materials have been fabricated in order to gain fundamental insight into the complex hierarchical materials found in nature and to engineer novel materials with enhanced mechanical properties. The materials fabricated here were mechanically characterized and compared to simple mechanics models to describe their behavior with the goal of applying the knowledge acquired to the design and synthesis of future engineered materials with novel properties.

  2. Estimation of radioactivity in structural materials of ETRR-1 reactor

    International Nuclear Information System (INIS)

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig

  3. Low Cost, Lightweight, Multifunctional Structural Shielding Materials Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR involves the development of a lightweight innovative material for use as structure and radiation shielding in one. APS has assembled a uniquely qualified...

  4. Handbook for structural analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    This paper described structural analysis method of radioactive material transport casks for use of a handbook of safety analysis and evaluation. Safety analysis conditions, computer codes for analyses and stress evaluation method are also involved in the handbook. (author)

  5. Application of Advanced Radiation Shielding Materials to Inflatable Structures Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This innovation is a weight-optimized, inflatable structure that incorporates radiation shielding materials into its construction, for use as a habitation module or...

  6. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    The document presents the status of nuclear reaction theory concerning optical model development, level density models and pre-equilibrium and direct processes used in calculation of neutron nuclear data for structural materials of fission and fusion reactors. 6 refs

  7. Classic and New Materials Used for Structural Rehabilitation. Case Study

    Science.gov (United States)

    Lute, M.

    2016-06-01

    New materials development with different combination of properties were always a challenge in terms of their adequate use in civil engineering. Introduction of carbon fibres as strength material for structures was a beginning of a new approach in structural rehabilitation, and sometimes meant the end of classic rehabilitation solution use. The present paper gives an example of a building rehabilitation that use a melt of both new and old solutions in order to achieve the optimum result for building itself. The problem was even more challenging, because the structure considered is only 22 years old, but having some design faults in terms of seismic behaviour and, in addition, one floor was added to existing structure. The chosen solution was a compromise between the use of old and new materials in places where their qualities were best suitable and their minuses could be compensated by the other material.

  8. A low-risk aqueous lithium salt blanket for engineering test reactors

    International Nuclear Information System (INIS)

    A simple blanket concept is proposed based on 1-3 wt.% lithium dissolved as a salt in low temperature (80 degrees C) and low pressure (0.1 MPa) water. This concept can provide, for example, a 0.5 tritium breeding ratio with 60% steel structure and 70% coverage. The use of neutron multipliers, other structural materials (especially zirconium alloys), higher coverage and higher lithium salt concentrations allows tritium breeding ratios over unity if necessary. Other advantages of this concept include the simple shield-like geometry, substantial structural volume for mechanical strength, excellent heat transfer ability of water coolant, efficient neutron and gamma shielding through the combination of high-Z structure and low-Z water, and conventional tritium recovery and control technology. This concept could initially provide the shielding needs for an engineering test reactor and later, by the addition of lithium salt and tritium recovery systems, also provide tritium breeding. This staged operation and liquid breeder/coolant allows control over the tritium inventory in the device without machine disassembly. 14 refs

  9. 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

  10. Material and structural instabilities of single-wall carbon nanotubes

    Institute of Scientific and Technical Information of China (English)

    J. Wu; K. C. Hwang; J. Song; Y. Huang

    2008-01-01

    The nonlinear atomistic interactions usually involve softening behavior. Instability resulting directly from this softening are called the material instability, while those unrelated to this softening are called the structural instability. We use the finite-deformation shell theory based on the interatomic potential to show that the tension instability of single-wall carbon nanotubes is the material instability, while the compression and torsion instabilities are structural instability.

  11. 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.)

  12. 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.

  13. 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

  14. 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

  15. Simultaneous dynamic electrical and structural measurements of functional materials

    Energy Technology Data Exchange (ETDEWEB)

    Vecchini, C.; Stewart, M.; Muñiz-Piniella, A.; Wooldridge, J. [National Physical Laboratory, Hampton Road, Teddington TW11 0LW (United Kingdom); Thompson, P.; McMitchell, S. R. C.; Bouchenoire, L.; Brown, S.; Wermeille, D.; Lucas, C. A. [XMaS, The UK-CRG, ESRF-The European Synchrotron, CS40220, F-38043, Grenoble Cedex 09 (France); Department of Physics, University of Liverpool, Liverpool L69 3BX (United Kingdom); Lepadatu, S. [National Physical Laboratory, Hampton Road, Teddington TW11 0LW (United Kingdom); Jeremiah Horrocks Institute, University of Central Lancashire, Preston PR1 2HE (United Kingdom); Bikondoa, O.; Hase, T. P. A. [XMaS, The UK-CRG, ESRF-The European Synchrotron, CS40220, F-38043, Grenoble Cedex 09 (France); Department of Physics, University of Warwick, Coventry CV4 7AL (United Kingdom); Lesourd, M. [ESRF-The European Synchrotron, CS40220, F-38043, Grenoble Cedex 09 (France); Dontsov, D. [SIOS Meßtechnik GmbH, Am Vogelherd 46, 98693 Ilmenau (Germany); Cain, M. G. [National Physical Laboratory, Hampton Road, Teddington TW11 0LW (United Kingdom); Electrosciences Ltd., Farnham, Surrey GU9 9QT (United Kingdom)

    2015-10-15

    A new materials characterization system developed at the XMaS beamline, located at the European Synchrotron Radiation Facility in France, is presented. We show that this new capability allows to measure the atomic structural evolution (crystallography) of piezoelectric materials whilst simultaneously measuring the overall strain characteristics and electrical response to dynamically (ac) applied external stimuli.

  16. Dosimetry methods for fuels, cladding and structural materials

    International Nuclear Information System (INIS)

    This volume of the proceedings of the symposium on reactor dosimetry covers the following topics: the metallurgy and dosimetry interface, radiation damage correlations of structural materials and damage analyses techniques, dosimetry for fusion materials, light water reactor pressure vessel surveillance in practice and irradiation experiments, fast reactor and reseach reactor characterization

  17. Analysis and Design of Biological Materials and Structures

    CERN Document Server

    Öchsner, Andreas; Altenbach, Holm

    2012-01-01

    This collection provides researchers and scientists with advanced analyses and materials design techniques in Biomaterials and presents mechanical studies of biological structures. In 16 contributions well known experts present their research on Stress and Strain Analysis, Material Properties, Fluid and Gas mechanics and they show related problems.

  18. Extended propagation model for interfacial crack in composite material structure

    Institute of Scientific and Technical Information of China (English)

    闫相桥; 冯希金

    2002-01-01

    An interfacial crack is a common damage in a composite material structure . An extended propaga-tion model has been established for an interfacial crack to study the dependence of crack growth on the relativesizes of energy release rates at left and right crack tips and the properties of interfacial material characterize thegrowth of interfacial crack better.

  19. Spectrum Slam Fatigue Loading of Sandwich Materials for Marine Structures

    OpenAIRE

    Burman, Magnus; Rosén, Anders; Zenkert, Dan

    2010-01-01

    Sandwich materials are more frequently used in high speed craft and ferries. The motivation is reduced weight and associated operational cost. The hull structure in these vessels is subjected to repeated (fatigue) slamming loads (high strain rate loading). Scantling societies treat sandwich materials differently in their design rules. In common reduction or safety factors on the static strength of sandwich materials are used calculating the design stress. In most rules there is no explicit co...

  20. Structure and properties of sintered tool gradient materials

    OpenAIRE

    L.A. Dobrzański; B. Dołżańska

    2010-01-01

    Purpose: The main objective of the presented is to elaborate the fabrication technology of novel sintered tool gradient materials on the basis of hard wolfram carbide phase with cobalt binding phase, and to carry out research studies on the structure and properties of the newly elaborated sintered tool gradient materials.Design/methodology/approach: The following research studies have been carried out to elaborate a new group of sintered tool gradient materials, wolfram carbide with cobalt ma...

  1. 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)

  2. Tunable structural color in organisms and photonic materials for design of bioinspired materials

    Directory of Open Access Journals (Sweden)

    Hiroshi Fudouzi

    2011-01-01

    Full Text Available In this paper, the key topics of tunable structural color in biology and material science are overviewed. Color in biology is considered for selected groups of tropical fish, octopus, squid and beetle. It is caused by nanoplates in iridophores and varies with their spacing, tilting angle and refractive index. These examples may provide valuable hints for the bioinspired design of photonic materials. 1D multilayer films and 3D colloidal crystals with tunable structural color are overviewed from the viewpoint of advanced materials. The tunability of structural color by swelling and strain is demonstrated on an example of opal composites.

  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. 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

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

    International Nuclear Information System (INIS)

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

  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. Drucker-Prager-Cap creep modelling of pebble beds in fusion blankets

    International Nuclear Information System (INIS)

    Modelling of thermal and mechanical behaviour of pebble beds for fusion blankets is an important issue to understand the interaction of solid breeder and beryllium pebble beds with the surrounding structural material. Especially the differing coefficients of thermal expansion of these materials cause high stresses and strains during irradiation induced volumetric heating. To describe this process, the coupled thermomechanical behaviour of both pebble bed materials has to be modelled. Additionally, creep has to be considered contributing to bed deformations and stress relaxation. Motivated by experiments, we use a continuum mechanical approach called Drucker-Prager/Cap theory to model the macroscopic pebble bed behaviour. The model accounts for pressure dependent shear failure, inelastic hardening, and volumetric creep. The elastic part is described by a nonlinear elasticity law. The model has been implemented by user-defined routines in the commercial finite-element code ABAQUS. To check the numerics, the implementation is compared to an analytical solution. Furthermore, the Drucker-Prager/Cap tool is applied to a single ceramic breeder bed subject to creep under volumetric heating

  8. FIRE PROTECTION OF TIMBER STRUCTURES STRENGTHENED WITH FRP MATERIALS

    Directory of Open Access Journals (Sweden)

    Radek Zigler

    2015-12-01

    Full Text Available Modern, progressive methods of structures’ strengthening based on the use of composite materials composed of high strength fibers (carbon, glass, aramid or basalt and matrices based on epoxy resins brings, among many indisputable advantages (low weight, high effectiveness, easy application etc. also some disadvantages. One of the major disadvantages is a low fire resistance of these materials due to the low glass transition temperature Tg of the resin used. Based on an extensive research of strengthening of historic structures with FRP materials [1], the article outlines possible approaches to this problem, especially while strengthening timber load- bearing structures of historic buildings.

  9. Spectral fine structure effects on material and doppler reactivity worth

    International Nuclear Information System (INIS)

    New formulations concerning the fine structure effects on the reactivity worth of resonances are developed and conclusions are derived following the extension to more general types of perturbations which include: the removal of resonance material at finite temperatures and the temperature variation of part of the resonance material. It is concluded that the flux method can overpredict the reactivity worth of resonance materials more than anticipated. Calculations on the Doppler worth were carried out; the results can be useful for asessing the contribution of the fine structure effects to the large discrepancy that exists between the calculated and measured small sample Doppler worths. (B.G.)

  10. Challenges of structural materials for innovative nuclear systems in Europe

    International Nuclear Information System (INIS)

    New fusion and fission reactors for generation IV are envisaged to operate at conditions well above the actual ones for commercial fission reactors. This type of reactor combined a high operation temperature with a high neutron dose and an aggressive coolant, which imply new challenges for structural materials. One of the key issues to assure the safety and feasibility of these new nuclear systems is the selection of the structural materials, especially for in core components. Beside the differences between them, especially the amount of transmutation He in fusion reactors, similar structural materials have been selected. Some of the selected materials are well characterized at least at medium temperatures, as conventional ferritic/martensitic steels, but the qualification for higher temperatures is needed. For other materials, as ODS steels, there is a need for a complete characterization and qualification. In this paper a review of the operating conditions and selected structural materials for generation IV and fusion reactors within Europe is made. The needs for a complete characterization of these candidate materials are identified in terms of high temperature behaviour, radiation damage and coolant compatibility. (author)

  11. Fluoridation and oxidation behavior of JLF-1 and NIFS-HEAT-2 low activation structural materials

    International Nuclear Information System (INIS)

    Full text of publication follows: Corrosion of structural materials, such as JLF-1 (Fe-9Cr-2W-0.1C-V-Ta) and NIFSHEAT- 2 (V-4Cr-4Ti), is paid attention most in Flibe (LiF-BeF2) blanket development. In a previous corrosion study on steels, only oxides were identified as corrosion products after a conventional Flibe exposure test, while fluorides were indicated in another report with high purity Flibe after dehydration treatment. In order to clarify the corrosion mechanism in Flibe, it is essential to understand the competitive process by fluoridation and oxidation. Purpose of the present study is to characterize corrosion products of the low activation materials after fluoridation, oxidation and Flibe exposure tests, and to evaluate corrosion resistance of the materials in various conditions. JLF-1 JOYO-II heat and NIFS-HEAT-2 were machined and polished into specimens with 1 x 10 x 20 mm or 1 x 10 x 15 mm in size. The same size specimens of pure Fe, Cr, W, V and Ti were also prepared. The specimens were exposed to HF-H2O solution at room temperature, and to flowing He-HF-O2 gas mixture and static molten salt Flibe at 673-873 K. High purity HF gas was made by a reaction between NiF2 and dehydrated He-H2 gas mixture. Flibe was dehydrated by HF gas bubbling treatment. After the exposure tests, weight change was measured. Corrosion products and microstructural change at the surface were characterized by scanning electron microscopy (SEM) and X-ray diffractometry (XRD). In a XRD spectrum of JLF-1 after Flibe exposure at 823 K for 2003 hr, some peaks considered as FeF2 and CrF2 were observed, however their intensity was not enough to identify the compounds. The morphology and composition of the corrosion products were analyzed by SEM and compared with the one made in accelerated fluoridation tests with various composition of He-HF-O2 gas mixture. (authors)

  12. 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

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

    Science.gov (United States)

    Bosia, G.; Ragona, R.

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-10

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

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

    International Nuclear Information System (INIS)

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

  16. New progress on design and R and D for solid breeder test blanket module in China

    Energy Technology Data Exchange (ETDEWEB)

    Feng, K.M., E-mail: fengkm@swip.ac.cn; Zhang, G.S.; Hu, G.; Chen, Y.J.; Feng, Y.J.; Li, Z.X.; Wang, P.H.; Zhao, Z.; Ye, X.F.; Xiang, B.; Zhang, L.; Wang, Q.J.; Cao, Q.X.; Zhao, F.C.; Wang, F.; Liu, Y.; Zhang, M.C.

    2014-10-15

    Highlights: • The new progress on design and R and D of Chinese solid breeder TBM are introduced. • The mock-up fabrication and component tests for Chinese HCCB TBM have being developed. • The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CFL-1 are being prepared. • The fabrication of 1/3 sized mock-up is being carried-out. • The key technology development is proceeding to the large-scale mock-up fabrication. - Abstract: ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R and D activities for each TBM module with the auxiliary system are introduced. The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R and D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.

  17. Point Defect Phenomena of Crystalline Structure in Some Common Structural Materials

    Institute of Scientific and Technical Information of China (English)

    RUAN Yu-Zhong; WU Ren-Ping; YU Yan

    2005-01-01

    The existence and its movement rule of crystalline structure defect are closely related to the diffusion, solid phase reaction, sintering, phase transformation as well as the physical and chemical properties of materials. Point defect theory has been widely applied in material mineralization research, unfavorable transformation controlling, material modification,the research and development of new materials and so on. Point defect theory is one of the important theories for new material research and development. Herein we mainly discuss the application of point defect theory in some structural material researches.

  18. NASA Subsonic Rotary Wing Project - Structures and Materials Discipline

    Science.gov (United States)

    Halbig, Michael C.; Johnson, Susan M.

    2008-01-01

    The Structures & Materials Discipline within the NASA Subsonic Rotary Wing Project is focused on developing rotorcraft technologies. The technologies being developed are within the task areas of: 5.1.1 Life Prediction Methods for Engine Structures & Components 5.1.2 Erosion Resistant Coatings for Improved Turbine Blade Life 5.2.1 Crashworthiness 5.2.2 Methods for Prediction of Fatigue Damage & Self Healing 5.3.1 Propulsion High Temperature Materials 5.3.2 Lightweight Structures and Noise Integration The presentation will discuss rotorcraft specific technical challenges and needs as well as details of the work being conducted in the six task areas.

  19. Corrosion of structural materials for Generation IV systems

    International Nuclear Information System (INIS)

    The Generation IV International Forum aims at developing future generation nuclear energy systems. Six systems have been selected for further consideration: sodium-cooled fast reactor (SFR), gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR) and very high temperature reactor (VHTR). CEA, in the frame of a national program, of EC projects and of the GIF, contributes to the structural materials developments and research programs. Particularly, corrosion studies are being performed in the complex environments of the GEN IV systems. As a matter of fact, structural materials encounter very severe conditions regarding corrosion concerns: high temperatures and possibly aggressive chemical environments. Therefore, the multiple environments considered require also a large diversity of materials. On the other hand, the similar levels of working temperatures as well as neutron spectrum imply also similar families of materials for the various systems. In this paper, status of the research performed in CEA on the corrosion behavior of the structural material in the different environments is presented. The materials studied are either metallic materials as austenitic (or Y, La, Ce doped) and ferrito-martensitic steels, Ni base alloys, ODS steels, or ceramics and composites. In all the environments studied, the scientific approach is identical, the objective being in all cases the understanding of the corrosion processes to establish recommendations on the chemistry control of the coolant and to predict the long term behavior of the materials by the development of corrosion models. (author)

  20. Low-Cost Composite Materials and Structures for Aircraft Applications

    Science.gov (United States)

    Deo, Ravi B.; Starnes, James H., Jr.; Holzwarth, Richard C.

    2003-01-01

    A survey of current applications of composite materials and structures in military, transport and General Aviation aircraft is presented to assess the maturity of composites technology, and the payoffs realized. The results of the survey show that performance requirements and the potential to reduce life cycle costs for military aircraft and direct operating costs for transport aircraft are the main reasons for the selection of composite materials for current aircraft applications. Initial acquisition costs of composite airframe components are affected by high material costs and complex certification tests which appear to discourage the widespread use of composite materials for aircraft applications. Material suppliers have performed very well to date in developing resin matrix and fiber systems for improved mechanical, durability and damage tolerance performance. The next challenge for material suppliers is to reduce material costs and to develop materials that are suitable for simplified and inexpensive manufacturing processes. The focus of airframe manufacturers should be on the development of structural designs that reduce assembly costs by the use of large-scale integration of airframe components with unitized structures and manufacturing processes that minimize excessive manual labor.

  1. Liquid-metal fast breeder reactor structural materials design considerations

    International Nuclear Information System (INIS)

    This paper gives a brief overview of the LMFBR, to describe its key components, addresses two key structural problems, reviews high-temperature materials utilized, and places bounds on expected operating conditions. The current status of materials utilization in the LMFBR is summarized as follows: with the exception of the reactor upper internals, design needs for the LMFBR can be met with currently approved Code materials; Inconel 718 can potentially solve the thermal striping problems in the reactor upper internals; temperature, stress-strain levels, and design lifetime of the LMFBR push currently approved Code materials toward their limits of usefulness

  2. Recent developments of discrete material optimization of laminated composite structures

    DEFF Research Database (Denmark)

    Lund, Erik; Sørensen, Rene

    2015-01-01

    This work will give a quick summary of recent developments of the Discrete Material Optimization approach for structural optimization of laminated composite structures. This approach can be seen as a multi-material topology optimization approach for selecting the best ply material and number of...... plies in a laminated composite structure. The conceptual combinatorial design problem is relaxed to a continuous problem such that well-established gradient based optimization techniques can be applied, and the optimization problem is solved on basis of interpolation schemes with penalization. The...... it possible also to include local criteria such as strength criteria in the formulations. This is illustrated by structural optimization of a corner hinged laminated plate in this paper, and at ICCM20 it will also be demonstrated for optimization of a main spar from a wind turbine blade....

  3. 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

  4. Basic reactions of osteoblasts on structured material surfaces

    Directory of Open Access Journals (Sweden)

    B�chter A.

    2005-04-01

    Full Text Available In order to assess how bone substitute materials determine bone formation in vivo it is useful to understand the mechanisms of the material surface/tissue interaction on a cellular level. Artificial materials are used in two applications, as biomaterials alone or as a scaffold for osteoblasts in a tissue engineering approach. Recently, many efforts have been undertaken to improve bone regeneration by the use of structured material surfaces. In vitro studies of bone cell responses to artificial materials are the basic tool to determine these interactions. Surface properties of materials surfaces as well as biophysical constraints at the biomaterial surface are of major importance since these features will direct the cell responses. Studies on osteoblastlike cell reactivity towards materials will have to focus on the different steps of protein and cell reactions towards defined surface properties. The introduction of new techniques allows nowadays the fabrication of materials with ordered surface structures. This paper gives a review of present knowledge on the various stages of osteoblast reactions on material surfaces, focused on basic cell events under in vitro conditions. Special emphasis is given to cellular reactions towards ordered nano-sized topographies.

  5. Irradiation environment and materials behavior

    International Nuclear Information System (INIS)

    Irradiation environment is unique for materials used in a nuclear energy system. Material itself as well as irradiation and environmental conditions determine the material behaviour. In this review, general directions of research and development of materials in an irradiation environment together with the role of materials science are discussed first, and then recent materials problems are described for energy systems which are already existing (LWR), under development (FBR) and to be realized in the future (CTR). Topics selected are (1) irradiation embrittlement of pressure vessel steels for LWRs, (2) high fluence performance of cladding and wrapper materials for fuel subassemblies of FBRs and (3) high fluence irradiation effects in the first wall and blanket structural materials of a fusion reactor. Several common topics in those materials issues are selected and discussed. Suggestions are made on some elements of radiation effects which might be purposely utilized in the process of preparing innovative materials. (J.P.N.) 69 refs

  6. 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

  7. Preliminary thermo-mechanical analysis of ITER breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

  8. Ripple effect and impact of test blanket module by using RAFM steels on the plasma of ITER

    International Nuclear Information System (INIS)

    The losses of high-energy particles from the plasma depend on the toroidal field (TF) ripple in Tokomak machine. TBM (test blanket module), using RAFM (reduced activation ferritic/martensitic) steels as structure material, impacts on TF ripple in International Thermonuclear Experimental Reactor (ITER). The aim in this paper was to investigate the impact of TBM on TF ripple in ITER. It was analyzed based on ANSYS code and the Chinese DFLL (Dual Function Lithium Lead)-TBM as instances of analysis. The results indicated the TF ripple was still beyond the acceptable level of ITER (δTF < 0.3%) while considering several kinds of configurations (different masses, different dimensions, and different distances to plasma) of the DFLL-TBM. The correction coil might be one way to further reduce the effect on ripple of TF, and the ferromagnetic inserts under TF coil need to continue optimized.

  9. Neutral gas blanket effects in a gaseous divertor

    International Nuclear Information System (INIS)

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

  10. Fluctuation electron microscopy studies of complex structured materials

    Science.gov (United States)

    Zhao, Gongpu; Rougée, Annick; Buseck, Peter; Treacy, Michael

    2008-03-01

    Fluctuation electron microscopy (FEM) is a hybrid imaging-diffraction technique. This technique is particularly sensitive to paracrystalline structures of dimension 0.5-2 nm, which are difficult to detect by either imaging or diffraction techniques alone. It has been successfully deployed to study paracrystalline structures in amorphous silicon, germanium thin film. This technique has also been used to study metallic glasses and oxide glasses. Until now, FEM has not been used to study disordered geological materials. In this talk we present our FEM studies of shungite, a naturally occurring disordered carbonaceous material, reveal that trace quantities of tightly curved graphene structures such as C60, or fragments of C60, is present in shungite. We also present results from our study of metamict zircon, whose crystal structure is destroyed by self-radiation during naturally occurring α decay events. Work is in progress to study the structural evolution during the metamictization process.

  11. 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

  12. 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...

  13. 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

  14. 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

  15. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  16. Nuclear materials - fissile, fertile and dual-use structural materials involved in nuclear reactors

    International Nuclear Information System (INIS)

    The article presents a brief account of nuclear materials, with special emphasis on fissile, fertile and some important dual-use structural materials generally involved in nuclear reactors. The dual-use structural materials utilized in nuclear reactors have got important applications in both nuclear and non-nuclear fields. In the hostile environment, the important phenomena such as interactions between fission products and the surrounding elemental species, radiation-induced effects, corrosion, generation of gases, swelling and so forth, become increasingly complex and the performance of a nuclear reactor system thus becomes very much dependent on the physicochemical stability and nuclear compatibility of the dual-use structural materials used in fuel sub-assembly towards the fuel elements. In this context, the dual-use structural materials like stainless steel, zirconium alloys, etc. as claddings; water, liquid sodium or gases, etc. as coolants and water, boron, etc. as moderators, having good reliability and appropriate nuclear compatibility with the fuels, are of prime importance in reactor technology. In advanced designed reactors, development of novel fuels coupled with efficient dual-use structural materials may mitigate the challenges involved in optimizing the efficiency of the power reactors under specific experimental conditions. (author)

  17. Structure and properties of sintered tool gradient materials

    Directory of Open Access Journals (Sweden)

    L.A. Dobrzański

    2010-12-01

    Full Text Available Purpose: The main objective of the presented is to elaborate the fabrication technology of novel sintered tool gradient materials on the basis of hard wolfram carbide phase with cobalt binding phase, and to carry out research studies on the structure and properties of the newly elaborated sintered tool gradient materials.Design/methodology/approach: The following research studies have been carried out to elaborate a new group of sintered tool gradient materials, wolfram carbide with cobalt matrix, to elaborate their fabrication technology and to determine their structure and properties: a fabrication technology of mixtures and the formation technology of wolfram carbide gradient materials with cobalt matrix WC-Co was applied and elaborated; sintering conditions were selected experimentally: time, temperature and sintering atmosphere as well as isostatic condensation, ensuring the best structure and properties; phase and chemical composition of the sintered gradient WC-Co materials was determined using EDX, EBSD methods and qualitative X-ray analysis; the structure of sintered gradient WC-Co materials was investigated using scanning microscopy and transmission electron microscopy; mechanical and physical properties of sintered gradient WC-Co materials was determined: porosity, density, hardness, resistance to abrasive wear, resistance to brittle cracking.Findings: The presented research results confirm that the newly elaborated technology of powder metallurgy, which consists in sequential coating of the moulding with layers having the increasing content of carbides and decreasing concentration of cobalt, and then sintering such a compact, ensures the acquisition of the required structure and properties, including the resistance to cracking and abrasive wear of tool gradient materials, due to earned high hardness and resistance to abrasive wear on the surface as well as high resistance to cracking in the core of the materials fabricated in such a

  18. Surface modification of active material structures in battery electrodes

    Science.gov (United States)

    Erickson, Michael; Tikhonov, Konstantin

    2016-02-02

    Provided herein are methods of processing electrode active material structures for use in electrochemical cells or, more specifically, methods of forming surface layers on these structures. The structures are combined with a liquid to form a mixture. The mixture includes a surface reagent that chemically reacts and forms a surface layer covalently bound to the structures. The surface reagent may be a part of the initial liquid or added to the mixture after the liquid is combined with the structures. In some embodiments, the mixture may be processed to form a powder containing the structures with the surface layer thereon. Alternatively, the mixture may be deposited onto a current collecting substrate and dried to form an electrode layer. Furthermore, the liquid may be an electrolyte containing the surface reagent and a salt. The liquid soaks the previously arranged electrodes in order to contact the structures with the surface reagent.

  19. Materializing a responsive interior: designing minimum energy structures

    DEFF Research Database (Denmark)

    Mossé, Aurélie; Kofod, Guggi; Ramsgaard Thomsen, Mette

    2011-01-01

    This paper discusses a series of design-led experiments investigating future possibilities for architectural materialization relying on minimum energy structures as an example of adaptive structure. The structures have been made as laminates of elastic membrane under high tension with flexible...... (Lendlein, Kelch 2002) or light (van Oosten, Bastiaansen et al. 2009). All in all, this approach could form a whole new design paradigm, in which efficient 2D-manufacturing can lead to highly flexible, low weight and adaptable 3D-structures. This is illustrated by the design and manufacture of electro...

  20. Strengthening of building structures using carbon composite materials

    Directory of Open Access Journals (Sweden)

    N.V. Paranicheva

    2010-03-01

    Full Text Available Currently, the question of ensuring the reliability of various building structures both at the stage of their construction and during operation is very urgent. There are a lot of different ways and constructive methods of structures strengthening. Аt the same time, traditional ways of concrete structures strengthening with steel reinforcement are such expensive, time consuming and in some cases require to interrupt the building operation. As an alternative, it is proposed to use composite materials based on high-strength carbon fibers.The authors consider the properties, advantages, disadvantages and the methods of application of these materials. This article presents results of a technical survey carried out in a public building in 2009. In this building the CFRP was used to strengthen concrete slabs, resting on the crossbar consoles. The calculation of the strength is adduced and the section selection is made. The authors demonstrate their conclusions about the feasibility of using carbon composite materials.