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Sample records for blanket shield module

  1. Manufacture of blanket shield modules for ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-11-15

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

  2. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

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

  3. ITER blanket, shield and material data base

    International Nuclear Information System (INIS)

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

  4. Strength analysis results for the RF 'modified' option of the shielding block for ITER blanket module

    International Nuclear Information System (INIS)

    In accordance with the combined analysis of 'reference' shield block (SB) design option the RF specialist have adopted the decision to develop the 'modified' one. The present article contains the results of strength analysis for 'modified' shield block design developed by the RF Domestic Agency. This stage of analysis is devoted to strength analysis of 'lid-SB'-welded joint for elastic and elasto-plastic approach. There two welding types have been considered: electron beam welding (EB-welding) and thermal welding in inert gas (TIG-welding). These results together with the technological R and D will be useful for final decision on the shield block manufacturing technology.

  5. 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

  6. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  9. The ITER EC H and CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    Energy Technology Data Exchange (ETDEWEB)

    Gessner, Robby, E-mail: robby.gessner@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ronden, Dennis [DIFFER – Dutch Institute for Fundamental Energy Physics, P.O. Box 1207, NL-3430 BE Nieuwegein (Netherlands); Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

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

    International Nuclear Information System (INIS)

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

  11. 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

  12. 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)

  13. 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

  14. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-15

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

  17. Ceramic helium-cooled blanket test module

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

  18. Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO

    International Nuclear Information System (INIS)

    Highlights: •A conceptual architectural model for a vertical maintenance DEMO is presented. •Novel concepts for a set of DEMO remote handling equipment are put forward. •Remote maintenance of a multi module segment blanket is found to be feasible. •The criticality of space in the vertical port is highlighted. -- Abstract: The anticipated high neutron flux, and the consequent damage to plasma-facing components in DEMO, results in the need to regularly replace the tritium breeding and radiation shielding blanket. The current European multi module segment (MMS) blanket concept favours a less invasive small port entry maintenance system over large sector transport concepts, because of the reduced impact on other tokamak systems – particularly the magnetic coils. This paper presents a novel conceptual remote maintenance strategy for a Vertical Maintenance Scheme DEMO, incorporating substantiated designs for an in-vessel mover, to detach and attach the blanket segments, and cask-housed vertical maintenance devices to open and close access ports, cut and join service connections, and extract blanket segments from the vessel. In addition, a conceptual architectural model for DEMO was generated to capture functional and spatial interfaces between the remote maintenance equipment and other systems. Areas of further study are identified in order to comprehensively establish the feasibility of the proposed maintenance system

  19. Detailed 3-D nuclear analysis of ITER blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, T.D., E-mail: tdbohm@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Sawan, M.E.; Marriott, E.P.; Wilson, P.P.H. [University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, M.; Bullock, J. [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-10-15

    In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.

  20. Low activity aluminum blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-03-01

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

  1. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  2. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    International Nuclear Information System (INIS)

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available

  3. 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

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

  7. Manufacture of a shield prototype for primary wall modules

    International Nuclear Information System (INIS)

    In the frame of the BLANKET MODULE (BM) development for ITER, an R and D programme was implemented for the manufacture of a shield prototype by powder Hot Isostatic Pressing (HIPping). The manufactured shield is a full scale module No. 11a. Starting from a forged block of 1200 x 1200 x 500 mm, the main machining steps as deep drilling (1200 mm), 3D machining and sawing were performed. Tubes were 3D bent and large number of small parts were designed and machined. By welding together all the sub-parts we erected the main part of the water coolant circuit. Once the water circuit was built; the shield was completed using powder HIPping together with forged block embedding the tubes and their in a final solid part. The powder/solid HIP is used to minimize the number of BM seal welds in front of plasma. It increases the reliability of the components during operation. About 300 kg of stainless steel powder was densified together with the forged block. 3D measurement was done before and after the HIP cycle to collect the data to be compared with theoretical model. It allows to predict the main distortions of the solid bulk. Ultrasonic examination of the densified powder on the Stainless steel bulk and around the bended tubes was performed as well as mechanical characterization of the samples. The recess for stub key attachment on the vacuum vessel side, the hydraulic connector, the key for the primary wall panel attachment on the front side and the link between the four parallel water coolant circuits were then machined to achieve the shield prototype. (orig.)

  8. Manufacture of a shield prototype for primary wall modules

    International Nuclear Information System (INIS)

    In the frame of the blanket module (BM) development for ITER, an R and D programme was implemented for the manufacture of a shield prototype by powder hot isostatic pressing (HIPping). The manufactured shield is a full-scale module No. 11a. Starting from a forged block of 1350 mm x 1300 mm x 450 mm, the main machining steps as deep drilling (1200 mm), 3D machining and sawing were performed. Tubes were 3D bent and large number of small parts were designed and machined. By welding together all the sub-parts we erected the main part of the water coolant circuit. Once the water circuit was built; the shield was completed using powder HIPping together with forged block embedding the tubes in a final solid part. The powder/solid HIP is used to minimize the number of BM seal welds in front of plasma. It increases the reliability of the components during operation. About 300 kg of stainless steel powder was densified together with the forged block. 3D measurement was done before and after the HIP cycle to collect the data to be compared with theoretical model. It allows to predict the main distortions of the solid bulk. Ultrasonic examination of the densified powder on the stainless steel bulk and around the bended tubes was performed as well as mechanical characterization of the samples. The recess for stub key attachment on the vacuum vessel side, the hydraulic connector, the key for the primary wall panel attachment on the front side and the link between the four parallel water coolant circuits were then machined to achieve the shield prototype

  9. ITER屏蔽包层活化分析%Activation analysis for ITER shielding blanket

    Institute of Scientific and Technical Information of China (English)

    杨琪; 李斌; 郑剑; 何桃; 蒋洁琼; 吴宜灿

    2016-01-01

    作为国际热核聚变实验堆(ITER)的重要部件之一,屏蔽包层承受高强度聚变中子辐照,需要定期更换和维修。当活化的屏蔽包层从 ITER 托卡马克装置移到热室时,可能会给工作人员造成严重的辐射照射,是 ITER大厅和热室屏蔽设计的重要辐射源。文中基于 ITER最新中子学分析基准模型和“二步法”停堆剂量计算方法,使用超级蒙特卡罗核计算仿真软件系统 SuperMC针对15号屏蔽包层建立精细的中子学模型,并计算分析包层的活化情况及最严重情况下的周围辐射剂量率,并初步应用于 ITER赤道窗口室的屏蔽分析。计算结果显示,单个包层周围最大剂量率为350 Sv/hr,当传送小车停留在赤道窗口室内时,窗口室屏蔽门外剂量率高于10 mSv/hr,不足以满足设计要求。%As one of the key components of the International thermonuclear experiment reactor (ITER),blankets will sustain radiation from fusion neutrons with high intensity and may need to be replaced and maintained regularly. During the maintenance,the cask with activated blankets will be transferred to hot cell from Tokamak,which will cause high level of radiation in the building and radiation exposure for workers. Employing the Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC),the activation of No.1 5 shielding blanket and the shutdown dose around was analyzed based on the latest ITER neutronics model named Blite-3. The results were applied in the shielding analysis for ITER equatorial port cell. From the results,the dose rate around one activated blanket should be as high as 350 Sv/hr. When the cask carrying four activated first walls was transferred to the equatorial port cell,the dose rate in the gallery outside the port cell could be more than 10 mSv/hr,not meeting with the design criteria.

  10. 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.

  11. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2010-06-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  12. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2007-08-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  13. Manufacturing and testing of full scale prototype for ITER blanket shield block

    International Nuclear Information System (INIS)

    Highlights: • 316L(N)-IG forged steel was successfully fabricated and qualified. • Related R&D activities were implemented to resolve the fabrication issues. • SB #8 FSP was successfully manufactured with conventional fabrication techniques. • All of the validation tests were carried out and met the acceptance criteria. - Abstract: Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D

  14. Manufacturing and testing of full scale prototype for ITER blanket shield block

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sa-Woong, E-mail: swkim12@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Duck-Hoi; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Sung-Ki [WONIL Co., Ltd., Haman (Korea, Republic of); Kang, Sung-Chan [POSCO Specialty Steel Co., Ltd., Changwon (Korea, Republic of); Zhang, Fu; Kim, Byoung-Yoon [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ahn, Hee-Jae; Lee, Hyeon-Gon; Jung, Ki-Jung [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-04-15

    Highlights: • 316L(N)-IG forged steel was successfully fabricated and qualified. • Related R&D activities were implemented to resolve the fabrication issues. • SB #8 FSP was successfully manufactured with conventional fabrication techniques. • All of the validation tests were carried out and met the acceptance criteria. - Abstract: Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D.

  15. Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

    Institute of Scientific and Technical Information of China (English)

    HUANG Qun-Ying; LI Jian-Gang; CHEN Yi-Xue

    2004-01-01

    @@ Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB)to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDS-FBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW. yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.

  16. Transient thermal and stress analyses of the ITER shielding blanket/first wall under off-normal conditions

    International Nuclear Information System (INIS)

    Transient thermal and stress analyses have been conducted with the following three off-normal conditions for the shielding blanket and first wall (FW) structure of International Thermonuclear Experimental Reactor (ITER). (1) Loss of Flow Accident (LOFA) (2) Loss of Coolant Accident (LOCA) (3) Power Excursion Condition (PEC) The main results obtained are as follows : 1) In case of FW LOFA/LOCA, time to reach 400degC is 18 s at Beryllium surface, in case of shield LOFA/LOCA, time to reach 400degC is 90 s at 316SS internal rib, and in case of FW and shield LOFA/LOCA, time to reach 400degC is 17 s at Beryllium surface. 2) In case of FW LOFA/LOCA, maximum temperatures to satisfy 3Sm limits are 280degC for FW Cu alloy and 285degC for 316SS internal rib and in case of shield LOFA/LOCA, maximum temperatures to satisfy 3Sm limits are 248degC for FW Cu alloy and 170degC for 316SS internal rib. However, detail design guideline for off-normal conditions should be established and the stress should be reevaluated in future. 3) In case of FW LOFA/LOCA, plasma must be shut-down in a few seconds after the initiation of these events so as to prevent excursion of FW temperature and stress, while plasma shut-down requirement could be relatively relaxed in case of shield LOFA/LOCA. 4) Stresses and displacements during FW LOFA and FW LOCA are nearly equal. So are those during shield LOFA and LOCA. 5) Power excursion up to 1.8 GW shows no problem. (author)

  17. RF test blanket sub-module with ceramic breeder and helium cooling for test in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kovalenko, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)]. E-mail: koval@nikiet.ru; Kapyshev, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Leshukov, A. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Poliksha, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Shatalov, G. [Russian Research Center ' Kurchatov Institute' , Kurchatov Square 1, 123182 Moscow (Russian Federation); Strebkov, Yu. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Strizhov, A. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Sviridenko, M. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)

    2006-02-15

    International thermonuclear experimental reactor (ITER) is anticipated as the only one step to DEMO fusion reactor. One of its main objectives is to demonstrate the availability and integration of technologies essential for a fusion reactor by testing of components for a future reactor including the test blanket modules (TBM) with different types of breeding materials. RF proposed to divide the TBM on two parts and to use two independent test blanket sub-modules (TBSM) which fixed on the frame in ITER horizontal experimental port for testing. CHC TBSM design description, its mechanical attachment on the frame, and principle schemes of helium cooling system and tritium cycle system are presented in this paper.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

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

  19. Study of the effects of corrugated wall structures due to blanket modules around ICRH antennas

    Energy Technology Data Exchange (ETDEWEB)

    Dumortier, Pierre; Louche, Fabrice; Messiaen, André; Vervier, Michel [LPP-ERM/KMS, EURATOM-Belgian State Association, TEC partner, B-1000 Brussels (Belgium)

    2014-02-12

    In future fusion reactors, and in ITER, the first wall will be covered by blanket modules. These blanket modules, whose dimensions are of the order of the ICRF wavelengths, together with the clearance gaps between them will constitute a corrugated structure which will interact with the electromagnetic waves launched by ICRF antennas. The conditions in which the grooves constituted by the clearance gaps between the blanket modules can become resonant are studied. Simple analytical models and numerical simulations show that mushroom type structures (with larger gaps at the back than at the front) can bring down the resonance frequencies, which could lead to large voltages in the gaps between the blanket modules and perturb the RF properties of the antenna if they are in the ICRF operating range. The effect on the wave propagation along the wall structure, which is acting as a spatially periodic (toroidally and poloidally) corrugated structure, and hence constitutes a slow wave structure modifying the wall boundary condition, is examined.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  1. Vehicle drive module having improved EMI shielding

    Science.gov (United States)

    Beihoff, Bruce C.; Kehl, Dennis L.; Gettelfinger, Lee A.; Kaishian, Steven C.; Phillips, Mark G.; Radosevich, Lawrence D.

    2006-11-28

    EMI shielding in an electric vehicle drive is provided for power electronics circuits and the like via a direct-mount reference plane support and shielding structure. The thermal support may receive one or more power electronic circuits. The support may aid in removing heat from the circuits through fluid circulating through the support. The support forms a shield from both external EMI/RFI and from interference generated by operation of the power electronic circuits. Features may be provided to permit and enhance connection of the circuitry to external circuitry, such as improved terminal configurations. Modular units may be assembled that may be coupled to electronic circuitry via plug-in arrangements or through interface with a backplane or similar mounting and interconnecting structures.

  2. Forces on liquid lithium modules in a tokamak blanket due to the pulsed poloidal magnetic field

    International Nuclear Information System (INIS)

    This paper treats cylindrical modules filled with liquid lithium in the presence of a steady toroidal magnetic field and a time-dependent poloidal field. Solutions for liquid lithium flows and formulas for the forces on the modules are presented for both axial and transverse poloidal fields. Numerical examples are presented for the design in the ORNL/Westinghouse Tokamak Blanket Study. The initial analysis ignores the ends of the modules and treats infinitely long pipes, but the effects of the ends are also treated. Calculations and conclusions based on the solutions for infinitely long pipes are not significantly altered by end effects

  3. Applicability of tungsten/EUROFER blanket module for the DEMO first wall

    International Nuclear Information System (INIS)

    In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14 MW/m2. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface

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

    International Nuclear Information System (INIS)

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

  5. Experimental estimate of tritium production parameters for RF test blanket module

    International Nuclear Information System (INIS)

    Tritium breeding ratio (TBR) is a most value among controlled fusion reactor parameters. One in targets of test blanket module (TBM) program is experimental investigation of the value. On the whole TBR can be submitted for consideration TBR = BTB/BTP (BTB: breaded tritium in blanket; BTP: burned tritium in plasma). To investigate a numerator of the formula a tritium production in breeding zone (TBZ) of the TBM has to be measured under ITER plasma experiments. Tritium and neutron monitoring system with some lithium and neutron sensors are proposed. Lithium ortho-silicate and lithium carbonate and the neutron detectors fit the task. Differences isotope lithum-6 and lithium-7 can be applied. For delivery/withdrawal of the detectors into/from the TBZ a pneumatic concept is suggested with using canals allocated in module. The canals pass through the module back wall and reach the attended area. These canals allow the insertion of activation foil and capsules with material probes during the dwell time or operational pauses. Casks for the detectors and the canal for conveying of the casks in the TBM before pulse and extraction after pulse are presented in this paper

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

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

  8. Preliminary piping layout and integration of European test blanket modules subsystems in ITER CVCS area

    Energy Technology Data Exchange (ETDEWEB)

    Tarallo, Andrea, E-mail: andrea.tarallo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Mozzillo, Rocco; Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Aiello, Antonio; Utili, Marco [ENEA UTIS, C.R. Brasimone, Bacino del Brasimone, I-40032 Camugnano, BO (Italy); Ricapito, Italo [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • The use of human modeling tools for piping design in view of maintenance is discussed. • A possible preliminary layout for TBM subsystems in CVCS area has been designed with CATIA. • A DHM-based method to quickly check for maintainability of piping systems is suggested. - Abstract: This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  10. 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.

  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. Achievements of the water cooled solid breeder test blanket module of Japan to the milestones for installation in ITER

    International Nuclear Information System (INIS)

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, Water Cooled Solid Breeder (WCSB) TBM is being developed. Six TBMs will be tested in ITER simultaneously, under the leadership of different countries. To ensure the installation of reliable TBMs, it is necessary to show feasibility on the TBM milestones for installation in ITER. This paper shows the recent achievements toward the milestones of ITER TBMs prior to the installation, that consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, it is necessary to show the consistency with ITER design on time with ITER design progress, targeting the detailed design final report in 2012. Structure design of the interfacing components between the WCSB TBM structure and the interfacing components (Common Frame and Backside Shielding) that are placed in a test port of ITER has been developed. The design work also consists of procedures of fabrication and replacement of TBM, the consistency with ITER port structure and TBM interface structure, and the layouts of the auxiliary systems of TBMs including the tritium extraction system and water cooling system. As for the module qualification, it is necessary to show fabrication capability and the integrity of prototypical size mockup in corresponding operation condition before the delivery of the TBM to ITER. A real scale first wall mock-up was successfully fabricated by using Hot Isostatic Pressing (HIP) method by structural material of reduced activation martensitic ferritic steel, F82H. High heat flux test with real cooling water condition is planned using this mock-up. Other essential R and Ds for the WCSB TBM also showed steady progress on investigation of mechanical behavior of breeder pebble beds, development of advanced breeder/multiplier pebble, neutron measurement technology for TBM and purge gas tritium recovery technology. As for safety milestones

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  15. Multiple Module Simulation of Water Cooled Breeding Blankets in K-DEMO Using Thermal-Hydraulic Analysis Code MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.

  16. Program plan for the DOE Office of Fusion Energy First Wall/Blanket/Shield Engineering Technology Program. Volume I. Summary, objectives and management. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    1982-08-01

    This document defines a plan for conducting selected aspects of the engineering testing required for magnetic fusion reactor FWBS components and systems. The ultimate product of this program is an established data base that contributes to a functional, reliable, maintainable, economically attractive, and environmentally acceptable commercial fusion reactor first wall, blanket, and shield system. This program plan updates the initial plan issued in November of 1980 by the DOE/Office of Fusion Energy (unnumbered report). The plan consists of two parts. Part I is a summary of activities, responsibilities and program management including reporting and interfaces with other programs. Part II is a compilation of the Detailed Technical Plans for Phase I (1982 to 1984) developed by the participants during Phase 0 of the program (July to December 1981).

  17. Major achievements of the European shield blanket R and D during the ITER EDA, and their relevance for future next step machines

    International Nuclear Information System (INIS)

    In the frame of the international thermonuclear experimental reactors (ITER) collaboration, the European home team (EU HT) has committed significant efforts on the R and D for the Shield Blanket. This paper summarises the main achievements of this programme, which have been obtained over the last 7 years. The depth of R and D extends from generic activities up to the manufacture of prototypes, but has, in accordance with the design progress, reached different stages of maturity for the various components. New ITER options being considered since early 1998 have not made these activities irrelevant. With few exceptions, the results are still applicable for less ambitious next step machines, or transferable to components with similar functions or requirements

  18. Major achievements of the European shield blanket R and D during the ITER EDA, and their relevance for future next step machines

    Energy Technology Data Exchange (ETDEWEB)

    Daenner, W. E-mail: daenner@ipp.mpg.de; Cardella, A.; Jones, L.; Lorenzetto, P.; Maisonnier, D.; Malavasi, G.; Peacock, A.; Rodgers, E.; Tavassoli, F

    2000-11-01

    In the frame of the international thermonuclear experimental reactors (ITER) collaboration, the European home team (EU HT) has committed significant efforts on the R and D for the Shield Blanket. This paper summarises the main achievements of this programme, which have been obtained over the last 7 years. The depth of R and D extends from generic activities up to the manufacture of prototypes, but has, in accordance with the design progress, reached different stages of maturity for the various components. New ITER options being considered since early 1998 have not made these activities irrelevant. With few exceptions, the results are still applicable for less ambitious next step machines, or transferable to components with similar functions or requirements.

  19. 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

  20. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for test blanket module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation four different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port two TBMs will be accommodated, in the port 16 will be the European helium-cooled pebble bed blanket. In different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for the exchange of TBMs. Two TBMs are mounted onto one common frame, into a port plug (PP), which offers a standardised interface to the vacuum vessel (VV). It is cantilevered with a flange to VV port extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs is also the standard operation of ITER. Several components of the helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system are integrated into the ancillary equipment unit (AEU), which during the operation will connect the port plug to the subsystems. The bigger part of the AEU is accommodated in the port cell and the rest part of it is penetrated into the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection/disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected here. A

  1. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for Test Blanket Module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation 4 different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port will be two TBMs accomodated, in the port 16 will be the the European Helium Cooled Pebble Bed blanket. In the different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for exchange the TBMs. Two TBMs are mounted into one common frame, into a Port Plug (PP), which offers a standardised interface to the Vacuum Vessel (VV). It is cantilevered with a flange to VV Port Extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs are also standard operation of ITER. Several components of the Helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system, etc. All of them is integrated into the Ancillary Equipment Unit (AEU) which during operation will connect the port plug to the sub systems. The bigger part of the AEU is accomodated in the Port Cell and the rest part of it is penetrate to the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection / disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected

  2. Fast Ion Effects During Test Blanket Module Simulation Experiments in DIII-D

    International Nuclear Information System (INIS)

    Fast beam-ion losses were studied in DIII-D in the presence of a scaled mockup of two Test Blanket Modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam-ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot which is predicted to be different among the various codes.

  3. The coolant purification system of the European test blanket modules: Preliminary design

    Energy Technology Data Exchange (ETDEWEB)

    Ciampichetti, A., E-mail: andrea.ciampichetti@enea.i [ENEA CR Brasimone, FPN-FISING, 40032 Camugnano (Bolivia, Plurinational State of) (Italy); Aiello, A.; Coccoluto, G. [ENEA CR Brasimone, FPN-FISING, 40032 Camugnano (Bolivia, Plurinational State of) (Italy); Ricapito, I. [Fusion for Energy, 08019 Barcelona (Spain); Liger, K. [CEA, DEN, DTN/STPA/LPC, Cadarache, 13108 St Paul-lez-Durance (France); Demange, D. [Karlsruhe Institute of Technology, ITEP-TLK, Postfach 36 40, 76021 Karlsruhe (Germany); Moreno, C. [EURATOM-CIEMAT Association, 28040 Madrid (Spain)

    2010-12-15

    The HCPB (Helium Cooled Pebble Bed) and HCLL (Helium Cooled Lithium Lead) Test Blanket Modules (TBMs), developed in EU to be tested in ITER, adopt helium at 80 bar as primary coolant. This paper contains a conceptual design of the TBMs Coolant Purification System (CPS) based on the need to remove permeated tritium and gas impurities. The following steps have been considered: identification of CPS design requirements; review of the purification systems developed for Helium Cooled Fission Reactors and proposed for DEMO Fusion Reactor; selection of the most promising technologies for CPS; indications on instrumentation and procedures for tritium balance. The proposed solution is a three-stage process constituted by an oxidiser to convert Q{sub 2} and CO to Q{sub 2}O and CO{sub 2}, an adsorption step, performed on molecular sieve at room temperature to remove Q{sub 2}O and CO{sub 2}, and a final step performed on a heated getter to remove residual impurities.

  4. 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)

    Solid breeder blankets are regarded as a near-at-hand blanket concept for a fusion power demonstration plant in Japan. Test blanket module (TBM) to be tested in ITER is the most important milestone to establish the fusion demonstration blanket. For the candidate TBM's, two types of TBM, water cooled solid breeder TBM, and a helium gas cooled solid breeder TBM have been proposed and designed in JAERI. For detailed performance study under operation and after shut down, detailed neutronics analysis gives the most important design conditions, such as, distribution of tritium breeding ratio, nuclear heating rate during operation, and induced activation and decay heat after termination of irradiation. In the analysis, neutron and gamma transportation was calculated by two dimensional analysis code, DOT3.5, for two TBMs. Nuclear reaction rate and induced activation rate were evaluated by APPLE-3 and ACT-4, respectively. The analysis model included configurations of thermo-mechanical test modules and surrounding common frames for both of He cooled and water cooled TBMs. By the neutronics analysis, TBR and contact dose rate by induced activation till one year after termination of the module testing have been evaluated. For the evaluation of induced activation level change and decay heat change, the transient decreases in one year after termination of the module testing have been calculated. The time duration of the module testing before termination of testing is assumed to be 133 continuous days of full power operation. The result of TBR analysis showed that TBR distribution in the toroidal direction of TBM is not significant, however, the neutron flux decreases in the region of sidewall of common frame made of SS and water. This result shows that there is relatively large neutron loss from the TBM to the common frame. Thus, it is considered that the TBR value observed in the TBM testing may be smaller than the estimation by one dimensional neutronics analysis which does

  5. Implementation of two-phase tritium models for helium bubbles in HCLL breeding blanket modules

    OpenAIRE

    Fradera, Jordi; Sedano, L.A.; Mas de les Valls Ortiz, Elisabet; Batet Miracle, Lluís

    2011-01-01

    Tritium self-sufficiency requirement of future DT fusion reactors involves large helium production rates in the breeding blankets; this might impact on the conceptual design of diverse fusion power reactor units, such as Liquid Metal (LM) blankets. Low solubility, long residence-times and high production rates create the conditions for Helium nucleation, which could mean effective T sinks in LM channels. A model for helium nano-bubble formation and tritium conjugate transport phen...

  6. ITER test blanket module error field simulation experiments at DIII-D

    Science.gov (United States)

    Schaffer, M. J.; Snipes, J. A.; Gohil, P.; de Vries, P.; Evans, T. E.; Fenstermacher, M. E.; Gao, X.; Garofalo, A. M.; Gates, D. A.; Greenfield, C. M.; Heidbrink, W. W.; Kramer, G. J.; La Haye, R. J.; Liu, S.; Loarte, A.; Nave, M. F. F.; Osborne, T. H.; Oyama, N.; Park, J.-K.; Ramasubramanian, N.; Reimerdes, H.; Saibene, G.; Salmi, A.; Shinohara, K.; Spong, D. A.; Solomon, W. M.; Tala, T.; Zhu, Y. B.; Boedo, J. A.; Chuyanov, V.; Doyle, E. J.; Jakubowski, M.; Jhang, H.; Nazikian, R. M.; Pustovitov, V. D.; Schmitz, O.; Srinivasan, R.; Taylor, T. S.; Wade, M. R.; You, K.-I.; Zeng, L.; DIII-D Team

    2011-10-01

    Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δv/v ~ 60% via non-resonant braking. Changes to global Δn/n, Δβ/β and ΔH98/H98 were ~3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

  7. Simbol-X Mirror Module Thermal Shields: II-Small Angle X-Ray Scattering Measurements

    International Nuclear Information System (INIS)

    The formation flight configuration of the Simbol-X mission implies that the X-ray mirror module will be open to Space on both ends. In order to reduce the power required to maintain the thermal stability and, therefore, the high angular resolution of the shell optics, a thin foil thermal shield will cover the mirror module. Different options are presently being studied for the foil material of these shields. We report results of an experimental investigation conducted to verify that the scattering of X-rays, by interaction with the thin foil material of the thermal shield, will not significantly affect the performances of the telescope.

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

  10. Residual stress in a laser welded EUROFER blanket module assembly using non-destructive neutron diffraction techniques

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, D.J., E-mail: d.hughes@warwick.ac.uk [WMG, University of Warwick, Coventry CV4 7AL (United Kingdom); Koukovini-Platia, E. [CERN, CH-1211 Geneva 23 (Switzerland); Heeley, E.L. [Department of Physical Sciences, Open University, Walton Hall, Milton Keynes MK7 6AA (United Kingdom)

    2014-02-15

    Highlights: • Residual stresses were determined in a welded EUROFER blanket assembly with integrated cooling channels. • Good agreement was seen between experimentally determined and predicted stresses. • We show that microstructure changes that occur in EUROFER steels during welding must be considered for residual stress determination. • An experimental route is proposed for validation of predicted stresses in reactor components using non-destructive diffraction techniques. - Abstract: Whilst the structural integrity and lifetime considerations in welded joints for blanket modules can be predicted using finite element software, it is essential to prove the validity of these simulations. This paper provides detailed analysis for the first time, of the residual stress state in a laser-welded sample with integral cooling channels. State-of-the-art non-destructive neutron diffraction was employed to determine the triaxial stress state and to understand microstructural changes around the heat affected zone. Synchrotron X-ray diffraction was used to probe the variation of strain-free lattice reference parameter around the weld zone allowing correction of the neutron measurements. This paper details an important experimental route to validation of predicted stresses in complex safety-critical reactor components for future applications.

  11. Residual stress in a laser welded EUROFER blanket module assembly using non-destructive neutron diffraction techniques

    CERN Document Server

    Hughes, D J; Heeley, E L

    2014-01-01

    Whilst the structural integrity and lifetime considerations in welded joints for blanket modules can be predicted using finite element software, it is essential to prove the validity of these simulations. This paper provides detailed analysis for the first time, of the residual stress state in a laser-welded sample with integral cooling channels. State-of-the-art non-destructive neutron diffraction was employed to determine the triaxial stress state and to understand microstructural changes around the heat affected zone. Synchrotron X-ray diffraction was used to probe the variation of strain-free lattice reference parameter around the weld zone allowing correction of the neutron measurements. This paper details an important experimental route to validation of predicted stresses in complex safety-critical reactor components for future applications.

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

    International Nuclear Information System (INIS)

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

  13. Modelling of 3D fields due to ferritic inserts and test blanket modules in toroidal geometry at ITER

    Science.gov (United States)

    Liu, Yueqiang; Äkäslompolo, Simppa; Cavinato, Mario; Koechl, Florian; Kurki-Suonio, Taina; Li, Li; Parail, Vassili; Saibene, Gabriella; Särkimäki, Konsta; Sipilä, Seppo; Varje, Jari

    2016-06-01

    Computations in toroidal geometry are systematically performed for the plasma response to 3D magnetic perturbations produced by ferritic inserts (FIs) and test blanket modules (TBMs) for four ITER plasma scenarios: the 15 MA baseline, the 12.5 MA hybrid, the 9 MA steady state, and the 7.5 MA half-field helium plasma. Due to the broad toroidal spectrum of the FI and TBM fields, the plasma response for all the n  =  1-6 field components are computed and compared. The plasma response is found to be weak for the high-n (n  >  4) components. The response is not globally sensitive to the toroidal plasma flow speed, as long as the latter is not reduced by an order of magnitude. This is essentially due to the strong screening effect occurring at a finite flow, as predicted for ITER plasmas. The ITER error field correction coils (EFCC) are used to compensate the n  =  1 field errors produced by FIs and TBMs for the baseline scenario for the purpose of avoiding mode locking. It is found that the middle row of the EFCC, with a suitable toroidal phase for the coil current, can provide the best correction of these field errors, according to various optimisation criteria. On the other hand, even without correction, it is predicted that these n  =  1 field errors will not cause substantial flow damping for the 15 MA baseline scenario.

  14. Breeding zone models of DEMO ceramic helium cooled blanket test module for testing in IVV-2M reactor

    International Nuclear Information System (INIS)

    The goal of DEMO ceramic helium cooled blanket test module (CHC BTM) is to demonstrate a breeding capability that would lead to tritium self-sufficiency in ITER reactor and to extract a high-grade heat suitable for electricity generation. Experimental validation of all the adopted design solutions is main important problem at design and calculation works carrying out in order to develop the CHC BTM. One important task for breeding zones feasibility validation is in-pile tests. Two models were developed and fabricated for testing in the fission IVV-2M reactor. Breeding zone is based on poloidal BIT-conception. The models structural material is ferrito-martensitic steel. Breeder material is lithium orthosilicate in pebble beds and pellet forms. Multiplier material is beryllium in pebble beds and porosity forms. The cooling is provided by helium at 10 MPa. The tritium produced in the breeder material is purged by the helium flow at 0.1-0.2 MPa. Designs of model description and experimental channel, results of neutronic and thermo-hydraulic calculations are presented in the paper. (orig.)

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

    International Nuclear Information System (INIS)

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

  16. Shielding experiments

    International Nuclear Information System (INIS)

    Shielding mock-up experiments for Prototype Fast Breeder Reactor (PFBR) and Advanced Heavy Water Reactor (AHWR) are carried out in shielding corner facility of APSARA reactor, to assess the overall accuracy of the codes and nuclear data used in reactor shield design. As APSARA is a swimming pool-type thermal reactor, for fast reactor experiments, typical fast reactor shielding facility was created by using uranium assemblies as spectrum converter. The flux was also enhanced by replacing water by air. Experiments have been carried out to study neutron attenuation through typical fast reactor radial and axial bulk shielding materials such as steel, sodium, graphite, borated graphite and boron carbide. A large number of reaction rates, sensitive to different regions of the neutron energy spectrum, were measured using foil activation and Solid State Nuclear Track Detector (SSNTD) techniques. These experimental results were analysed using computational tools normally used in design calculations, viz., discrete ordinate transport codes with multigroup cross section sets. Comparison of measured reaction rates with calculations provided suitable bias factors for parameters relevant to shield design, such as sodium activation, fast neutron fluence, fission equivalent fluxes etc. The measured neutron spectrum on the incident face of shield model compares well with the calculated fast reactor blanket leakage neutron spectrum. The comparison of calculated reaction rates within shield model indicate that the calculations suffer from considerable uncertainties, in shield models with boron carbide/borated graphite. For AHWR shielding experiments, no spectrum converter was used as it is also a thermal reactor. Radiation streaming studies through penetrations/ducts of various shapes and sizes relevant to AHWR shielding were carried out. (author)

  17. 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)

  18. Rotation Braking and Error Field Correction of the Test Blanket Module Induced Magnetic Field Error in ITER

    International Nuclear Information System (INIS)

    Full text: Experiments on DIII-D confirm that the tritium breeding test blanket modules (TBMs) in ITER will lead to a decrease of the plasma rotation in H-modes. Moreover, they suggest that long-wavelength correction fields applied with non-axisymmetric saddle coils will only be able to ameliorate a fraction of such a rotation reduction. The new finding obtained in rotating H-modes contrasts previous experiments, which showed that saddle coils are very effective in restoring resilience to locked modes in L-mode plasmas. The experiments use a TBM mock-up coil that has been especially designed to simulate the error field induced by the ferromagnetic steel of a pair of TBMs in one of ITER port. The TBM field is applied in rotating H-mode plasmas with shape, β and safety factor similar to the ITER baseline scenario. The n = 1 error field correction (EFC) is applied with a set of non-axisymmetric saddle coils (I-coil), whose currents are optimized in the presence of the TBM mock-up field using a newly developed non- disruptive technique that maximizes the plasma rotation. However, a test of the effectiveness of the TBM EFC yields that the optimized EFC can only recover approximately a quarter of the 30% rotation decrease attributed to the TBM error field. An alternative criterion to evaluate the 'goodness' of an EFC has been its effectiveness in canceling the n = 1 plasma response to the error field. Plasma response measurements in the TBM experiment show that the I-coil can indeed cancel the magnetic measurements of the n = 1 plasma response to the TBM mock-up field. The required currents are consistent with ideal MHD predictions using the IPEC code, but differ significantly from the currents that maximize the plasma rotation. The contrast between the limited effectiveness of n = 1 EFC in rotating H-modes and their ability to recover a low locking density in L-mode plasmas shows that the components of the non-axisymmetric field that braking the plasma at high rotation

  19. 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

  20. Progress on DCLL Blanket Concept

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-01

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

  1. Progress in design and study of ITER test blanket modules%ITER氚增殖实验包层设计研究进展

    Institute of Scientific and Technical Information of China (English)

    刘松林; 柏云清; 陈红丽; 李春京; 黄群英; 吴宜灿; FDS团队

    2009-01-01

    The International Thermonuclear Experimental Reactor (ITER) will be the first experimental D-T fusion reactor to provide an exclusive test platform of physics and engineering technology for research and development of fusion, where the technology of Test Blanket Module (TBM) in ITER is one of the most critical kernels to achieve fusion power in the future. According to defined concepts of DEMO blanket, the parties had proposed DEMOrelevant TBM, respectively, which would be to be tested during ITER operation. Design of proposed TBM concepts, R&D status, and recommended port allocation in ITER are introduced in this contribution.%国际热核实验反应堆(ITER)为人类开发聚变能提供重要的物理和工程技术实验平台,ITER氚增殖实验包层模块(TBM)技术是必须掌握的关键技术.参与ITER计划的成员国根据本国商用演示堆包层发展策略,分别提出了各自的实验包层概念,以便在ITER运行期间进行实验.本文对ITER-TBM目前已经开展和正在进行的主要设计研究工作进展进行总结,介绍了各方提出的设计方案、支撑设计的相关技术研究进展,以及合作实验窗口的分配现状.

  2. Simbol-X Mirror Module Thermal Shields: I - Design and X-Ray Transmission

    International Nuclear Information System (INIS)

    The Simbol-X mission is designed to fly in formation flight configuration. As a consequence, the telescope has both ends open to space, and thermal shielding at telescope entrance and exit is required to maintain temperature uniformity throughout the mirrors. Both mesh and meshless solutions are presently under study for the shields. We discuss the design and the X-ray transmission.

  3. Development of Solid Breeder Blanket at JAERI

    International Nuclear Information System (INIS)

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

  4. Shielding member for thermonuclear device

    International Nuclear Information System (INIS)

    In a thermonuclear device for shielding fast neutrons by shielding members disposed in a shielding vessel (vacuum vessel and structures such as a blanket disposed in the vacuum vessel), the shielding member comprises a large number of shielding wires formed fine and short so as to have elasticity. The shielding wires are sealed in a shielding vessel together with water, and when the width of the shielding vessel is changed, the shielding wires follow after the change of the width while elastically deforming in the shielding vessel, so that great stress and deformation are not formed thereby enabling to improve reliability. In addition, the length, the diameter and the shape of each of the shielding wires can be selected in accordance with the shielding space of the shielding vessel. Even if the shape of the shielding vessel is complicated, the shielding wires can be inserted easily. Accordingly, the filling rate of the shielding members can be changed easily. It can be produced more easily compared with a conventional spherical pebbles. It can be produced more easily than existent spherical shielding pebbles thereby enabling to reduce the production cost. (N.H.)

  5. 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.)

  6. Breeding blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-03-01

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

  7. Breeding blanket for Demo

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-12-31

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

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

    Science.gov (United States)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

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

  9. Assessment of the activation, decay heat, and waste disposal of the US helium-cooled ceramic breeder test blanket module in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Youssef, Mahmoud Z. [University of California-Los Angeles, Los Angeles, CA 90025 (United States); University of Wisconsin-Madison (United States)], E-mail: youssef@fusion.ucla.edu; Ying, Alice [University of California-Los Angeles, Los Angeles, CA 90025 (United States)

    2008-12-15

    The radioactivity inventory and decay heat in the US helium-cooled ceramic breeder (HCCB) test blanket module (TBM) have been assessed at shutdown and for several times thereafter. The sub-module will have its own FW and structural container box that houses the breeder and beryllium pebble bed units, arranged in an edge-on-configuration. Low activation ferritic steel (F82H) is used as the structure and helium is used as a coolant. The breeder beds are made of Li{sub 2}TiO{sub 3} pebbles in which lithium has been enriched up to 75% in Li-6. Pulsed operation mode is assumed. During operation in the D-T phase, the total heating rate in the TBM is {approx}263 kW. The total amount of tritium generated in the breeder and the beryllium multiplier is {approx}9 g and 0.07 g, respectively, after reaching the 0.3 MWa/m{sup 2} fluence limit. At shutdown, the total radioactivity and decay heat levels are {approx}0.89 MCi and {approx}0.002 MW, respectively. These values drop sharply after 1 min to {approx}0.098 MCi and {approx}0.0006 MW. The contribution from the F82H structure is the dominant one up to {approx}10 years following shutdown. After {approx}10 years, the contribution to the total activation and decay heat from the breeder material is the dominant one due to the generated tritium. The WDR of various components are far below unity and thus are well within ITER regulatory guidelines.

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  12. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Assessment of the activation, decay heat, and waste disposal of the US helium-cooled ceramic breeder test blanket module in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Youssef, M.; Ying, A. [California Univ., Los Angeles, CA (United States)

    2007-07-01

    The radioactivity inventory and after heat in the U.S. helium-cooled ceramic breeder (HCCB) test blanket module (TBM) have been accessed at shut down and for several times thereafter. Also assessed is the waste disposal rating (WDR) of its various components. The objectives are: (1) to provide the information needed for further safety assessment of the generated radionuclides and their volatility, as well as after heat on the safety operation of ITER, and (2) to aid in determining the waiting cooling period prior to removing and transporting the TBM for further treatment outside ITER site. The TBM is proposed to be placed in one of the three dedicated test ports of ITER. The current proposal is that it will occupy 1/3 of the horizontal upper half of a port next to Japan and Korea sub-modules. The sub-module will have its own FW and structural container box that houses the breeder and beryllium pebble bed units, arranged in an edge-on-configuration. Helium is used to cool the FW, sides of the box, and the internal plates. Conventional ferritic steel (F82H) is used as the structure. The sub-module has 71 cm height, 38.9 cm wide and 60 cm depth in the radial direction. The breeder beds are made of Li{sub 2}TiO{sub 3} pebbles with 94% theoretical density and 62% packing factor (as the beryllium pebbles). Lithium-6 is enriched to 75%. A 2 mm thick beryllium layer is used as a plasma facing material on the FW area subjected to 0.78 MW/m{sup 2} neutron wall load. Pulsed operation mode is assumed. Each pulse is assumed to be 400 s full flat top followed by 1800 s dwell time, during which the decay of the generated radionuclides are accounted for. The 500 MW pulses are assumed to be generated one after another until a fluence limit of 0.3 MWa/m{sup 2} is reached without replacing the TBM. This gives upper conservative estimates for the radioactive inventory and decay heat. During operation in the D-T phase, the total heating rate in the TBM is {proportional_to}263 KW. The

  16. 扩散连接技术在核聚变反应堆包层模块制造中的应用%Application of Diffusion Bonding Technique in Fabrication of Blanket Module Components of Nuclear Fusion Reactor

    Institute of Scientific and Technical Information of China (English)

    刘晨曦; 刘永长; 周晓胜; 马宗青; 王颖; 李会军; 杨建国

    2015-01-01

    国际受控热核聚变实验堆计划是全球规模最大、影响最深远的国际科研合作项目之一,有望彻底解决能源危机。核聚变反应堆关键部件———包层模块的结构复杂、体积庞大,且服役环境恶劣,焊接接头成为影响反应堆安全运行的薄弱环节。以扩散连接为代表的固相焊接技术对接头性能及组织影响较小,已逐渐取代熔化焊应用于包层模块复杂构件制造。在简要介绍扩散连接及其原理的基础上,对包层模块构件扩散连接的研究进展进行了阐述,包括低活化铁素体/马氏体钢及氧化物弥散强化钢构件的扩散连接,Be,W,SiC等其他先进高温材料的扩散连接等。%International Thermonuclear Experimental Reactor is one of the world′s largest and the most far-reaching in-ternational scientific collaborative projects, which is expected to solve the energy crisis.As a key built-up part, blanket module has complex structure and large size, and serves under harsh service conditions.The welding joints of blanket mod-ule have become the weak links affecting the operation of the nuclear fusion reactor.Solid-phase welding technology, repre-sented by diffusion bonding, have relatively low effect on the mechanical properties and microstructure of the joints, and has gradually taken the place of the fusion welding technology used for fabrication of the blanket module complex compo-nents.Based on the brief presentation of diffusion bondingand its bonding mechanism, the research progress in diffusion bonding of blanket module components was discussed in this paper, including the diffusion bonding of reduced activation ferritic/martensitic steels and oxide dispersion strengthened steels, and the diffusion bonding of Be, W, SiC and/or other advanced high-temperature materials.

  17. MFTF-α + T shield design

    International Nuclear Information System (INIS)

    MFTF-α+T is a DT upgrade option of the Tandem Mirror Fusion Test Facility (MFTF-B) to study better plasma performance, and test tritium breeding blankets in an actual fusion reactor environment. The central cell insert, designated DT axicell, has a 2-MW/m2 neutron wall loading at the first wall for blanket testing. This upgrade is completely shielded to protect the reactor components, the workers, and the general public from the radiation environment during operation and after shutdown. The shield design for this upgrade is the subject of this paper including the design criteria and the tradeoff studies to reduce the shield cost

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

    International Nuclear Information System (INIS)

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

  19. ITER-FEAT vacuum vessel and blanket design features and implications for the R and D programme

    International Nuclear Information System (INIS)

    A tight fitting configuration of the VV to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the TF ripple. The blanket modules are supported directly by the VV. A full-scale VV sector model has provided critical information related to fabrication technology, and the magnitude of welding distortions and achievable tolerances. This R and D validated the fundamental feasibility of the double-wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and the robustness of solid HIP joining was demonstrated in R and D, by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal. (author)

  20. 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

  1. Flexible armored blanket development

    Energy Technology Data Exchange (ETDEWEB)

    Roth, E.S.

    1978-05-01

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

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

    International Nuclear Information System (INIS)

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

  3. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m2 and a particle heat flux of 1 MW/m2. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  4. Multiphysics Engineering Analysis for an Integrated Design of ITER Diagnostic First Wall and Diagnostic Shield Module Design

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Loesser, G. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Smith, M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Udintsev, V. [ITER Org, F-13115 St Paul Les Durance, France.; Giacomin, T., T. [ITER Org, F-13115 St Paul Les Durance, France.; Khodak, A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Johnson, D, [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Feder, R, [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    2015-07-01

    ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

  5. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-10-01

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

  7. Fast Breeder Blanket Facility FBBF. Annual report, January 1, 1981-December 31, 1981

    International Nuclear Information System (INIS)

    This annual report contains a summmary of fission rate, spectra, and gamma-ray heating rate measurements made in the first blanket of the Purdue Fast Breeder Blanket Facility. The first blanket consisted of aluminum clad, natural UO2 fuel rods with a secondary cladding of stainless steel or aluminum. The blanket was arranged in two concentric regions around the neutron source and converter regions. A neutron diffusion code, 2DB, and a Monte Carlo code, VIM, both using homogeneous cross section groups have been used to calculate the reaction rates. Calculated to experimental values for a number of important reactions are presented. A modified method of applying Bondarenko self-shielding factors to correct for the self shielding of resonance energy neutrons in aluminum, stainless steel and UO2 has improved the agreement between the calculations and experiment, but does not account for all of the differences

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  10. 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

  11. 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

  12. 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.)

  13. 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.)

  14. 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

  15. 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)

  16. 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.

  17. SHIELD verification and validation report

    Energy Technology Data Exchange (ETDEWEB)

    Boman, C.

    1992-02-01

    This document outlines the verification and validation effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system code. Along with its predecessors, SHIELD has been in use at the Savannah River Site (SRS) for more than ten years. During this time the code has been extensively tested and a variety of validation documents have been issued. The primary function of this report is to specify the features and capabilities for which SHIELD is to be considered validated, and to reference the documents that establish the validation.

  18. 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

  19. Small Engine Technology (SET) - Task 13 ANOPP Noise Prediction for Small Engines: Jet Noise Prediction Module, Wing Shielding Module, and System Studies Results

    Science.gov (United States)

    Lieber, Lysbeth; Golub, Robert (Technical Monitor)

    2000-01-01

    This Final Report has been prepared by AlliedSignal Engines and Systems, Phoenix, Arizona, documenting work performed during the period May 1997 through June 1999, under the Small Engines Technology Program, Contract No. NAS3-27483, Task Order 13, ANOPP Noise Prediction for Small Engines. The report specifically covers the work performed under Subtasks 4, 5 and 6. Subtask 4 describes the application of a semi-empirical procedure for jet noise prediction, subtask 5 describes the development of a procedure to predict the effects of wing shielding, and subtask 6 describes the results of system studies of the benefits of the new noise technology on business and regional aircraft.

  20. Shielding Effectiveness of Laminated Shields

    Directory of Open Access Journals (Sweden)

    B. P. Rao

    2008-12-01

    Full Text Available Shielding prevents coupling of undesired radiated electromagnetic energy into equipment otherwise susceptible to it. In view of this, some studies on shielding effectiveness of laminated shields with conductors and conductive polymers using plane-wave theory are carried out in this paper. The plane wave shielding effectiveness of new combination of these materials is evaluated as a function of frequency and thickness of material. Conductivity of the polymers, measured in previous investigations by the cavity perturbation technique, is used to compute the overall reflection and transmission coefficients of single and multiple layers of the polymers. With recent advances in synthesizing stable highly conductive polymers these lightweight mechanically strong materials appear to be viable alternatives to metals for EM1 shielding.

  1. Safety Analysis on Dual-functional Lithium Lead Test Blanket Module With RELAP5%基于 RELAP5的双功能液态锂铅实验包层模块安全分析

    Institute of Scientific and Technical Information of China (English)

    李伟; 田文喜; 秋穗正; 苏光辉

    2013-01-01

    利用嵌入了液态锂铅(LiPb)的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅(DFLL)实验包层模块(TBM)的安全特性进行评价。对DFLL-TBM 及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁(FW )结构材料表面最高温度低于允许值550℃。事故工况下氦气泄漏引起的ITER真空室(VV)、窗口设备室(port cell)以及托卡马克冷却水系统大厅拱顶(TCWS vault)的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM 的设计可满足ITER对其热工水力安全方面的要求。%Safety assessment on the dual-functional lithium lead test blanket module (DFLL-TBM) was performed with a modified version of RELAP5/MOD3 code in which the LiPb eutectic thermal-hydraulic sub-module was inserted .The DFLL-TBM and its ancillary cooling systems were modeled to conduct the computation and analysis for steady-state operation ,anticipated operational incidents and relevant accidents .Compu-tational results indicate that the maximum surface temperature of the first wall (FW) structural material is lower than the allowable value of 550 ℃ .For the accident analy-ses ,none of the pressure increases in ITER vacuum vessel (VV) ,port cell and TCWS vault induced by helium leaking is beyond the ITER safety limit of 0.2 MPa .No melting of the TBM box is found and the decay heat can be removed efficiently by the radiation heat transfer .With the current design ,DFLL-TBM can meet the thermal-hydraulic safety requirements from IT ER .

  2. Shielded syringe

    International Nuclear Information System (INIS)

    This patent specification relates to a partially disposable shielded syringe for injecting radioactive material into a patient. It is claimed that the technique overcomes the problems of non-standardisation of syringe size. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

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

  4. Numerical benchmarks TRIPOLI - MCNP with use of MCAM on FNG ITER bulk shield and FNG HCLL TBM mock-up experiments

    Energy Technology Data Exchange (ETDEWEB)

    Fausser, Clement, E-mail: clement.fausser@cea.fr [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Lee, Yi-Kang [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Zeng Qin; Zhang Junjun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Serikov, Arkady [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (Germany); Trama, Jean-Christophe; Gabriel, Franck [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France)

    2011-10-15

    3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium-lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI - MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI - MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.

  5. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

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

  6. Crucial issues on liquid metal blanket design

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-01

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

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

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

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

  8. RF-transparent solar shield

    International Nuclear Information System (INIS)

    By combining durable Kapton films with quartz fibers, an effective solar shield or blanket is produced which also serves as an efficient RF-transparent window. The window consists of a series of Kapton film envelopes sandwiching thin quartz paper. Not only must the window prevent the sun from overheating the electronics and distorting mechanically aligned antennas, it must also prevent radiant heat loss from inside the satellite when it is in shadow and radiating to space at approx. 40K. The guidelines for achieving an effective high-frequency RF window are a low dielectric constant to keep reflections down, a low loss tangent so RF absorption and molecular movement will be minimal, and low mass with tin and lightweight materials. Because these guidelines were followed, the RF insertion loss of the multiple envelope shield is less than 1/4 dB at high frequency. This paper concentrates on the material and processing aspects of an RF-transparent solar shield

  9. 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

  10. Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

    International Nuclear Information System (INIS)

    three dimensional Monte Carlo calculation is required for the shielding calculation in the tokamak-type DT nuclear fusion reactor with many penetrations. 2) In Chapter 3, radiation streaming through the slit between the blanket modules is described, in Chapter 4, that through the small circular duct in the blanket modules is described, in Chapter 5, and that through the large opening duct in the vacuum vessel is described. The nuclear properties of the blanket, the vacuum vessel and the TF coil are systematically calculated for the various configurations. Based on the obtained results, the analytical formulas of these nuclear properties are deduced, and the guideline is proposed for the shielding design. 3) In Chapter 6, in order to evaluate the decay gamma ray dose rate around the duct due to radiation streaming through the large opening duct in the vacuum vessel, the evaluation method is proposed using the decay gamma ray Monte Carlo calculation. By replacing the prompt gamma-ray spectrum to the decay one in the Monte Carlo code, the decay gamma ray Monte Carlo transport calculation is conducted. The effective variance reduction method is developed for the decay gamma ray Monte Carlo calculation in the over-all tokamak region with drastically reducing the calculation time. Using this method, the shielding calculation is conducted for the ITER duct penetration, and the effectiveness of this method is demonstrated. (author)

  11. Main maintenance operations for Test Blanket Systems in ITER TBM port cells

    International Nuclear Information System (INIS)

    Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues

  12. An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

    Science.gov (United States)

    Yu, Guanying; Liu, Xufeng; Liu, Songlin

    2016-10-01

    The Chinese Fusion Engineering Tokamak Reactor (CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder (WCCB) blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic (RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement. supported by National Natural Science Foundation of China (No. 11175207) and the National Magnetic Confinement Fusion Program of China (No. 2013GB108004)

  13. Design of ITER vacuum vessel in-wall shielding

    Energy Technology Data Exchange (ETDEWEB)

    Wang, X., E-mail: xiaoyu.wang@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Morimoto, M. [Mitsubishi Heavy Industries, 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe (Japan); Choi, C.H.; Utin, Y.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); TaiLhardat, O. [Assystem EOS, ZAC SAINT MARTIN, 23 rue Benjamin Franklin, 84120 Pertuis (France); Mille, B.; Terasawa, A.; Gribov, Y.; Barabash, V.; Polunovskiy, E.; Dani, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Pathak, H.; Raval, J. [ITER-India, Institute for Plasma Research, Gandhinagar 382025 (India); Liu, S.; Lu, M.; Du, S. [Institute of Plasma Physics, China Academy of Sciences, Shushanhu Road 350, Hefei (China)

    2014-10-15

    The ITER vacuum vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with in-wall shielding (IWS) and cooling water. The main purpose of the in-wall shielding is to provide neutron shielding together with the blanket and VV shells and water during ITER plasma operation and to reduce the ripple of the Toroidal magnetic field. Based on ITER vacuum vessel structure and related requirements, in-wall shielding are designed as about 8900 individual blocks with different sizes and several different materials distributed over nine vessel sectors and nine field joints of vessel sectors. This paper presents the design of the IWS, considering loads, structural stresses and assembly method, and also shows neutron shielding effect and TF ripple reduced by the IWS.

  14. Using the shield for thermal energy storage in pulsar

    Energy Technology Data Exchange (ETDEWEB)

    Sager, G.T. [General Atomics, San Diego, CA (United States); Sze, D.K. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Wong, C.P.C. [General Atomics, San Diego, CA (United States); Bathke, C.G. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Blanchard, J.P. [University of Wisconsin at Madison, Madison, WI 53706 (United States); Brimer, C. [McDonnell Douglas Aerospace, PO Box 516, Mail Code 3064204, St. Louis, MO 63166-0516 (United States); Cheng, E.T. [TSI Research Inc., 225 Stevens Avenue, Suite 203, Solana Beach, CA 92075 (United States); El-Guebaly, L.A. [University of Wisconsin at Madison, Madison, WI 53706 (United States); Hasan, M.Z. [University of California at Los Angeles, Los Angeles, CA 90024 (United States); Najmabadi, F. [University of California at San Diego, San Diego, CA 92037 (United States); Sharafat, S. [University of California at Los Angeles, Los Angeles, CA 90024 (United States); Sviatoslavski, I.N. [University of Wisconsin at Madison, Madison, WI 53706 (United States); Waganer, L. [McDonnell Douglas Aerospace, PO Box 516, Mail Code 3064204, St. Louis, MO 63166-0516 (United States)

    1995-03-01

    The PULSAR pulsed tokamak power plant design utilizes the outboard shield for thermal energy storage to maintain full 1000MW(e) output during the dwell period of 200s. Thermal energy resulting from direct nuclear heating is accumulated in the shield during the 7200s fusion power production phase. The maximum shield temperature may be much higher than that for the blanket because radiation damage is significantly reduced. During the dwell period, thermal power discharged from the shield and coolant temperature are simultaneously regulated by controlling the coolant mass flow rate at the shield inlet. This is facilitated by throttled coolant bypass. Design concepts using helium and lithium coolant have been developed. Two-dimensional time-dependent thermal hydraulic calculations were performed to confirm performance capabilities required of the design concepts. The results indicate that the system design and performance can accommodate uncertainties in material limits or the length of the dwell period. (orig.).

  15. Solar energy apparatus with apertured shield

    Science.gov (United States)

    Collings, Roger J. (Inventor); Bannon, David G. (Inventor)

    1989-01-01

    A protective apertured shield for use about an inlet to a solar apparatus which includesd a cavity receiver for absorbing concentrated solar energy. A rigid support truss assembly is fixed to the periphery of the inlet and projects radially inwardly therefrom to define a generally central aperture area through which solar radiation can pass into the cavity receiver. A non-structural, laminated blanket is spread over the rigid support truss in such a manner as to define an outer surface area and an inner surface area diverging radially outwardly from the central aperture area toward the periphery of the inlet. The outer surface area faces away from the inlet and the inner surface area faces toward the cavity receiver. The laminated blanket includes at least one layer of material, such as ceramic fiber fabric, having high infra-red emittance and low solar absorption properties, and another layer, such as metallic foil, of low infra-red emittance properties.

  16. Shielding requirements in helical tomotherapy

    International Nuclear Information System (INIS)

    Helical tomotherapy is a relatively new intensity-modulated radiation therapy (IMRT) treatment for which room shielding has to be reassessed for the following reasons. The beam-on-time needed to deliver a given target dose is increased and leads to a weekly workload of typically one order of magnitude higher than that for conventional radiation therapy. The special configuration of tomotherapy units does not allow the use of standard shielding calculation methods. A conventional linear accelerator must be shielded for primary, leakage and scatter photon radiations. For tomotherapy, primary radiation is no longer the main shielding issue since a beam stop is mounted on the gantry directly opposite the source. On the other hand, due to the longer irradiation time, the accelerator head leakage becomes a major concern. An analytical model based on geometric considerations has been developed to determine leakage radiation levels throughout the room for continuous gantry rotation. Compared to leakage radiation, scatter radiation is a minor contribution. Since tomotherapy units operate at a nominal energy of 6 MV, neutron production is negligible. This work proposes a synthetic and conservative model for calculating shielding requirements for the Hi-Art II TomoTherapy unit. Finally, the required concrete shielding thickness is given for different positions of interest

  17. 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

  18. Shield verification and validation action matrix summary

    Energy Technology Data Exchange (ETDEWEB)

    Boman, C.

    1992-02-01

    WSRC-RP-90-26, Certification Plan for Reactor Analysis Computer Codes, describes a series of action items to be completed for certification of reactor analysis computer codes used in Technical Specifications development and for other safety and production support calculations. Validation and verification are integral part of the certification process. This document identifies the work performed and documentation generated to satisfy these action items for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system, it is not certification of the complete SHIELD system. Complete certification will follow at a later date. Each action item is discussed with the justification for its completion. Specific details of the work performed are not included in this document but can be found in the references. The validation and verification effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system computer code is completed.

  19. Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

    International Nuclear Information System (INIS)

    Highlights: • Optimized water distribution in ITER blanket modules is presented. • All key challenging constraints are included. • The methodology and the successful result are presented. - Abstract: This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements. The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented

  20. 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

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  5. Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

    International Nuclear Information System (INIS)

    In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

  6. Divertor and gas blanket impurity control study

    Energy Technology Data Exchange (ETDEWEB)

    El Derini, Z; Stacey, Jr, W M

    1979-04-01

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

  7. 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.

  8. Two-dimensional TBR calculations for conceptual compact reversed-field pinch reactor blanket

    Science.gov (United States)

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

    A detailed two-dimensional nucleonic analysis was performed for a conceptual first wall, blanket, and shield design for the Compact Reversed-Field Pinch Reactor. The design includes significant two-dimensional aspects presented by the limiter, vacuum ducts, and coolant manifolds; these aspects seriously degrade the tritium-breeding reaction (TBR) predicted by one-dimensional calculations. A range of design change to increase the TBR were investigated within the two-dimensional analysis. The results of this investigation indicated that an adequate TBR could be achieved with a thinning copper first wall, a (6)Li enrichment near 90%, the proper selection of reflector, and a small addition to the blanket thickness, determined by the one-dimensional analysis.

  9. Two-dimensional TBR calculations for conceptual compact reversed-field pinch reactor blanket

    International Nuclear Information System (INIS)

    A detailed two-dimensional nucleonic analysis was performed for a conceptual first wall, blanket, and shield design for the Compact Reversed-Field Pinch Reactor. The design includes significant two-dimensional aspects presented by the limiter, vacuum ducts, and coolant manifolds; these aspects seriously degrade the tritium-breeding reaction (TBR) predicted by one-dimensional calculations. A range of design change to increase the TBR were investigated within the two-dimensional analysis. The results of this investigation indicated that an adequate TBR could be achieved with a thinner copper first wall, a 6Li enrichment near 90%, the proper selection of reflector, and a small addition to the blanket thickness, determined by the one-dimensional analysis

  10. 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.

  11. 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

  12. Physiologic implications of two different heat shields for infants under radiant warmers

    Energy Technology Data Exchange (ETDEWEB)

    Baumgart, S.; Fox, W.W.; Polin, R.A.

    1982-05-01

    To study the effects of two different heat shields commonly used under radiant warmers (thin plastic blanket vs a plastic body hood) on premature newborn infants, eight neonates were studied to measure insensible water loss and radiant power density at the skin under control and two shielded conditions. The body hood was found to diminish transmission of radiant heat to the infant's skin by 80%, absorbing heat and becoming warm (P less than 0.001). The plastic blanket absorbed only 15% of radiant warmer heat output (P less than 0.01). Insensible water loss was significantly less under the plastic blanket (1.88 +/- 0.4) ml/kg/hour) than under control (2.70 +/- 0.50 ml/kg/hour, P less than 0.01) and hood (2.86 +/- 0.32 ml/kg/hour, P less than 0.05) conditions. There was no decrease in insensible water loss under the hood compared to controls. This study demonstrates that a thin plastic blanket is the more effective shield against insensible water loss under a radiant warmer. Caution should be exercised with any shielding device to prevent interference with radiant heat delivery.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  15. Shielding analysis methods available in the scale computational system

    International Nuclear Information System (INIS)

    Computational tools have been included in the SCALE system to allow shielding analysis to be performed using both discrete-ordinates and Monte Carlo techniques. One-dimensional discrete ordinates analyses are performed with the XSDRNPM-S module, and point dose rates outside the shield are calculated with the XSDOSE module. Multidimensional analyses are performed with the MORSE-SGC/S Monte Carlo module. This paper will review the above modules and the four Shielding Analysis Sequences (SAS) developed for the SCALE system. 7 refs., 8 figs

  16. Handout on shielding calculation

    International Nuclear Information System (INIS)

    In order to avoid the difficulties of the radioprotection supervisors in the tasks related to shielding calculations, is presented in this paper the basic concepts of shielding theory. It also includes exercises and examples. (author)

  17. Magnetic shielding design analysis

    International Nuclear Information System (INIS)

    Two passive magnetic-shielding-design approaches for static external fields are reviewed. The first approach uses the shielding solutions for spheres and cylinders while the second approach requires solving Maxwell's equations. Experimental data taken at LLNL are compared with the results from these shieldings-design methods, and improvements are recommended for the second method. Design considerations are discussed here along with the importance of material gaps in the shield

  18. Electromagnetically shielded building

    International Nuclear Information System (INIS)

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs

  19. Study on the Influence of Intermittent Shielding on Power Generation of PV Module%间歇性遮挡对光伏组件发电量影响的研究

    Institute of Scientific and Technical Information of China (English)

    周磊; 刘翼; 雷涛; 童亦斌

    2012-01-01

    当光伏组件或者光伏阵列中部分光伏电池受到间歇性遮挡时,输出伏安特性曲线呈现阶梯状,对应的功率电压曲线包含多个局域峰值,导致光伏阵列发电效率下降,系统发电量降低.为了正确评估该类问题,通过工程案例重点研究和分析了不同遮挡情况下光伏电池输出特性,建立MATLAB池板单元级、方阵级遮挡的计算仿真模型,统计出由于组件间歇性遮挡带来的系统发电量的损失,文中的仿真统计结果与实际系统发电量进行了比较,验证了该方法的可行性.文中最后提出了利用分布式MPPT方法挽回遮挡对系统发电量的影响,通过系统监控得知,分布式MPPT最高可挽回15.52%的发电量,具有较高的工程应用价值.%When some PV cells of PV module or PV array are shielded intermittently, the output voltage-current curve has the feature of stepladder shape, and corresponding power-voltage curve contains multiple local peak values, which decrease efficiency and power generation of PV array. In order to evaluate the affected value properly, the I-U characteristic curve under different intermittent shielding is analyzed mainly through project case, and the simulation model of PV cell and PV array are built in MATLAB. Then the loss of power generation caused by intermittent shielding between components is calculated, and the power generation obtained by simulation is compared with that of practical system, which verifies the feasibility of the method. In the end, study shows that influence of shielding on system power generation will be overcome by distributed MPPT control method, and maximum salvage power generation will reach 15. 52% through system monitoring, which has high engineering applications value.

  20. 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.)

  1. 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.

  2. Rotating shielded crane system

    Science.gov (United States)

    Commander, John C.

    1988-01-01

    A rotating, radiation shielded crane system for use in a high radiation test cell, comprises a radiation shielding wall, a cylindrical ceiling made of radiation shielding material and a rotatable crane disposed above the ceiling. The ceiling rests on an annular ledge intergrally attached to the inner surface of the shielding wall. Removable plugs in the ceiling provide access for the crane from the top of the ceiling into the test cell. A seal is provided at the interface between the inner surface of the shielding wall and the ceiling.

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 47 CFR 22.353 - Blanketing interference.

    Science.gov (United States)

    2010-10-01

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

  9. 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

  10. 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.

  11. ITER driver blanket, European Community design

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

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

  12. Fidget Blankets: A Sensory Stimulation Outreach Program.

    Science.gov (United States)

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

    2016-01-01

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

  13. Robot vision system R and D for ITER blanket remote-handling system

    Energy Technology Data Exchange (ETDEWEB)

    Maruyama, Takahito, E-mail: maruyama.takahito@jaea.go.jp [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Tesini, Alessandro [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France)

    2014-10-15

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system.

  14. Robot vision system R and D for ITER blanket remote-handling system

    International Nuclear Information System (INIS)

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system

  15. James Webb Space Telescope (JWST) Integrated Science Instruments Module (ISIM) Electronics Compartment (IEC) Conformal Shields Composite Bond Structure Qualification Test Method

    Science.gov (United States)

    Yew, Calinda; Stephens, Matt

    2015-01-01

    The JWST IEC conformal shields are mounted onto a composite frame structure that must undergo qualification testing to satisfy mission assurance requirements. The composite frame segments are bonded together at the joints using epoxy, EA 9394. The development of a test method to verify the integrity of the bonded structure at its operating environment introduces challenges in terms of requirements definition and the attainment of success criteria. Even though protoflight thermal requirements were not achieved, the first attempt in exposing the structure to cryogenic operating conditions in a thermal vacuum environment resulted in approximately 1 bonded joints failure during mechanical pull tests performed at 1.25 times the flight loads. Failure analysis concluded that the failure mode was due to adhesive cracks that formed and propagated along stress concentrated fillets as a result of poor bond squeeze-out control during fabrication. Bond repairs were made and the structures successfully re-tested with an improved LN2 immersion test method to achieve protoflight thermal requirements.

  16. Under the Rape Shield

    OpenAIRE

    Roman, Denise

    2011-01-01

    This article focuses on the Rape Shield Laws and their evolution in the United States, one of the pioneers in this field. The article also discusses constitutional and feminist critiques of present Rape Shield Laws, and ends with a comparative perspective throughout the Anglo-American legal space today. Finally, although the Rape Shield Laws can be approached from a variety of discourses, this article engages specifically with a discourse that intersects legal and feminist analyses.

  17. 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

  18. 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

  19. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  20. 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

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

    Science.gov (United States)

    Dillard, P. A.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. 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...

  5. 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

  6. TO THE SUBSTANTIATION OF CALCULATION PRINCIPLES OF ROAD CONSTRUCTION STRENGTH HAVING BLANKET REGULATING WATER-THERMIC REGIME

    OpenAIRE

    Savenko, V.; Petrovich, V.; Chechuga, O.

    2005-01-01

    Principles of road construction strength with the blanket regulating water-thermal conditions are surveyed. The affect of dampness on a general module of elasticity as well as the basic activities which promote necessary frost resistance of the construction are considered. Synthetic fibre application is surveyed.

  7. Neutronics analysis of inboard shielding capability for a DEMO fusion reactor CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin; Li, Jiangang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Zheng, Shanliang, E-mail: shanliang.zheng@ccfe.ac.uk [Culham Centre for Fusion Energy, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Mitchell, Neil [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    The inboard shielding of a fusion reactor can be a crucial issue due to the limited space available in a tokamak configuration. It is necessary to assess the inboard shielding capability of DEMO for its initial design. In this paper, 1D and 3D neutronics models were developed based on a reference design of the Chinese Fusion Engineering Testing Reactor (CFETR). The neutron wall load (NWL) is in the range of 1.5–3 MW/m{sup 2} and the inboard shielding thickness is constrained within 40–70 cm in order to achieve the tritium self-sufficiency of the reactor. Referring to the detailed design of the ITER Toroidal Field Coils (TFCs) and using radiation hardening technology developed for ITER, the inboard blanket shielding capability and nuclear responses of the TFC are investigated for both FLiBe and Li{sub 4}SiO{sub 4} breeding blanket concepts. The impact of the gaps on shielding performance is discussed. Some suggestions on improving the inboard shielding performance for DEMO are also proposed.

  8. 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.

  9. 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)

  10. A Precambrian proximal ejecta blanket from Scotland

    Science.gov (United States)

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

    2008-04-01

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

  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. 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 ...

  13. 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

  14. 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.

  15. 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.

  16. TCT hybrid preconceptual blanket design studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-01-01

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

  17. 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

  18. Scale-PC shielding analysis sequences

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    1996-05-01

    The SCALE computational system is a modular code system for analyses of nuclear fuel facility and package designs. With the release of SCALE-PC Version 4.3, the radiation shielding analysis community now has the capability to execute the SCALE shielding analysis sequences contained in the control modules SAS1, SAS2, SAS3, and SAS4 on a MS- DOS personal computer (PC). In addition, SCALE-PC includes two new sequences, QADS and ORIGEN-ARP. The capabilities of each sequence are presented, along with example applications.

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

    International Nuclear Information System (INIS)

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

  20. Radiation shielding device

    International Nuclear Information System (INIS)

    Purpose: To lower the shielding cost by providing a shielding wall having cavities and charging spherical shiedling materials in the cavities only when the shielding is required. Constitution: The structure comprises two parallel steel side plates aparting from each other to form a space therebetween and reinforcements such as H-type steels vertically provided between the side plates. The upper and the lower ends of the reinforcements are aparted from the upper and the lower edges of the side plates by a predetermined distance to form lateral passage between the top plate and the bottom plate. A guide plate having a plurality of openings is mounted on the upper ends of the reinforcements. If it is required for the structure to serve as the shield, spherical radioactive shielding materials are supplied through an injection port onto the guide plate while opening the injection port is opened and closing discharge port. The spherical radioactive shielding materials are fallen through the openings and filled in the space to thereby providing the structure with shielding performance. (Yoshino, Y.)

  1. 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

  2. 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

  3. Shielding against galactic cosmic rays

    Science.gov (United States)

    Schimmerling, W.; Wilson, J. W.; Nealy, J. E.; Thibeault, S. A.; Cucinotta, F. A.; Shinn, J. L.; Kim, M.; Kiefer, R.

    1996-01-01

    Ions of galactic origin are modified but not attenuated by the presence of shielding materials. Indeed, the number of particles and the absorbed energy behind most shield materials increases as a function of shield thickness. The modification of the galactic cosmic ray composition upon interaction with shielding is the only effective means of providing astronaut protection. This modification is intimately conntected with the shield transport porperties and is a strong function of shield composition. The systematic behavior of the shield properites in terms of microscopic energy absorption events will be discussed. The shield effectiveness is examined with respect to convectional protection practice and in terms of a biological endpoint: the efficiency for reduction of the probability of transformation of shielded C3H1OT1/2 mouse cells. The relative advantage of developing new shielding technologies is discussed in terms of a shield performance as related to biological effect and the resulting uncertainty in estimating astronaut risk.

  4. 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

  5. iSHIELD - A Line Source Application of SHIELD11

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, W.R.; Rokni, S.H.; /SLAC

    2006-04-27

    iSHIELD11 performs a line-source numerical integration of radiation source terms that are defined by the iSHIELD11 computer code[1] . An example is provided to demonstrate how one can use iSHIELD11 to perform a shielding analysis for a 250 GeV electron linear accelerator.

  6. 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...

  7. 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

  8. Scintillation counter, segmented shield

    International Nuclear Information System (INIS)

    A scintillation counter, particularly for counting gamma ray photons, includes a massive lead radiation shield surrounding a sample-receiving zone. The shield is disassembleable into a plurality of segments to allow facile installation and removal of a photomultiplier tube assembly, the segments being so constructed as to prevent straight-line access of external radiation through the shield into radiation-responsive areas. Provisions are made for accurately aligning the photomultiplier tube with respect to one or more sample-transmitting bores extending through the shield to the sample receiving zone. A sample elevator, used in transporting samples into the zone, is designed to provide a maximum gamma-receiving aspect to maximize the gamma detecting efficiency. (U.S.)

  9. Alternate shield material feasibility

    Energy Technology Data Exchange (ETDEWEB)

    Specht, E.R.; Levitt, L.B.

    1984-04-01

    The feasibility and cost/benefit of using materials other than stainless steel for in-vessel neutron shielding in large LMFBRs were investigated. Canned vibratorally compacted B/sub 4/C powder shields were found to be much more economical than stainless steel (a savings of $1.1M in loop plant designs and $9.4M in pool plant designs). The helium gas pressure buildup in B/sub 4/C shields placed around LMFBR in-vessel components (direct reactor heat exchangers in a loop reactor and intermediate heat exchangers in a pool reactor) would only be 0.04 atm after 40 y of reactor operation (with 80% dense powder). The irradiation-induced swelling of the B/sub 4/C would only be 0.002%. No adverse reactor impact would occur if the B/sub 4/C escaped from the B/sub 4/C shields.

  10. Alternate shield material feasibility

    International Nuclear Information System (INIS)

    The feasibility and cost/benefit of using materials other than stainless steel for in-vessel neutron shielding in large LMFBRs were investigated. Canned vibratorally compacted B4C powder shields were found to be much more economical than stainless steel (a savings of $1.1M in loop plant designs and $9.4M in pool plant designs). The helium gas pressure buildup in B4C shields placed around LMFBR in-vessel components (direct reactor heat exchangers in a loop reactor and intermediate heat exchangers in a pool reactor) would only be 0.04 atm after 40 y of reactor operation (with 80% dense powder). The irradiation-induced swelling of the B4C would only be 0.002%. No adverse reactor impact would occur if the B4C escaped from the B4C shields

  11. Radiation shielding curtain

    International Nuclear Information System (INIS)

    A radiation shield is described in the form of a stranded curtain made up of bead-chains whose material and geometry are selected to produce a cross-sectional density that is the equivalent of 0.25 mm or more of lead and which curtain may be mounted on various radiological devices to shield against scattered radiation while offering a minimum of obstruction to the radiologist

  12. Evidence for an extensive Phanerozoic sediment cover on the Canadian and Fenno-Scandian shields

    Energy Technology Data Exchange (ETDEWEB)

    Laine, E.P.; Dickson, S.M.

    1985-01-01

    Examination of the age and diameter of 75 terrestrial meteorite impact craters taken from platform and shield regions throughout the world suggest that both the Canadian and Fenno-Scandian Shields were covered by a sedimentary blanket during a portion of the Phanerozoic. Subsequent erosion, fostered perhaps by a combination of glacial and tectonic processes, has exposed both of these shields to reveal an anomalous distribution of craters through time. The primary evidence for sedimentary cover and subsequent erosion is in the form of a 280 Myr gap in the record of craters less than 15 km in diameter. Small craters of Cambrian, Ordovician and Silurian age are found in shield regions, suggesting either a thin or non-existent sediment cover during this period. However, there is no record of small diameter craters on either shield of Devonian, Carboniferous, Permian, Triassic, or Jurassic age (400 to 120 Myr). This 280 Myr gap suggests that the shields were protected from smaller body impacts by a sedimentary cover. In contrast, the record of impacts on platform sediments implies no such hiatus in the infall of cosmic bodies to the earth's surface between the Devonian and the Early Cretaceous. Subsequent erosion, perhaps by Early Cretaceous time, exposed the shields to further bombardment. In addition, pre-Devonian craters became exhumed. Thus, the record of impact craters suggests that the Canadian and Fenno-Scandian Shields were covered by sediments while part of Pangaea.

  13. Thermal, Radiation and Impact Protective Shields (TRIPS) for Robotic and Human Space Exploration Missions

    Science.gov (United States)

    Loomis, M. P.; Arnold, J. L.

    2005-01-01

    New concepts for protective shields for NASA s Crew Exploration Vehicles (CEVs) and planetary probes offer improved mission safety and affordability. Hazards include radiation from cosmic rays and solar particle events, hypervelocity impacts from orbital debris/ micrometeorites, and the extreme heating environment experienced during entry into planetary atmospheres. The traditional approach for the design of protection systems for these hazards has been to create single-function shields, i.e. ablative and blanket-based heat shields for thermal protection systems (TPS), polymer or other low-molecular-weight materials for radiation shields, and multilayer, Whipple-type shields for protection from hypervelocity impacts. This paper introduces an approach for the development of a single, multifunctional protective shield, employing nanotechnology- based materials, to serve simultaneously as a TPS, an impact shield and as the first line of defense against radiation. The approach is first to choose low molecular weight ablative TPS materials, (existing and planned for development) and add functionalized carbon nanotubes. Together they provide both thermal and radiation (TR) shielding. Next, impact protection (IP) is furnished through a tough skin, consisting of hard, ceramic outer layers (to fracture the impactor) and sublayers of tough, nanostructured fabrics to contain the debris cloud from the impactor before it can penetrate the spacecraft s interior.

  14. 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.

  15. Preliminary Thermal Design of Cryogenic Radiation Shielding

    Science.gov (United States)

    Li, Xiaoyi; Mustafi, Shuvo; Boutte, Alvin

    2015-01-01

    Cryogenic Hydrogen Radiation Shielding (CHRS) is the most mass efficient material radiation shielding strategy for human spaceflight beyond low Earth orbit (LEO). Future human space flight, mission beyond LEO could exceed one year in duration. Previous radiation studies showed that in order to protect the astronauts from space radiation with an annual allowable radiation dose less than 500 mSv, 140 kgm2 of polyethylene is necessary. For a typical crew module that is 4 meter in diameter and 8 meter in length. The mass of polyethylene radiation shielding required would be more than 17,500 kg. The same radiation study found that the required hydrogen shielding for the same allowable radiation dose is 40 kgm2, and the mass of hydrogen required would be 5, 000 kg. Cryogenic hydrogen has higher densities and can be stored in relatively small containment vessels. However, the CHRS system needs a sophisticated thermal system which prevents the cryogenic hydrogen from evaporating during the mission. This study designed a cryogenic thermal system that protects the CHRS from hydrogen evaporation for one to up to three year mission. The design also includes a ground based cooling system that can subcool and freeze liquid hydrogen. The final results show that the CHRS with its required thermal protection system is nearly half of the mass of polyethylene radiation shielding.

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

    International Nuclear Information System (INIS)

    This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared

  17. An assessment of the base blanket for ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-31

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

  18. An assessment of the base blanket for ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  19. 75 FR 51482 - Woven Electric Blankets From China

    Science.gov (United States)

    2010-08-20

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

  20. Mechanical shielded hot cell

    International Nuclear Information System (INIS)

    A plan to erect a mechanical shielded hot cell in the process hall of the Radiochemical Laboratory at Inchas is described. The hot cell is designed for safe handling of spent fuel bundles, from the Inchas reactor, and for dismantling and cutting the fuel rods in preparation for subsequent treatment. The biological shielding allows for the safe handling of a total radioactivity level up to 10,000 MeV-Ci. The hot cell consists of an α-tight stainless-steel box, connected to a γ-shielded SAS, through an air-lock containing a movable carriage. The α-box is tightly connected with six dry-storage cavities for adequate storage of the spent fuel bundles. Both the α-box, with the dry-storage cavities, and the SAS are surrounded by 200-mm thick biological lead shielding. The α-box is equipped with two master-slave manipulators, a lead-glass window, a monorail crane and Padirac and Minirag systems. The SAS is equipped with a lead-glass window, tong manipulator, a shielded pit and a mechanism for the entry of the spent fuel bundle. The hot cell is served by adequate ventilation and monitoring systems. (author)

  1. Neutron shielding material

    International Nuclear Information System (INIS)

    From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

  2. LMFBR Blanket Physics Project progress report No. 6

    International Nuclear Information System (INIS)

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

  3. LMFBR Blanket Physics Project progress report No. 6

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J. (ed.)

    1975-06-30

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

  4. MIT LMFBR blanket research project. Final summary report

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J.

    1983-08-01

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

  5. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

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

  6. Designer's guidebook for first wall/blanket/shield assembly, maintenance, and repair

    International Nuclear Information System (INIS)

    This is the initial issue of the guidebook. Since a guidebook of this type must incorporate information concerning a wide range of subjects, much additional data will eventually be included. The guidebook will document, in summary and easily referenceable form, data, designs, design concepts, design guidelines and background information useful to the FWBS and to the Maintenance System designer. In providing guidelines for the AMR of the FWBS, the guidebook must, of necessity, include guidelines for all aspects of maintenance associated with the FWBS. These include most maintenance operations within the reactor room necessary to gain access, identify faults, and handle equipment related to FWBS maintenance. In addition, the guidelines include those required to define facility requirements for handling and repair of FWBS and related reactor components external to the reactor room. Particular emphasis is given to remote maintenance design and operations

  7. Designer's guidebook for first wall/blanket/shield assembly, maintenance, and repair

    Energy Technology Data Exchange (ETDEWEB)

    1983-12-30

    This is the initial issue of the guidebook. Since a guidebook of this type must incorporate information concerning a wide range of subjects, much additional data will eventually be included. The guidebook will document, in summary and easily referenceable form, data, designs, design concepts, design guidelines and background information useful to the FWBS and to the Maintenance System designer. In providing guidelines for the AMR of the FWBS, the guidebook must, of necessity, include guidelines for all aspects of maintenance associated with the FWBS. These include most maintenance operations within the reactor room necessary to gain access, identify faults, and handle equipment related to FWBS maintenance. In addition, the guidelines include those required to define facility requirements for handling and repair of FWBS and related reactor components external to the reactor room. Particular emphasis is given to remote maintenance design and operations.

  8. Shields-1, A SmallSat Radiation Shielding Technology Demonstration

    Science.gov (United States)

    Thomsen, D. Laurence, III; Kim, Wousik; Cutler, James W.

    2015-01-01

    The NASA Langley Research Center Shields CubeSat initiative is to develop a configurable platform that would allow lower cost access to Space for materials durability experiments, and to foster a pathway for both emerging and commercial-off-the-shelf (COTS) radiation shielding technologies to gain spaceflight heritage in a relevant environment. The Shields-1 will be Langleys' first CubeSat platform to carry out this mission. Radiation shielding tests on Shields-1 are planned for the expected severe radiation environment in a geotransfer orbit (GTO), where advertised commercial rideshare opportunities and CubeSat missions exist, such as Exploration Mission 1 (EM-1). To meet this objective, atomic number (Z) graded radiation shields (Zshields) have been developed. The Z-shield properties have been estimated, using the Space Environment Information System (SPENVIS) radiation shielding computational modeling, to have 30% increased shielding effectiveness of electrons, at half the thickness of a corresponding single layer of aluminum. The Shields-1 research payload will be made with the Z-graded radiation shields of varying thicknesses to create dose-depth curves to be compared with baseline materials. Additionally, Shields-1 demonstrates an engineered Z-grade radiation shielding vault protecting the systems' electronic boards. The radiation shielding materials' performances will be characterized using total ionizing dose sensors. Completion of these experiments is expected to raise the technology readiness levels (TRLs) of the tested atomic number (Z) graded materials. The most significant contribution of the Z-shields for the SmallSat community will be that it enables cost effective shielding for small satellite systems, with significant volume constraints, while increasing the operational lifetime of ionizing radiation sensitive components. These results are anticipated to increase the development of CubeSat hardware design for increased mission lifetimes, and enable

  9. Radiolysis Experiments for the Aqueous Self-Cooled Blanket. Final report

    International Nuclear Information System (INIS)

    The results of Fusion Technology Task NAB 1.1 (Radiolytic Experiments for the ASCB), are reported . In the Aqueous Self-Cooled Blanket (ASCB) concept, an aqueous 6Li solution in a metallic structure is used as a shielding-breeding blanket for fusion reactors. Radiolysis could be very important with respect to the design and the use of an ASCB. The objectives of this project were to quantify the radiolytic decomposition of neutron irradiated aqueous lithium solutions and to demonstrate, if possible, the suppression of this decomposition by the initial addition of a small amount of hydrogen. Closed capsules, with the solutions and an inert gas or hydrogen as cover gas, were irradiated with thermal neutrons in a fission reactor. Radiolysis products hydrogen and oxygen (from hydrogen peroxide) as well as tritium were measured after irradiation. Tritium served as an internal dosimeter. The experimental results with LiNO3 , Li2SO4 and LiOH solutions indicate that the radiolytic gas production in an ASCB is proportional to the absorbed radiation energy. The observed radiation chemical yields allow the preliminary estimation of the radiolysis effects for a specific ASCB design. Contrary to the theoretical predictions, the use of hydrogen as a cover gas at up to 1 MPa had no measurable effect on the radiolytic gas production. Probably it will thus not be possible to suppress the radiolytic decomposition of a low-pressure ASCB by the addition of H2. Catalytic recombination will be required

  10. Analysis of the thorium axial blanket experiments in the proteus reactor

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.; Ingersoll, D.T.

    1980-12-01

    Detailed analysis has been completed for the ThO/sub 2/ and Th-metal axial blanket experiments performed at the Swiss PROTEUS critical facility in order to compare reaction rates and neutron spectra measured in prototypic GCFR configurations with calculated results. The PROTEUS configurations allowed the analysis of infinitely dilute thorium data in a PuO/sub 2//UO/sub 2/ fast lattice spectrum at core center as well as the analysis of resonance self-shielding effects in the thorium-bearing axial blankets. These comparisons indicate that significant deficiencies still exist in the latest evaluated infinitely dilute thorium data file. Specifically, the analysis showed that the /sup 232/Th capture is underpredicted by ENDF/B-IV data, and the discrepancies are further exaggerated by ENDF/B-V data. On the other hand, ENDF/B-V /sup 232/Th fission data appear to be significantly improved relative to ENDF/B-IV data, while discrepancies are extremely large for the (n,2n) process in both data files. Finally, the (n,n') cross sections for thorium also appear improved in ENDF/B-V, except for a small energy range just above the 50 keV threshold. Therefore, these combined data deficiencies suggest that relatively large uncertainties should be associated with many of the results obtained from recent fast reactor alternate fuel cycle analyses. 38 figures, 12 tables.

  11. Advanced Multifunctional MMOD Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — MMOD toughened, smart thermal blankets. The project was successful over the past two years in developing and demonstrating by test (hypervelocity impact and thermal...

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

    Science.gov (United States)

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

    2016-02-01

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

  13. Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. 12-month progress report, July 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    This report presents the conceptual design and preliminary feasibility assessment for the hybrid blanket and power conversion system of the Mirror Hybrid Fusion-Fission Reactor. Existing gas-cooled fission reactor technology is directly applicable to the Mirror Hybrid Reactor. There are a number of aspects of the present conceptual design that require further design and analysis effort. The blanket and power conversion system operating parameters have not been optimized. The method of supporting the blanket modules and the interface between these modules and the primary loop helium ducting will require further design work. The means of support and containment of the primary loop components must be studied. Nevertheless, in general, the conceptual design appears quite feasible

  14. Simulation of sludge blanket height in clarifiers

    Institute of Scientific and Technical Information of China (English)

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

    2009-01-01

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

  15. Shield For Flexible Pipe

    Science.gov (United States)

    Ponton, Michael K.; Williford, Clifford B.; Lagen, Nicholas T.

    1995-01-01

    Cylindrical shield designed to fit around flexible pipe to protect nearby workers from injury and equipment from damage if pipe ruptures. Designed as pressure-relief device. Absorbs impact of debris ejected radially from broken flexible pipe. Also redirects flow of pressurized fluid escaping from broken pipe onto flow path allowing for relief of pressure while minimizing potential for harm.

  16. Radiation shielding materials

    International Nuclear Information System (INIS)

    Purpose: To obtain putty-like shielding materials excellent in the radiation shielding and packing workability for use in penetrations of electrical wires or pipeways in a nuclear installation. Constitution: A putty-like material is prepared from 100 parts by weight of a binder comprising a grease or the like having viscosity of greater than 5000 cst or an immiscible consistency of greater than 100 (JIS K 2220 (1980) para. 5.3.4) at 25 0C and from 1200 to 4000 parts by weight of high density inorganic powder such as lead powder or lead oxide powder having a density of greater than 5 g/cm3 and such a particle size that more than 95 % thereof passes through a 145 mesh sieve. The putty-like material is adjusted such that it has 1 - 35 mm of softness (JIS A 5752) at normal temperature, more than 1 g/5 sec of injection amount and a density of greater than 4 g/cm3. In this way, non-curable radiation shielding agent with excellent X-ray or γ-ray shielding property and being capable of packed densely to void portions can be obtained. (Ikeda, J.)

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

    Science.gov (United States)

    Scialdone, John J.

    1990-01-01

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

  18. Hinged Shields for Machine Tools

    Science.gov (United States)

    Lallande, J. B.; Poland, W. W.; Tull, S.

    1985-01-01

    Flaps guard against flying chips, but fold away for tool setup. Clear plastic shield in position to intercept flying chips from machine tool and retracted to give operator access to workpiece. Machine shops readily make such shields for own use.

  19. Analysis of the ORNL/TSF GCFR Grid-Plate Shield Design Confirmation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.; Cramer, S.N.; Ingersoll, D.T.

    1979-08-01

    The results of the analysis of the GCFR Grid-Plate Shield Design Confirmation Experiment are presented. The experiment, performed at the ORNL Tower Shielding Facility, was designed to test the adequacy of methods and data used in the analysis of the GCFR design. In particular, the experiment tested the adequacy of methods to calculate: (1) axial neutron streaming in the GCFR core and axial blanket, (2) the amount and location of the maximum fast-neutron exposure to the grid plate, and (3) the neutron source leaving the top of the grid plate and entering the upper plenum. Other objectives of the experiment were to verify the grid-plate shielding effectiveness and to assess the effects of fuel-pin and subassembly spacing on radiation levels in the GCFR. The experimental mockups contained regions representing the GCFR core/blanket region, the grid-plate shield section, and the grid plate. Most core design options were covered by allowing: (1) three different spacings between fuel subassemblies, (2) two different void fractions within a subassembly by variation of the number of fuel pins, and (3) a mockup of a control-rod channel.

  20. Analysis of the ORNL/TSF GCFR Grid-Plate Shield Design Confirmation Experiment

    International Nuclear Information System (INIS)

    The results of the analysis of the GCFR Grid-Plate Shield Design Confirmation Experiment are presented. The experiment, performed at the ORNL Tower Shielding Facility, was designed to test the adequacy of methods and data used in the analysis of the GCFR design. In particular, the experiment tested the adequacy of methods to calculate: (1) axial neutron streaming in the GCFR core and axial blanket, (2) the amount and location of the maximum fast-neutron exposure to the grid plate, and (3) the neutron source leaving the top of the grid plate and entering the upper plenum. Other objectives of the experiment were to verify the grid-plate shielding effectiveness and to assess the effects of fuel-pin and subassembly spacing on radiation levels in the GCFR. The experimental mockups contained regions representing the GCFR core/blanket region, the grid-plate shield section, and the grid plate. Most core design options were covered by allowing: (1) three different spacings between fuel subassemblies, (2) two different void fractions within a subassembly by variation of the number of fuel pins, and (3) a mockup of a control-rod channel

  1. Spacecraft Electrostatic Radiation Shielding

    Science.gov (United States)

    2008-01-01

    This project analyzed the feasibility of placing an electrostatic field around a spacecraft to provide a shield against radiation. The concept was originally proposed in the 1960s and tested on a spacecraft by the Soviet Union in the 1970s. Such tests and analyses showed that this concept is not only feasible but operational. The problem though is that most of this work was aimed at protection from 10- to 100-MeV radiation. We now appreciate that the real problem is 1- to 2-GeV radiation. So, the question is one of scaling, in both energy and size. Can electrostatic shielding be made to work at these high energy levels and can it protect an entire vehicle? After significant analysis and consideration, an electrostatic shield configuration was proposed. The selected architecture was a torus, charged to a high negative voltage, surrounding the vehicle, and a set of positively charged spheres. Van de Graaff generators were proposed as the mechanism to move charge from the vehicle to the torus to generate the fields necessary to protect the spacecraft. This design minimized complexity, residual charge, and structural forces and resolved several concerns raised during the internal critical review. But, it still is not clear if such a system is costeffective or feasible, even though several studies have indicated usefulness for radiation protection at energies lower than that of the galactic cosmic rays. Constructing such a system will require power supplies that can generate voltages 10 times that of the state of the art. Of more concern is the difficulty of maintaining the proper net charge on the entire structure and ensuring that its interaction with solar wind will not cause rapid discharge. Yet, if these concerns can be resolved, such a scheme may provide significant radiation shielding to future vehicles, without the excessive weight or complexity of other active shielding techniques.

  2. INDRA: a program system for calculating the neutronics and photonics characteristics of a fusion reactor blanket

    International Nuclear Information System (INIS)

    INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU)

  3. 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

  4. Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm.

  5. Capacitive Proximity Sensors With Additional Driven Shields

    Science.gov (United States)

    Mcconnell, Robert L.

    1993-01-01

    Improved capacitive proximity sensors constructed by incorporating one or more additional driven shield(s). Sensitivity and range of sensor altered by adjusting driving signal(s) applied to shield(s). Includes sensing electrode and driven isolating shield that correspond to sensing electrode and driven shield.

  6. 18 CFR 284.402 - Blanket marketing certificates.

    Science.gov (United States)

    2010-04-01

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

  7. 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

  8. 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...

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

    Data.gov (United States)

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

  10. Combustor bulkhead heat shield assembly

    Energy Technology Data Exchange (ETDEWEB)

    Zeisser, M.H.

    1990-06-19

    This paper describes a gas turbine engine having an annular combustion chamber defined by an annular, inner liner, a concentric outer liner, and an upstream annular combustor head, wherein the head includes a radially extending bulkhead having circumferentially distributed openings for each receiving an individual fuel nozzle therethrough. It comprises: a segmented heat shield assembly, disposed between the combustion chamber interior and the bulkhead, including generally planar, sector shaped heat shields, each shield abutting circumferentially with two next adjacent shields and extending radially from proximate the inner liner to proximate the outer liner, the plurality of shields collectively defining an annular protective barrier, and wherein each sector shaped shield further includes an opening, corresponding to one of the bulkhead nozzle openings for likewise receiving the corresponding nozzle therethrough, the shield opening further including an annular lip extending toward the bulkhead and being received within the bulkhead opening, raised ridges on the shield backside, the ridges contacting the facing bulkhead surface and defining a flow path for a flow of cooling air issuing from a sized supply opening disposed in the bulkhead, the flow path running ultimately from adjacent the annular lip to the edges of each shield segment, wherein the raised edges extend fully along the lateral, circumferentially spaced edges of each shield segment and about the adjacent shield segments wherein the raised ridges further extend circumferentially between the annular lip and the abutting edge ridges.

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

    International Nuclear Information System (INIS)

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

  12. Impact of prescribed burning on blanket peat hydrology

    OpenAIRE

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

    2015-01-01

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

  13. 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.)

  14. Axial blanket for 16NGF Angra 1 fuel type

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  15. Radioisotope Power System Facility shielding analysis

    International Nuclear Information System (INIS)

    A series of calculations for the Radioisotope Power System Facility have been performed. These analyses have determined the shielding required for storage, testing, and transport of 238Pu heat source modules using the Monte Carlo code MCNP3B. The source terms and the assumptions used have been verified by comparison of calculated dose rates with measured ones. This paper describes the methodology used for shielding designs and the utilization of available variance reduction techniques to improve the computational efficiency. The new version of MCNP (MCNP3B) with a repeated structure capability was used. It decreased the chance for computer model errors and greatly decreased the model setup time. 2 refs., 3 figs., 2 tabs

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

    International Nuclear Information System (INIS)

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

  17. 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

  18. Current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

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

  19. 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

  20. 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

  1. A Ballistic Limit Analysis Program for Shielding Against Micrometeoroids and Orbital Debris

    Science.gov (United States)

    Ryan, Shannon; Christiansen, Erie

    2010-01-01

    A software program has been developed that enables the user to quickly and simply perform ballistic limit calculations for common spacecraft structures that are subject to hypervelocity impact of micrometeoroid and orbital debris (MMOD) projectiles. This analysis program consists of two core modules: design, and; performance. The design module enables a user to calculate preliminary dimensions of a shield configuration (e.g., thicknesses/areal densities, spacing, etc.) for a ?design? particle (diameter, density, impact velocity, incidence). The performance module enables a more detailed shielding analysis, providing the performance of a user-defined shielding configuration over the range of relevant in-orbit impact conditions.

  2. A contribution to shielding effectiveness analysis of shielded tents

    Directory of Open Access Journals (Sweden)

    Vranić Zoran M.

    2004-01-01

    Full Text Available An analysis of shielding effectiveness (SE of the shielded tents made of the metallised fabrics is given. First, two electromagnetic characteristic fundamental for coupling through electrically thin shield, the skin depth break frequency and the surface resistance or transfer impedance, is defined and analyzed. Then, the transfer function and the SE are analyzed regarding to the frequency range of interest to the Electromagnetic Compatibility (EMC Community.

  3. SHIELD 1.0: development of a shielding calculator program in diagnostic radiology; SHIELD 1.0: desenvolvimento de um programa de calculo de blindagem em radiodiagnostico

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Romulo R.; Real, Jessica V.; Luz, Renata M. da [Hospital Sao Lucas (PUCRS), Porto Alegre, RS (Brazil); Friedrich, Barbara Q.; Silva, Ana Maria Marques da, E-mail: ana.marques@pucrs.br [Pontificia Universidade Catolica do Rio Grande do Sul (PUCRS), Porto Alegre, RS (Brazil)

    2013-08-15

    In shielding calculation of radiological facilities, several parameters are required, such as occupancy, use factor, number of patients, source-barrier distance, area type (controlled and uncontrolled), radiation (primary or secondary) and material used in the barrier. The shielding design optimization requires a review of several options about the physical facility design and, mainly, the achievement of the best cost-benefit relationship for the shielding material. To facilitate the development of this kind of design, a program to calculate the shielding in diagnostic radiology was implemented, based on data and limits established by National Council on Radiation Protection and Measurements (NCRP) 147 and SVS-MS 453/98. The program was developed in C⌗ language, and presents a graphical interface for user data input and reporting capabilities. The module initially implemented, called SHIELD 1.0, refers to calculating barriers for conventional X-ray rooms. The program validation was performed by the comparison with the results of examples of shielding calculations presented in NCRP 147.

  4. Measurement of the transient shielding effectiveness of shielding cabinets

    Directory of Open Access Journals (Sweden)

    H. Herlemann

    2008-05-01

    Full Text Available Recently, new definitions of shielding effectiveness (SE for high-frequency and transient electromagnetic fields were introduced by Klinkenbusch (2005. Analytical results were shown for closed as well as for non closed cylindrical shields. In the present work, the shielding performance of different shielding cabinets is investigated by means of numerical simulations and measurements inside a fully anechoic chamber and a GTEM-cell. For the GTEM-cell-measurements, a downscaled model of the shielding cabinet is used. For the simulations, the numerical tools CONCEPT II and COMSOL MULTIPHYSICS were available. The numerical results agree well with the measurements. They can be used to interpret the behaviour of the shielding effectiveness of enclosures as function of frequency. From the measurement of the electric and magnetic fields with and without the enclosure in place, the electric and magnetic shielding effectiveness as well as the transient shielding effectiveness of the enclosure are calculated. The transient SE of four different shielding cabinets is determined and discussed.

  5. Light weight polarized polypropylene foam for noise shielding

    Science.gov (United States)

    Zelfer, Travis J.; Warne, Derik S.; Korde, Umesh A.

    2009-03-01

    The high levels of noise generated during launch can destroy sensitive equipment on space craft. Passive damping systems, like acoustic blankets, work to reduce the high frequency noise but do little to the low frequency noise (foams with high piezoelectric coupling constants are being used as new types of actuators and sensors. Further impedance control through the inverse piezoelectric effect will lead to a new "semi-active" approach that will reduce low frequency noise levels. Combining layers of conventional nonpiezoelectric foam and ferroelectret materials with a multiple loop feedback system will give a total damping effect that is adaptable over a wide band of low frequencies. This paper covers the manufacturing methods that were used to make polarized polypropylene foam, to test the foam for its polarized response and its noise shielding ability.

  6. Mechanical characteristics and position control of vehicle/manipulator for ITER blanket remote maintenance

    International Nuclear Information System (INIS)

    In International Thermonuclear Experimental Reactor (ITER), blanket maintenance requires the 4-tonne module handling with high positioning accuracy of ±2 mm. In order to meet this requirement, it is essential to suppress the dynamic deflection and vibration of the remote handling equipment due to sudden transfer of the module weight from/to the back-plate supports to/from the equipment itself during installation and removal. A new control scheme was proposed and tested so as to suppress the dynamic behaviors. As a result, the dynamic deflection of the rail and the acceleration of the manipulator were successfully decreased to nearly zero. Based on the test results, the proposed control scheme was concluded to be effective so as to suppress this kind of dynamic effect during heavy component handling

  7. On an optimized neutron shielding for an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    The molten salt reactor technology has gained renewed interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner core vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all core internal structures. On the basis of this new geometry a model for neutron physics calculation is presented and applied for a shielding optimization. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system has to be significantly increased and will finally be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem. (author)

  8. New Toroid shielding design

    CERN Multimedia

    Hedberg V

    On the 15th of June 2001 the EB approved a new conceptual design for the toroid shield. In the old design, shown in the left part of the figure above, the moderator part of the shielding (JTV) was situated both in the warm and cold areas of the forward toroid. It consisted both of rings of polyethylene and hundreds of blocks of polyethylene (or an epoxy resin) inside the toroid vacuum vessel. In the new design, shown to the right in the figure above, only the rings remain inside the toroid. To compensate for the loss of moderator in the toroid, the copper plug (JTT) has been reduced in radius so that a layer of borated polyethylene can be placed around it (see figure below). The new design gives significant cost-savings and is easier to produce in the tight time schedule of the forward toroid. Since the amount of copper is reduced the weight that has to be carried by the toroid is also reduced. Outgassing into the toroid vacuum was a potential problem in the old design and this is now avoided. The main ...

  9. Drip Shield Emplacement Gantry Concept

    Energy Technology Data Exchange (ETDEWEB)

    Silva, R.A.; Cron, J.

    2000-03-29

    This design analysis has shown that, on a conceptual level, the emplacement of drip shields is feasible with current technology and equipment. A plan for drip shield emplacement was presented using a Drip Shield Transporter, a Drip Shield Emplacement Gantry, a locomotive, and a Drip Shield Gantry Carrier. The use of a Drip Shield Emplacement Gantry as an emplacement concept results in a system that is simple, reliable, and interfaces with the numerous other exising repository systems. Using the Waste Emplacement/Retrieval System design as a basis for the drip shield emplacement concept proved to simplify the system by using existing equipment, such as the gantry carrier, locomotive, Electrical and Control systems, and many other systems, structures, and components. Restricted working envelopes for the Drip Shield Emplacement System require further consideration and must be addressed to show that the emplacement operations can be performed as the repository design evolves. Section 6.1 describes how the Drip Shield Emplacement System may use existing equipment. Depending on the length of time between the conclusion of waste emplacement and the commencement of drip shield emplacement, this equipment could include the locomotives, the gantry carrier, and the electrical, control, and rail systems. If the exisiting equipment is selected for use in the Drip Shield Emplacement System, then the length of time after the final stages of waste emplacement and start of drip shield emplacement may pose a concern for the life cycle of the system (e.g., reliability, maintainability, availability, etc.). Further investigation should be performed to consider the use of existing equipment for drip shield emplacement operations. Further investigation will also be needed regarding the interfaces and heat transfer and thermal effects aspects. The conceptual design also requires further design development. Although the findings of this analysis are accurate for the assumptions made

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

    Directory of Open Access Journals (Sweden)

    Han Bing

    2010-04-01

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

  11. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

  12. Direct Lit Electrolysis In A Metallic Lithium Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-13

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

  13. Reflective Shields for Artificial Satellites

    Science.gov (United States)

    Bouquet, F. L.

    1986-01-01

    Report proposes reflective shield that protects spacecraft from radiant energy. Also gives some protection against particle beams and cosmic rays. Conceptual shield essentially advanced version of decorative multifaceted mirror balls often hung over dance floors. Mirror facets disperse radiant energy in many directions.

  14. The shield effect

    DEFF Research Database (Denmark)

    Toft, Søren; Albo, Maria J

    2016-01-01

    Several not mutually exclusive functions have been ascribed to nuptial gifts across different taxa. Although the idea that a nuptial prey gift may protect the male from pre-copulatory sexual cannibalism is attractive, it has previously been considered of no importance based on indirect evidence and...... rejected by experimental tests. We reinvestigated whether nuptial gifts may function as a shield against female attacks during mating encounters in the spider Pisaura mirabilis and whether female hunger influences the likelihood of cannibalistic attacks. The results showed that pre-copulatory sexual...... cannibalism was enhanced when males courted without a gift and this was independent of female hunger. We propose that the nuptial gift trait has evolved partly as a counteradaptation to female aggression in this spider species....

  15. Sulphate resistant shielding material

    International Nuclear Information System (INIS)

    The shielding material of the present invention is provided with sulfuric acid resistance and contains bentonite put to ion exchange treatment with barium ions as an effective ingredient. When mortars and concretes are exposed to the circumstance of sulfate, the effective ingredient functions to take place reaction between intruding sulfate and the barium ions to form insoluble barium sulfate thereby reducing chemical corrosion of mortars and concretes caused by sulfate. Cement materials, water and aggregates can optionally be contained in addition to bentonite and bentonite put to ion exchange treatment. Chemical corrosion of concretes and mortars due to intrusion of the sulfate can be prevented, and it is useful as an artificial barrier, for example, in radioactive active waste processing facilities. (T.M.)

  16. Welding shield for coupling heaters

    Science.gov (United States)

    Menotti, James Louis

    2010-03-09

    Systems for coupling end portions of two elongated heater portions and methods of using such systems to treat a subsurface formation are described herein. A system may include a holding system configured to hold end portions of the two elongated heater portions so that the end portions are abutted together or located near each other; a shield for enclosing the end portions, and one or more inert gas inlets configured to provide at least one inert gas to flush the system with inert gas during welding of the end portions. The shield may be configured to inhibit oxidation during welding that joins the end portions together. The shield may include a hinged door that, when closed, is configured to at least partially isolate the interior of the shield from the atmosphere. The hinged door, when open, is configured to allow access to the interior of the shield.

  17. First wall and blanket concepts for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

  19. Tritium and heat management in ITER Test Blanket Systems port cell for maintenance operations

    International Nuclear Information System (INIS)

    Highlights: •The ITER TBM Program is one of the ITER missions. •We model a TBM port cell with CFD to optimize the design choices. •The heat and tritium releases management in TBM port cells has been optimized. •It is possible to reduce the T-concentration below one DAC in TBM port cells. •The TBM port cells can have human access within 12 h after shutdown. -- Abstract: Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown

  20. A ceramic breeder in a poloidal tube blanket for a tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Amici, A.; Anzidei, L.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.; Zampaglione, V.; Petrizzi, L. (Associazione Euratom-CNEN sulla Fusione, Centro di Frascati (Italy))

    1989-04-01

    A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods ({gamma}LiAlO/sub 2/) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to {gamma}LiAlO/sub 2/ volume ratio is 4/1. The He inlet and outlet branches are cooling Be and {gamma}LiAlO/sub 2/, respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load=2 MW/m/sup 2/), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature=570/sup 0/C; inlet He temperature=250/sup 0/; total extracted power=2700 MW; He pumping power percentage=2%; minimum/maximum {gamma}LiAlO/sub 2/ temperature=400/900/sup 0/C; maximum structural temperature=475/sup 0/C; and maximum Be temperature=525/sup 0/C. (orig.).

  1. 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 ...

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

    OpenAIRE

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

    2010-01-01

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

  3. 75 FR 60095 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-09-29

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

  4. Radiation shields for a shelter

    International Nuclear Information System (INIS)

    A simple and cheap closure and radiation shield arrangement is described for the entrance of an underground shelter. The shelter can serve as a blast-proof, biological or nuclear shelter. The radiation shield is positioned above the habitable space of the shelter and below a blast-proof, dust-proof outer cover. The shield consists of a box containing a filling, e.g. coke with a concrete screed, is closed by bolted panels and is horizontally moveable by sliding on castors. (author)

  5. New Materials for EMI Shielding

    Science.gov (United States)

    Gaier, James R.

    1999-01-01

    Graphite fibers intercalated with bromine or similar mixed halogen compounds have substantially lower resistivity than their pristine counterparts, and thus should exhibit higher shielding effectiveness against electromagnetic interference. The mechanical and thermal properties are nearly unaffected, and the shielding of high energy x-rays and gamma rays is substantially increased. Characterization of the resistivity of the composite materials is subtle, but it is clear that the composite resistivity is substantially lowered. Shielding effectiveness calculations utilizing a simple rule of mixtures model yields results that are consistent with available data on these materials.

  6. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

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

  7. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m2. Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

  8. 75 FR 11557 - Woven Electric Blankets From China

    Science.gov (United States)

    2010-03-11

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

  9. 18 CFR 33.1 - Applicability, definitions, and blanket authorizations.

    Science.gov (United States)

    2010-04-01

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

  10. Shielding synchrotron light sources: Advantages of circular shield walls tunnels

    Science.gov (United States)

    Kramer, S. L.; Ghosh, V. J.; Breitfeller, M.

    2016-08-01

    Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produce significantly higher neutron component dose to the experimental floor than lower energy injection and ramped operations. High energy neutrons produced in the forward direction from thin target beam losses are a major component of the dose rate outside the shield walls of the tunnel. The convention has been to provide thicker 90° ratchet walls to reduce this dose to the beam line users. We present an alternate circular shield wall design, which naturally and cost effectively increases the path length for this forward radiation in the shield wall and thereby substantially decreasing the dose rate for these beam losses. This shield wall design will greatly reduce the dose rate to the users working near the front end optical components but will challenge the beam line designers to effectively utilize the longer length of beam line penetration in the shield wall. Additional advantages of the circular shield wall tunnel are that it's simpler to construct, allows greater access to the insertion devices and the upstream in tunnel beam line components, as well as reducing the volume of concrete and therefore the cost of the shield wall.

  11. Radiation shielding for neutron guides

    Energy Technology Data Exchange (ETDEWEB)

    Ersez, T. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia)]. E-mail: tez@ansto.gov.au; Braoudakis, G. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia); Osborn, J.C. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia)

    2006-11-15

    Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions.

  12. Performance test on shielding concrete

    International Nuclear Information System (INIS)

    The cylinder of the shielding concrete is made from common Portland cement and home-made coarse or fine aggregates. Orthogonal design experiment and regression analysis are adopted to study the effects of the water content, sand percentage and water-cement ratio on the property of shielding concrete and the difference between them. The test shows that the tensile strength is in inverse proportion with water-cement ratio, and the influence is quite significant. Another factor is the type of aggregates. The effect of the age on its density is not obvious. Similarly, the concrete shielding γ rays shares the same influencing factors with that shielding neutron rays on density, slump and tensile strength. And both have the same change rules regarding to mechanical property. (authors)

  13. SNF shipping cask shielding analysis

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Pace, J.V. III

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan.

  14. Shielding of moving line charges

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Youmei; He, Bingyu [Department of Physics, School of Science, Hangzhou Dianzi University, Hangzhou 310018 (China); Yu, Wei [Shanghai Institute of Optics and Fine Mechanics, Chinese Academy of Sciences, Shanghai 201800 (China); Yu, M.Y., E-mail: myyu@zju.edu.cn [Institute for Fusion Theory and Simulation and Department of Physics, Zhejiang University, Hangzhou 310027 (China); Institute for Theoretical Physics I, Ruhr University, D-44780 Bochum (Germany)

    2015-07-03

    A charged object moving in plasma can excite plasma waves that inevitably modify its Debye shielding characteristics. When the excited waves propagate sufficiently fast, the shielding can even break down. Here the properties of finite amplitude plasma waves excited by a moving line charge are investigated. It is found that when the speed of the latter is close to but less than the thermal speed of the background plasma electrons, only a localized disturbance in the form of a soliton that moves together with the line charge is excited. That is, the line charge is well shielded even though it is moving at a high speed and has generated a large local electrostatic field. However, for a pair of line charges moving together, such complete shielding behavior could not be found.

  15. A fail-safe and cost effective fabrication route for blanket First Walls

    Science.gov (United States)

    Commin, L.; Rieth, M.; Dafferner, B.; Zimmermann, H.; Bolich, D.; Baumgärtner, S.; Ziegler, R.; Dichiser, S.; Fabry, T.; Fischer, S.; Hildebrand, W.; Palussek, O.; Ritz, H.; Sponda, A.

    2013-11-01

    Helium Cooled Lithium Lead and Helium Cooled Pebble Bed concepts have been selected as European Test Blanket Modules (TBM) for ITER. The TBM fabrication will need the assembly of six Reduced Activation Ferritic Martensitic steel sub-components, namely First Wall, Caps, Stiffening Grid, Breeding Units, Back Plates/Manifolds, and Attachment system. The fabrication of the First Wall requires the production of cooling channels inside 30 mm thick bended plates. For this specific component, the main issues consist of the lack of accessibility of some areas to join, the process tolerances, the dimensional stability and the resulting assembly mechanical properties. Several fabrication routes have been already investigated, which involve diffusion welding and fusion welding (electron beam, laser beam, hybrid MIG/laser).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Shielding vacuum fluctuations with graphene

    OpenAIRE

    Ribeiro, Sofia; Scheel, Stefan

    2013-01-01

    The Casimir-Polder interaction of ground-state and excited atoms with graphene is investigated with the aim to establish whether graphene systems can be used as a shield for vacuum fluctuations of an underlying substrate. We calculate the zero-temperature Casimir-Polder potential from the reflection coefficients of graphene within the framework of the Dirac model. For both doped and undoped graphene we show limits at which graphene could be used effectively as a shield. Additional results are...

  20. Composite Aerogel Multifoil Protective Shielding

    Science.gov (United States)

    Jones, Steven M.

    2013-01-01

    New technologies are needed to survive the temperatures, radiation, and hypervelocity particles that exploration spacecraft encounter. Multilayer insulations (MLIs) have been used on many spacecraft as thermal insulation. Other materials and composites have been used as micrometeorite shielding or radiation shielding. However, no material composite has been developed and employed as a combined thermal insulation, micrometeorite, and radiation shielding. By replacing the scrims that have been used to separate the foil layers in MLIs with various aerogels, and by using a variety of different metal foils, the overall protective performance of MLIs can be greatly expanded to act as thermal insulation, radiation shielding, and hypervelocity particle shielding. Aerogels are highly porous, low-density solids that are produced by the gelation of metal alkoxides and supercritical drying. Aerogels have been flown in NASA missions as a hypervelocity particle capture medium (Stardust) and as thermal insulation (2003 MER). Composite aerogel multifoil protective shielding would be used to provide thermal insulation, while also shielding spacecraft or components from radiation and hypervelocity particle impacts. Multiple layers of foil separated by aerogel would act as a thermal barrier by preventing the transport of heat energy through the composite. The silica aerogel would act as a convective and conductive thermal barrier, while the titania powder and metal foils would absorb and reflect the radiative heat. It would also capture small hypervelocity particles, such as micrometeorites, since it would be a stuffed, multi-shock Whipple shield. The metal foil layers would slow and break up the impacting particles, while the aerogel layers would convert the kinetic energy of the particles to thermal and mechanical energy and stop the particles.

  1. Effect of heat shielding on convective and evaporative heat losses and on radiant heat transfer in the premature infant

    Energy Technology Data Exchange (ETDEWEB)

    Baumgart, S.; Engle, W.D.; Fox, W.W.; Polin, R.A.

    1981-12-01

    Ten premature infants nursed on servocontrolled radiant warmer beds were studied in three environments designed to alter one or more factors affecting heat transfer (convection, evaporation, and radiation). In the control environment, infants were nursed supine on an open warmer bed. The second environment (walled chamber) was designed to reduce convection and evaporation by placing plastic walls circumferentially around the bed. In the third environment convection and evaporation were minimized by covering infants with a plastic blanket. Air turbulence, insensible water loss, and radiant warmer power were measured in each environment. There was a significant reduction in mean air velocity in the walled chamber and under the plastic blanket when compared to the control environment. A parallel decrease in insensible water loss occurred. In contrast, radiant power demand was the same for control and walled environments, but decreased significantly when infants were covered by the plastic blanket. This study suggests that convection is an important factor influencing evaporation in neonates nursed under radiant warmers. The thin plastic blanket was the most effective shield, significantly reducing radiant power demand.

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

    International Nuclear Information System (INIS)

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m2 and a surface heat flux of 1 MW/m2. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO2 rods. The helium coolant pressure is 5 MPa, entering the module at 2970C and exiting at 5500C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter

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

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A. (eds.)

    1986-09-01

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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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.

  9. MFTF-B Upgrade for blanket-technology testing

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    1988-01-01

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

  11. 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.

  12. 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

  13. Model problem of MHD flow in a lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Cherepanov, V.Y.

    1978-01-01

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

  14. 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.

  15. Radiation Shielding Materials and Containers Incorporating Same

    Energy Technology Data Exchange (ETDEWEB)

    Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

    2005-11-01

    An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

  16. Modular Shielded Solar Panels Made using Low Cost Lamination Process

    OpenAIRE

    Stern, Theodore

    2007-01-01

    A modular, laminated solar panel approach has been demonstrated which uses high efficiency cells and allows application to a variety of smallsat footprints without customizing each application. The module design includes a contiguous integral shield which protects the solar cells more effectively from a variety of space environments, and also provides electromagnetic cleanliness to protect sensitive instruments from solar array generated EMI. The laminated approach also provides significant c...

  17. 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.)

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

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 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.

  1. A Novel Radiation Shielding Material Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Radiation shielding simulations showed that epoxy loaded with 10-70% polyethylene would be an excellent shielding material against GCRs and SEPs. Milling produced...

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    2009-07-01

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

  4. Progress of R and D and design of blanket remote handling equipment for ITER

    International Nuclear Information System (INIS)

    The design of in-vessel transporter (IVT) including vehicle manipulator has been updated according to the design changes such as blanket segmentation and structure, taking account of the interface between modules and vehicle manipulator. In particular, the updated design of the vehicle manipulator and rail has been carried out because of collision avoidance between modules and vehicle manipulator. According to the updated design, the vehicle manipulator has been reduced by about 30% in weight, compared with the reference design. In parallel with design activities, the R and D to clarify the specifications of the IVT design in detail is also performed, i.e., simulation system to provide the visual information during maintenance, dry lubricant to prevent the lubricant oil from spreading in the vacuum vessel (VV). The rail connection and cable handling in the transfer cask, which are critical issues for IVT system, are under preparation of the demonstration tests to finalize the design of the IVT system. Connection of the rail joint and cable handling test facilities are planned and under fabrication now. These test facility will be installed by the end of March 2008, and the performance tests will be carried out from April 2008

  5. Paddle-based rotating-shield brachytherapy

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Yunlong; Xu, Weiyu [Department of Electrical and Computer Engineering, University of Iowa, 4016 Seamans Center, Iowa City, Iowa 52242 (United States); Flynn, Ryan T.; Kim, Yusung; Bhatia, Sudershan K.; Buatti, John M. [Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States); Dadkhah, Hossein [Department of Biomedical Engineering, University of Iowa, 1402 Seamans Center, Iowa City, Iowa 52242 (United States); Wu, Xiaodong, E-mail: xiaodong-wu@uiowa.edu [Department of Electrical and Computer Engineering, University of Iowa, 4016 Seamans Center, Iowa City, Iowa 52242 and Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States)

    2015-10-15

    Purpose: The authors present a novel paddle-based rotating-shield brachytherapy (P-RSBT) method, whose radiation-attenuating shields are formed with a multileaf collimator (MLC), consisting of retractable paddles, to achieve intensity modulation in high-dose-rate brachytherapy. Methods: Five cervical cancer patients using an intrauterine tandem applicator were considered to assess the potential benefit of the P-RSBT method. The P-RSBT source used was a 50 kV electronic brachytherapy source (Xoft Axxent™). The paddles can be retracted independently to form multiple emission windows around the source for radiation delivery. The MLC was assumed to be rotatable. P-RSBT treatment plans were generated using the asymmetric dose–volume optimization with smoothness control method [Liu et al., Med. Phys. 41(11), 111709 (11pp.) (2014)] with a delivery time constraint, different paddle sizes, and different rotation strides. The number of treatment fractions (fx) was assumed to be five. As brachytherapy is delivered as a boost for cervical cancer, the dose distribution for each case includes the dose from external beam radiotherapy as well, which is 45 Gy in 25 fx. The high-risk clinical target volume (HR-CTV) doses were escalated until the minimum dose to the hottest 2 cm{sup 3} (D{sub 2cm{sup 3}}) of either the rectum, sigmoid colon, or bladder reached their tolerance doses of 75, 75, and 90 Gy{sub 3}, respectively, expressed as equivalent doses in 2 Gy fractions (EQD2 with α/β = 3 Gy). Results: P-RSBT outperformed the two other RSBT delivery techniques, single-shield RSBT (S-RSBT) and dynamic-shield RSBT (D-RSBT), with a properly selected paddle size. If the paddle size was angled at 60°, the average D{sub 90} increases for the delivery plans by P-RSBT on the five cases, compared to S-RSBT, were 2.2, 8.3, 12.6, 11.9, and 9.1 Gy{sub 10}, respectively, with delivery times of 10, 15, 20, 25, and 30 min/fx. The increases in HR-CTV D{sub 90}, compared to D-RSBT, were 16

  6. 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)

  7. Impact of prescribed burning on blanket peat hydrology

    Science.gov (United States)

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

    2015-08-01

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

  8. One Year assessment of shielding for a multi-energy linear accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Gi; Carlson, Joel; Lee, Hyun Seok; Ye, Sung Joon [Seoul National University Graduate School of Convergence Science and Technology, Seoul (Korea, Republic of); Chung, Jin Beom; Kim, Jae Sung [Dept. of Radiation Oncology, Seoul National University Bundang Hospital, Seoul (Korea, Republic of); Kim, Jung In [Dept. of of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of)

    2014-11-15

    In 2005, the publication of Report No. 151 of the National Council on Radiation Protection and Measurements (NCRP) suggested shielding methodologies along with shielding data. Recently, intensity modulated radiation therapy (IMRT) and volumetric modulated arc therapy (VMAT) have become more widely used for cancer treatment. Thus, we analyzed shielding parameters for a multi-energy medical linear accelerator using the VMAT technique. Calculated total workload was similar to the recommendation of NCRP Report No. 49 and No. 51. However, these results were higher than the previous results in the NCRP Report No. 151. Also, the VMAT technique uses an intensity modulated beams with various gantry angles so that scattered and leakage doses should be carefully considered by retrospective analysis using the treatment data from each facility.

  9. Recovery of tritium from a liquid lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Talbot, J.B.

    1981-01-01

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

  10. 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

  11. A new radiation shielding material: Amethyst ore

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay, E-mail: turgaykorkut@hotmail.co [Faculty of Science and Art, Department of Physics, Ibrahim Cecen University, Agri (Turkey); Korkut, Hatun [Faculty of Science and Art, Department of Physics, Ibrahim Cecen University, Agri (Turkey); Karabulut, Abdulhalik; Budak, Goekhan [Faculty of Science, Department of Physics, Atatuerk University, Erzurum (Turkey)

    2011-01-15

    This paper describes a new radiation shielding material, amethyst ore. We have determined the elemental composition of amethyst using WDXRF spectroscopy technique. To see the shielding capability of amethyst for several photon energies, these results have been used in simulation process by FLUKA Monte Carlo radiation transport code. Linear attenuation coefficients have been calculated according to the simulation results. Then, these values have been compared to a fine shielding concrete material. The results show that amethyst shields more gamma beams than concrete. This investigation is the first study about the radiation shielding properties of amethyst ore.

  12. Facility target insert shielding assessment

    Energy Technology Data Exchange (ETDEWEB)

    Mocko, Michal [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-06

    Main objective of this report is to assess the basic shielding requirements for the vertical target insert and retrieval port. We used the baseline design for the vertical target insert in our calculations. The insert sits in the 12”-diameter cylindrical shaft extending from the service alley in the top floor of the facility all the way down to the target location. The target retrieval mechanism is a long rod with the target assembly attached and running the entire length of the vertical shaft. The insert also houses the helium cooling supply and return lines each with 2” diameter. In the present study we focused on calculating the neutron and photon dose rate fields on top of the target insert/retrieval mechanism in the service alley. Additionally, we studied a few prototypical configurations of the shielding layers in the vertical insert as well as on the top.

  13. New facility shield design criteria

    International Nuclear Information System (INIS)

    The purpose of the criteria presented here is to provide standard guidance for the design of nuclear radiation shields thoughout new facilities. These criteria are required to assure a consistent and integrated design that can be operated safely and economically within the DOE standards. The scope of this report is confined to the consideration of radiation shielding for contained sources. The whole body dose limit established by the DOE applies to all doses which are generally distributed throughout the trunk of the body. Therefore, where the whole body is the critical organ for an internally deposited radionuclide, the whole body dose limit applies to the sum of doses received must assure control of the concentration of radionuclides in the building atmosphere and thereby limit the dose from internal sources

  14. Shielding calculational system for plutonium

    International Nuclear Information System (INIS)

    A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

  15. Stellar activity and magnetic shielding

    CERN Document Server

    Grießmeier, J -M; Lammer, H; Grenfell, J L; Stadelmann, A; Motschmann, U; 10.1017/S1743921309992961

    2010-01-01

    Stellar activity has a particularly strong influence on planets at small orbital distances, such as close-in exoplanets. For such planets, we present two extreme cases of stellar variability, namely stellar coronal mass ejections and stellar wind, which both result in the planetary environment being variable on a timescale of billions of years. For both cases, direct interaction of the streaming plasma with the planetary atmosphere would entail servere consequences. In certain cases, however, the planetary atmosphere can be effectively shielded by a strong planetary magnetic field. The efficiency of this shielding is determined by the planetary magnetic dipole moment, which is difficult to constrain by either models or observations. We present different factors which influence the strength of the planetary magnetic dipole moment. Implications are discussed, including nonthermal atmospheric loss, atmospheric biomarkers, and planetary habitability.

  16. Paramagnetism shielding in drilling fluid

    OpenAIRE

    Li, Zhuo

    2013-01-01

    In drilling operations, drilling fluid containing magnetic materials is used when drilling a well. The materials can significantly shield the Earth’s magnetic field as measured by magnetic sensors inside the drilling strings. The magnetic property of the drilling fluid is one of the substantial error sources for the determination of magnetic azimuth for wellbores. Both the weight material, cuttings, clay and other formation material plus metal filings from the tubular wear m...

  17. Photonic Bandgap (PBG) Shielding Technology

    Science.gov (United States)

    Bastin, Gary L.

    2007-01-01

    Photonic Bandgap (PBG) shielding technology is a new approach to designing electromagnetic shielding materials for mitigating Electromagnetic Interference (EM!) with small, light-weight shielding materials. It focuses on ground planes of printed wiring boards (PWBs), rather than on components. Modem PSG materials also are emerging based on planar materials, in place of earlier, bulkier, 3-dimensional PBG structures. Planar PBG designs especially show great promise in mitigating and suppressing EMI and crosstalk for aerospace designs, such as needed for NASA's Constellation Program, for returning humans to the moon and for use by our first human visitors traveling to and from Mars. Photonic Bandgap (PBG) materials are also known as artificial dielectrics, meta-materials, and photonic crystals. General PBG materials are fundamentally periodic slow-wave structures in I, 2, or 3 dimensions. By adjusting the choice of structure periodicities in terms of size and recurring structure spacings, multiple scatterings of surface waves can be created that act as a forbidden energy gap (i.e., a range of frequencies) over which nominally-conductive metallic conductors cease to be a conductor and become dielectrics. Equivalently, PBG materials can be regarded as giving rise to forbidden energy gaps in metals without chemical doping, analogous to electron bandgap properties that previously gave rise to the modem semiconductor industry 60 years ago. Electromagnetic waves cannot propagate over bandgap regions that are created with PBG materials, that is, over frequencies for which a bandgap is artificially created through introducing periodic defects

  18. Steam generator hand hole shielding.

    Science.gov (United States)

    Cox, W E

    2000-05-01

    Seabrook Station is an 1198 MWE Pressurized Water Reactor (PWR) that began commercial operation in 1990. Expensive and dose intensive Steam Generator Replacement Projects among PWR operators have led to an increase in steam generator preventative maintenance. Most of this preventative maintenance is performed through access ports in the shell of the steam generator just above the tube sheet known as secondary side hand holes. Secondary side work activities performed through the hand holes are typically performed without the shielding benefit of water in the secondary side of the steam generator. An increase in cleaning and inspection work scope has led to an increase in dose attributed to steam generator secondary side maintenance. This increased work scope and the station goal of maintaining personnel radiation dose ALARA led to the development of the shielding concept described in this article. This shield design saved an estimated 2.5 person-rem (25 person-Smv) the first time it was deployed and is expected to save an additional 50 person-rem (500 person-mSv) over the remaining life of the plant. PMID:10770158

  19. ATLAS Award for Shield Supplier

    CERN Multimedia

    2004-01-01

    ATLAS technical coordinator Dr. Marzio Nessi presents the ATLAS supplier award to Vojtech Novotny, Director General of Skoda Hute.On 3 November, the ATLAS experiment honoured one of its suppliers, Skoda Hute s.r.o., of Plzen, Czech Republic, for their work on the detector's forward shielding elements. These huge and very massive cylinders surround the beampipe at either end of the detector to block stray particles from interfering with the ATLAS's muon chambers. For the shields, Skoda Hute produced 10 cast iron pieces with a total weight of 780 tonnes at a cost of 1.4 million CHF. Although there are many iron foundries in the CERN member states, there are only a limited number that can produce castings of the necessary size: the large pieces range in weight from 59 to 89 tonnes and are up to 1.5 metres thick.The forward shielding was designed by the ATLAS Technical Coordination in close collaboration with the ATLAS groups from the Czech Technical University and Charles University in Prague. The Czech groups a...

  20. Water shielding nuclear reactor container

    International Nuclear Information System (INIS)

    The reactor container of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevated inner pressure and keeping airtightness, and shielding water is filled inside from a water injection port. It is endurable to a great inner pressure satisfactorily and keep airtightness by the two spaced relatively thin steel plates. It exhibits radiation shielding effect by filling water substantially the same as that of a conventional reactor container made of iron reinforced concretes. Then, it is no more necessary to use concretes for the construction of the reactor container, which shortens the term of the construction, and saves the construction cost. In addition, a cooling effect for the reactor container is provided. Syphons are disposed contiguously to a water injection port and the top end of the syphon is immersed in an equipment temporarily storage pool, and further, pipelines are connected to the double steel plate walls or the syphons for supplying shielding water to enhance the cooling effect. (N.H.)

  1. 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

  2. Fission power flattening in hybrid blankets using mixed fuel

    International Nuclear Information System (INIS)

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

  3. 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

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

    Science.gov (United States)

    Zikanov, Oleg; Liu, Li

    2015-11-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. 77 FR 76013 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-12-26

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

  8. 75 FR 19954 - Cheniere Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-04-16

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

  9. 78 FR 4400 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2013-01-22

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

  10. Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries

    International Nuclear Information System (INIS)

    We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1

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

    Directory of Open Access Journals (Sweden)

    N. Lukwa

    2013-04-01

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

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

    International Nuclear Information System (INIS)

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

  13. Thermal and structural design issues of breeding blankets for testing in the Next European Torus (NET)

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.

    1988-05-01

    A review of the breeding blankets under study in Europe for testing in the Next European Torus is presented. In many concepts, the breeder modules are enclosed in boxes whose side walls in front of the plasma act as the first wall of the machine. Various types of breeder modules are investigated, involving both liquid and solid breeders, namely: - Pb-17Li liquid breeder concepts, the coolant being either water or Pb-17Li itself; - solid (ceramic) breeder concepts, the coolant being in all cases helium. The various ceramic concepts differ in the breeder/coolant arrangement (breeder-out-of-tube and breeder-in-tube), the orientation of the coolant tubes (poloidal or toroidal) and the breeder geometry (rods, plates or pebble bed). For each of these concepts the main design features are shown and the thermomechanical problems are discussed. The problems related to a coolant tube rupture are in many cases the most severe from the structural design point of view. The first wall box enclosing the breeder modules appears to be a weak secondary containment barrier. The liquid breeder-water cooled concept looks manageable from the thermal and structural design of point view. In the case of the self-cooled liquid breeder concept, the main problems are related to the magnetohydrodynamic effects. Solutions are envisaged to overcome these difficulties. In the case of ceramic breeders, the use of plates implies small dimensions in order to limit the thermal stresses and a poor exploitation of the permitted temperature operation window. Solutions involving rods associated with a multipass cooling scheme or pebble bed enable achievement of better thermomechanical conditions and, therefore, are preferred in the current investigations. However, they lead to design complications and require experimental verification which is in progress at the European laboratories.

  14. EMI Shields made from intercalated graphite composites

    Science.gov (United States)

    Gaier, James R.; Terry, Jennifer

    1995-01-01

    Electromagnetic interference (EMI) shielding typically makes up about twenty percent of the mass of a spacecraft power system. Graphite fiber/polymer composites have significantly lower densities and higher strengths than aluminum, the present material of choice for EMI shields, but they lack the electrical conductivity that enables acceptable shielding effectiveness. Bromine intercalated pitch-based graphite/epoxy composites have conductivities fifty times higher than conventional structural graphite fibers. Calculations are presented which indicate that EMI shields made from such composites can have sufficient shielding at less than 20% of the mass of conventional aluminum shields. EMI shields provide many functions other than EMI shielding including physical protection, thermal management, and shielding from ionizing radiation. Intercalated graphite composites perform well in these areas also. Mechanically, they have much higher specific strength and modulus than aluminum. They also have shorter half thicknesses for x-rays and gamma radiation than aluminum. Thermally, they distribute infra-red radiation by absorbing and re-radiating it rather than concentrating it by reflection as aluminum does. The prospects for intercalated graphite fiber/polymer composites for EMI shielding are encouraging.

  15. Verification of Monte Carlo transport codes FLUKA, Mars and Shield

    International Nuclear Information System (INIS)

    The present study is a continuation of the project 'Verification of Monte Carlo Transport Codes' which is running at GSI as a part of activation studies of FAIR relevant materials. It includes two parts: verification of stopping modules of FLUKA, MARS and SHIELD-A (with ATIMA stopping module) and verification of their isotope production modules. The first part is based on the measurements of energy deposition function of uranium ions in copper and stainless steel. The irradiation was done at 500 MeV/u and 950 MeV/u, the experiment was held at GSI from September 2004 until May 2005. The second part is based on gamma-activation studies of an aluminium target irradiated with an argon beam of 500 MeV/u in August 2009. Experimental depth profiling of the residual activity of the target is compared with the simulations. (authors)

  16. Highly heat-removing radiation shielding material

    International Nuclear Information System (INIS)

    Highly heat-removing radiation shielding material is constituted with fine particles prepared by coating metals of high heat conductivity to fine particles comprising materials having excellent radiation shielding performance. Then, the fine particles applied with the coating are mixed and filled in a shielding container or applied with hot press into a layerous form and used as a shielding member. In view of the above, since the coated fine particles provide the shielding performance against radiation such as neutrons and gamma rays, and the coating metals provide the heat removing performance, they act as a shielding material having heat removing performance as a whole. The combination of the coated fine particles and the coating metals are selected depending on the respective conditions for use. With such a constitution, radioactive wastes involving a problem of heat generation can be transported or stored safely. (T.M.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-06-05

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

  19. VAPOR SHIELD FOR INDUCTION FURNACE

    Science.gov (United States)

    Reese, S.L.; Samoriga, S.A.

    1958-03-11

    This patent relates to a water-cooled vapor shield for an inductlon furnace that will condense metallic vapors arising from the crucible and thus prevent their condensation on or near the induction coils, thereby eliminating possible corrosion or shorting out of the coils. This is accomplished by placing, about the top, of the crucible a disk, apron, and cooling jacket that separates the area of the coils from the interior of the cruclbIe and provides a cooled surface upon whlch the vapors may condense.

  20. 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.)

  1. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-26

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-02

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

  3. 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.

  4. 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.

  5. Cosmetic wastewater treatment by upflow anaerobic sludge blanket reactor

    International Nuclear Information System (INIS)

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

  6. Cosmetic wastewater treatment by upflow anaerobic sludge blanket reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-30

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

  7. Radiation shielding concrete made of Basalt aggregates.

    Science.gov (United States)

    Alhajali, S; Yousef, S; Kanbour, M; Naoum, B

    2013-04-01

    In spite of the fact that Basalt is a widespread type of rock, there is very little available information on using it as aggregates for concrete radiation shielding. This paper investigates the possibility of using Basalt for the aforementioned purpose. The results have shown that Basalt could be used successfully for preparing radiation shielding concrete, but some attention should be paid to the choice of the suitable types of Basalt and for the neutron activation problem that could arise in the concrete shield.

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

    International Nuclear Information System (INIS)

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

  9. 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)

  10. Shielding for thermoacoustic tomography with RF excitation

    Science.gov (United States)

    Mitchell, M.; Becker, G.; Dey, P.; Generotzky, J.; Patch, S. K.

    2008-02-01

    Radiofrequency (RF) pulses used to generate thermoacoustic computerized tomography (TCT) signal couple directly into the pulser-receiver and oscilloscope, swamping true TCT signal. We use a standard RF enclosure housing both RF amplifier and object being imaged. This is similar to RF shielding of magnetic resonance imaging (MRI) suites and protects electronics outside from stray RF. Unlike MRI, TCT receivers are ultrasound transducers, which must also be shielded from RF. A transducer housing that simultaneously shields RF and permits acoustic transmission was developed specifically for TCT. We compare TCT signals measured with and without RF shielding.

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

    Directory of Open Access Journals (Sweden)

    S. Chakraborty

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

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

  14. TPX remote maintenance and shielding

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment machine design incorporates comprehensive planning for efficient and safe component maintenance. Three programmatic decisions have been made to insure the successful implementation of this objective. First, the tokamak incorporates radiation shielding to reduce activation of components and limit the dose rate to personnel working on the outside of the machine. This allows most of the ex-vessel equipment to be maintained through conventional ''hands-on'' procedures. Second, to the maximum extent possible, low activation materials will be used inside the shielding volume. This resulted in the selection of Titanium (Ti-6Al-4V) for the vacuum vessel and PFC structures. The third decision stipulated that the primary in-vessel components will be replaced or repaired via remote maintenance tools specifically provided for the task. The component designers have been given the responsibility of incorporating maintenance design and for proving the maintainability of the design concepts in full-scale mockup tests prior to the initiation of final fabrication. Remote maintenance of the TPX machine is facilitated by general purpose tools provided by a special purpose design team. Major tools will include an in-vessel transporter, a vessel transfer system and a large component transfer container. In addition, tools such as manipulators and remotely operable impact wrenches will be made available to the component designers by this group. Maintenance systems will also provide the necessary controls for this equipment

  15. TPX remote maintenance and shielding

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) machine design incorporates comprehensive planning for efficient and safe component maintenance. Three programmatic decisions have been made to insure the successful implementation of this objective. First, the tokamak incorporates radiation shielding to reduce activation of components and limit the dose rate to personnel working on the outside of the machine. This allows most of the ex-vessel equipment to be maintained through conventional open-quotes hands-onclose quotes procedures. Second, to the maximum extent possible, low activation materials will be used inside the shielding volume. This resulted in the selection of Titanium (Ti-6Al-4V) for the vacuum vessel and Plasma Facing Components (PFC) structures. The third decision stipulated that the primary in-vessel components will be replaced or repaired via remote maintenance tools specifically provided for the task. The component designers have been given the responsibility of incorporating maintenance design and for proving the maintainability of the design concepts in full-scale mockup tests prior to the initiation of final fabrication. Remote maintenance of the TPX machine is facilitated by general purpose tools provided by a special purpose design team. Major tools will include an in-vessel transporter, a vessel transfer system and a large component transfer container. In addition, tools such as manipulators and remotely operable impact wrenches will be made available to the component designers by this group. Maintenance systems will also provide the necessary controls for this equipment

  16. Safety analysis and lay-out aspects of shieldings against particle radiation at the example of spallation facilities in the megawatt range; Sicherheitstechnische Analyse und Auslegungsaspekte von Abschirmungen gegen Teilchenstrahlung am Beispiel von Spallationsanlagen im Megawatt Bereich

    Energy Technology Data Exchange (ETDEWEB)

    Hanslik, R.

    2006-08-15

    This paper discusses the shielding of particle radiation from high current accelerators, spallation neutron sources and so called ADS-facilities (Accelerator Driven Systems). ADS-facilities are expected to gain importance in the future for transmutation of long-lived isotopes from fission reactors as well as for energy production. In this paper physical properties of the radiation as well as safety relevant requirements and corresponding shielding concepts are discussed. New concepts for the layout and design of such shielding are presented. Focal point of this work will be the fundamental difference between conventional fission reactor shielding and the safety relevant issues of shielding from high-energy radiation. Key point of this paper is the safety assessment of shielding issues of high current accelerators, spallation targets and ADS-blanket systems as well as neutron scattering instruments at spallation neutron sources. Safety relevant shielding requirements are presented and discussed. For the layout and design of the shielding for spallation sources computer base calculations methods are used. A discussion and comparison of the most important methods like semi-empirical, deterministic and stochastic codes are presented. Another key point within the presented paper is the discussion of shielding materials and their shielding efficiency concerning different types of radiation. The use of recycling material, as a cost efficient solution, is discussed. Based on the conducted analysis, flowcharts for a systematic layout and design of adequate shielding for targets and accelerators have been developed and are discussed in this paper. By use of these flowcharts layout and engineering design of future ADS-facilities can be performed. (orig.)

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

    Institute of Scientific and Technical Information of China (English)

    王红艳; 吴宜灿; 何晓雄

    2002-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-01

    The DEMO specifications defined so as to ensure the consistency of the various blanket conceptual design studies performed within the framework of the European Test Blanket Programme have been recently updated. A very first attempt has been made to adapt the European Ceramic Breeder Inside-Tube DEMO blanket to these new specifications. Two solutions have been investigated. The first would ensure tritium self-sufficiency of the plant with a large safety margin. The other one, which fully preserves the design simplicity and reliability of the initial design, appears to be somewhat marginal from the tritium breeding capability point of view, but to offer good improvement prospects. (orig.).

  19. 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)

  20. Realization of a flat fission power density in a hybrid blanket over long operation periods

    International Nuclear Information System (INIS)

    A straightforward numerical graphical method is applied to achieve a flat fission power density (FPD) in a hybrid blanket by using a mixed fuel (ThO2 and natural UO2) with variable fractions of the fuel components in the radial direction. The neutronic analysis is carried out on a blanket with a hard neutron spectrum in the fissionable zone by simply omitting the moderating beryllium neutron multiplier. Mainly due to this precaution in the blanket design, the FPD could be kept quasi-constant over a relatively long plant lifetime

  1. Electrical behaviour of ceramic breeder blankets in pebble form after γ-radiation

    Directory of Open Access Journals (Sweden)

    E. Carella

    2015-07-01

    Full Text Available Lithium orthosilicate (Li4SiO4 ceramics in from of pebble bed is the European candidate for ITER testing HCPB (Helium Cooled Pebble Bed breeding modules. The breeder function and the shielding role of this material, represent the areas upon which attention is focused. Electrical measurements are proposed for monitoring the modification created by ionizing radiation and at the same time provide information on lithium movement in this ceramic structure. The electrical tests are performed on pebbles fabricated by Spray-dryer method before and after gamma-irradiation through a 60Co source to a fluence of 4.8 Gy/s till a total dose of 5 ∗ 105 Gy. The introduction of thermal annealing treatments during the electrical impedance spectroscopy (EIS measurements points out the recombination effect of the temperature on the γ-induced defects.

  2. Flexible shielding system for radiation protection

    Science.gov (United States)

    Babin, A.

    1972-01-01

    Modular construction of low cost flexible radiation shielding panels consists of water filled steels cans, zinc bromide windows, turntable unit, master-slave manipulators, and interlocking lead bricks. Easy modifications of shielding wall thicknesses are obtained by rearranging overall geometry of portable components.

  3. Neutron shielding heat insulation material

    International Nuclear Information System (INIS)

    Purpose: To improve decceleration and absorption of neutrons by incorporating neutron moderators and neutron absorbers in asbestos to thereby increase hydrogen concentration. Constitution: A mixture consisting of crysotile asbestos, surface active agent and water is well stirred and compounded to open the crysotile asbestos filaments and prepare a high viscosity slurry. After adding hydroxides such as magnesium hydroxide, hydrated salts such as magnesium borate hydrate or water containing minerals such as alumina cement hydrate, or boron compound to the slurry, the slurry is charged in a predetermined die, and dried and compressed to prepare shielding heat insulation products. The crysotile asbestos has 18 - 15 wt.% of water of crystallinity in the structure and contains a considerably high hydrogen concentration that acts as neutron moderators. (Kawakami, Y.)

  4. Shielding superconductors with thin films

    CERN Document Server

    Posen, Sam; Catelani, Gianluigi; Liepe, Matthias U; Sethna, James P

    2015-01-01

    Determining the optimal arrangement of superconducting layers to withstand large amplitude AC magnetic fields is important for certain applications such as superconducting radiofrequency cavities. In this paper, we evaluate the shielding potential of the superconducting film/insulating film/superconductor (SIS') structure, a configuration that could provide benefits in screening large AC magnetic fields. After establishing that for high frequency magnetic fields, flux penetration must be avoided, the superheating field of the structure is calculated in the London limit both numerically and, for thin films, analytically. For intermediate film thicknesses and realistic material parameters we also solve numerically the Ginzburg-Landau equations. It is shown that a small enhancement of the superheating field is possible, on the order of a few percent, for the SIS' structure relative to a bulk superconductor of the film material, if the materials and thicknesses are chosen appropriately.

  5. Radiation Shielding Systems Using Nanotechnology

    Science.gov (United States)

    Chen, Bin (Inventor); McKay, Christoper P. (Inventor)

    2011-01-01

    A system for shielding personnel and/or equipment from radiation particles. In one embodiment, a first substrate is connected to a first array or perpendicularly oriented metal-like fingers, and a second, electrically conducting substrate has an array of carbon nanostructure (CNS) fingers, coated with an electro-active polymer extending toward, but spaced apart from, the first substrate fingers. An electric current and electric charge discharge and dissipation system, connected to the second substrate, receives a current and/or voltage pulse initially generated when the first substrate receives incident radiation. In another embodiment, an array of CNSs is immersed in a first layer of hydrogen-rich polymers and in a second layer of metal-like material. In another embodiment, a one- or two-dimensional assembly of fibers containing CNSs embedded in a metal-like matrix serves as a radiation-protective fabric or body covering.

  6. Electromagnetic interference shielding effectiveness of monolayer graphene.

    Science.gov (United States)

    Hong, Seul Ki; Kim, Ki Yeong; Kim, Taek Yong; Kim, Jong Hoon; Park, Seong Wook; Kim, Joung Ho; Cho, Byung Jin

    2012-11-16

    We report the first experimental results on the electromagnetic interference (EMI) shielding effectiveness (SE) of monolayer graphene. The monolayer CVD graphene has an average SE value of 2.27 dB, corresponding to ~40% shielding of incident waves. CVD graphene shows more than seven times (in terms of dB) greater SE than gold film. The dominant mechanism is absorption rather than reflection, and the portion of absorption decreases with an increase in the number of graphene layers. Our modeling work shows that plane-wave theory for metal shielding is also applicable to graphene. The model predicts that ideal monolayer graphene can shield as much as 97.8% of EMI. This suggests the feasibility of manufacturing an ultrathin, transparent, and flexible EMI shield by single or few-layer graphene. PMID:23085718

  7. Results of shielding characteristics tests in Monju

    Energy Technology Data Exchange (ETDEWEB)

    Usami, Shin; Suzuoki, Zenro; Deshimaru, Takehide; Nakashima, Fumiaki [Japan Nuclear Cycle Development Inst., Tsuruga, Fukui (Japan)

    2001-06-01

    In the prototype fast breeder reactor Monju, the shielding characteristics tests were made around the reactor core, the primary heat transport system, and the fuel handling and storage system as a part of the system start-up tests from 0% to 45% of rated power from October 1993 through December 1995. The results of the measurements, analyses and evaluations in these tests validated the FBR shielding analysis methods and demonstrated that there was a safe shielding design margin in Monju. The important basic data for use in future FBR shielding design were successfully acquired. In order to obtain more substantial basic data and to improve the accuracy of the analyses, the next shielding measurements are planned for the period of the system start-up tests at the restart of Monju. (author)

  8. Radiation Shielding for Nuclear Thermal Propulsion

    Science.gov (United States)

    Caffrey, Jarvis A.

    2016-01-01

    Design and analysis of radiation shielding for nuclear thermal propulsion has continued at Marshall Space Flight Center. A set of optimization tools are in development, and strategies for shielding optimization will be discussed. Considerations for the concurrent design of internal and external shielding are likely required for a mass optimal shield design. The task of reducing radiation dose to crew from a nuclear engine is considered to be less challenging than the task of thermal mitigation for cryogenic propellant, especially considering the likely implementation of additional crew shielding for protection from solar particles and cosmic rays. Further consideration is thus made for the thermal effects of radiation absorption in cryogenic propellant. Materials challenges and possible methods of manufacturing are also discussed.

  9. 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

  10. Integrated-blanket-coil (IBC) concept applied to the OH-coil for spherical tori

    International Nuclear Information System (INIS)

    This concept combines blanket and coil functions into a single component. The objectives of the concept are to: (1) provide design options, (2) simplify overall configuration, (3) enhance compactness, and (4) reduce costs. Some drawings of the system are given

  11. 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 ...

  12. Normal Operation (NO) of APT Blanket System and its Components Based on Initial Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  13. 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.)

  14. 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)

  15. The use of nipple shields: A review

    Directory of Open Access Journals (Sweden)

    Selina eChow

    2015-10-01

    Full Text Available A nipple shield is a breastfeeding aid with a nipple-shaped shield that is positioned over the nipple and areola prior to nursing. Nipple shields are usually recommended to mothers with flat nipples or in cases in which there is a failure of the baby to effectively latch onto the breast within the first two days postpartum. The use of nipple shields is a controversial topic in the field of lactation. Its use has been an issue in the clinical literature since some older studies discovered reduced breast milk transfer when using nipple shields, while more recent studies reported successful breastfeeding outcomes. The purpose of this review was to examine the evidence and outcomes with nipple shield use. Methods: A literature search was conducted in Ovid MEDLINE, OLDMEDLINE, EMBASE Classic, EMBASE, Cochrane Central Register of Controlled Trials and CINAHL. The primary endpoint was any breastfeeding outcome following nipple shield use. Secondary endpoints included the reasons for nipple shield use and the average/median length of use. For the analysis, we examined the effect of nipple shield use on physiological responses, premature infants, mothers’ experiences, and health professionals’ experiences. Results: The literature search yielded 261 articles, 14 of which were included in this review. Of these 14 articles, three reported on physiological responses, two reported on premature infants, eight reported on mothers’ experiences, and one reported on health professionals’ experiences. Conclusion: Through examining the use of nipple shields, further insight is provided on the advantages and disadvantages of this practice, thus allowing clinicians and researchers to address improvements on areas that will benefit mothers and infants the most.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  17. Climate-driven expansion of blanket bogs in Britain during the Holocene

    Directory of Open Access Journals (Sweden)

    A. V. Gallego-Sala

    2015-10-01

    Full Text Available Blanket bog occupies approximately 6 % of the area of the UK today. The Holocene expansion of this hyperoceanic biome has previously been explained as a consequence of Neolithic forest clearance. However, the present distribution of blanket bog in Great Britain can be predicted accurately with a simple model (PeatStash based on summer temperature and moisture index thresholds, and the same model correctly predicts the highly disjunct distribution of blanket bog worldwide. This finding suggests that climate, rather than land-use history, controls blanket-bog distribution in the UK and everywhere else. We set out to test this hypothesis for blanket bogs in the UK using bioclimate envelope modelling compared with a database of peat initiation age estimates. We used both pollen-based reconstructions and climate model simulations of climate changes between the mid-Holocene (6000 yr BP, 6 ka and modern climate to drive PeatStash and predict areas of blanket bog. We compiled data on the timing of blanket-bog initiation, based on 228 age determinations at sites where peat directly overlies mineral soil. The model predicts large areas of northern Britain would have had blanket bog by 6000 yr BP, and the area suitable for peat growth extended to the south after this time. A similar pattern is shown by the basal peat ages and new blanket bog appeared over a larger area during the late Holocene, the greatest expansion being in Ireland, Wales and southwest England, as the model predicts. The expansion was driven by a summer cooling of about 2 °C, shown by both pollen-based reconstructions and climate models. The data show early Holocene (pre-Neolithic blanket-bog initiation at over half of the sites in the core areas of Scotland, and northern England. The temporal patterns and concurrence of the bioclimate model predictions and initiation data suggest that climate change provides a parsimonious explanation for the early Holocene distribution and later

  18. Polyimide based amorphous silicon solar modules

    Science.gov (United States)

    Jeffrey, Frank R.; Grimmer, Derrick P.; Martens, Steven A.; Abudagga, Khaled; Thomas, Michael L.; Noak, Max

    1993-01-01

    Requirements for space power are increasingly emphasizing lower costs and higher specific powers. This results from new fiscal constraints, higher power requirements for larger applications, and the evolution toward longer distance missions such as a Lunar or Mars base. The polyimide based a-Si modules described are being developed to meet these needs. The modules consist of tandem a-Si solar cell material deposited directly on a roll of polyimide. A laser scribing/printing process subdivides the deposition into discrete cell strips which are series connected to produce the required voltage without cutting the polymer backing. The result is a large, monolithic, blanket type module approximately 30 cm wide and variable in length depending on demand. Current production modules have a specific power slightly over 500 W/Kg with room for significant improvement. Costs for the full blanket modules range from $30/Watt to $150/Watt depending on quantity and engineering requirements. Work to date focused on the modules themselves and adjusting them for the AMO spectrum. Work is needed yet to insure that the modules are suitable for the space environment.

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

    OpenAIRE

    Lukwa, N; A. Makuwaza; T. Chiwade; S.L. Mutambu; M. Zimba; P. Munosiyei

    2013-01-01

    The effect of permethrin-treated Africa University (AU) mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes) and repellence (ability to prevent ≥80% of mosquito bites) properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up t...

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

    OpenAIRE

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Smith, D. L.

    1999-10-07

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

  2. Foam-Reinforced Polymer Matrix Composite Radiation Shields Project

    Data.gov (United States)

    National Aeronautics and Space Administration — New and innovative lightweight radiation shielding materials are needed to protect humans in future manned exploration vehicles. Radiation shielding materials are...

  3. Improved Metal-Polymeric Laminate Radiation Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this proposed Phase I program, a multifunctional lightweight radiation shield composite will be developed and fabricated. This structural radiation shielding...

  4. Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor

    Science.gov (United States)

    Ragheb, Magdi M. H.; Maynard, Charles W.

    1981-10-01

    A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components. To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×1015 n/(m2·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 1017 n/(m2·s) for a 1 MW/m2 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.

  5. Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

    1999-08-01

    Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

  6. Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Humrickhouse, Paul Weston [Idaho National Laboratory; Merrill, Brad Johnson [Idaho National Laboratory

    2014-11-01

    It is envisioned that tritium will be extracted from DCLL blankets using a vacuum permeator. We derive here an analytical solution for the extraction efficiency of a permeator tube, which is a function of only two dimensionless numbers: one that indicates whether radial transport is limited in the PbLi or in the solid membrane, and another that is the ratio of axial and radial transport times in the PbLi. The permeator efficiency is maximized by decreasing the velocity and tube diameter, and increasing the tube length. This is true regardless of the mass transport correlation used; we review several here and find that they differ little, and the choice of correlation is not a source of significant uncertainty here. The PbLi solubility, on the other hand, is a large source of uncertainty, and we identify upper and lower bounds from the literature data. Under the most optimistic assumptions, we find that a ferritic steel permeator operating at 550 °C will need to be at least an order of magnitude larger in volume than previous conceptual designs using niobium and operating at higher temperatures.

  7. Physics aspects of metal fuelled fast reactors with thorium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Mohapatra, D.K., E-mail: dina@igcar.gov.in; Singh, S.S.; Riyas, A.; Mohanakrishnan, P.

    2013-12-15

    Metal fuelled fast breeder reactors (MFBR) with high breeding ratio will play a major role in meeting the high nuclear power growth envisaged in India. In this regard several conceptual reactor designs with alloys of U–Pu–Zr fuel have been suggested for commercial operations. This study focusses on the physics design aspects of a sodium cooled U–Pu–6%Zr fuelled 1000 MWe fast breeder reactor, which can attain a breeding ratio of nearly 1.5. The calculation results on reactor kinetics and safety parameters of the 1000 MWe MFBR are presented. The changes in the breeding ratio by introduction of thorium in the blankets of the MFBR are also investigated. Burnup analyses are carried out to compare the core burnup effects in MOX and metal fuelled FBRs. Since the MOX fuelled 500 MWe prototype fast breeder is getting constructed at IGCAR, for burnup comparisons a MFBR of similar design is considered. The results of this study indicate that the loss of reactivity in the metal core with burnup is less than half that of a MOX core and its breeding ratio remains nearly constant. It is also found that the isotopic composition of plutonium (Pu-vector composition) remains more steady with burnup in a metal core.

  8. Surface property variations in Venusian fluidized ejecta blanket craters

    Science.gov (United States)

    Johnson, Jeffrey R.; Baker, Victor R.

    1994-01-01

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

  9. SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-06

    The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.

  10. SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [ORNL

    2016-06-01

    The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.

  11. Extraterrestrial Regolith Derived Atmospheric Entry Heat Shields

    Science.gov (United States)

    Hogue, Michael D.; Mueller, Robert P.; Sibille, Laurent; Hintze, Paul E.; Rasky, Daniel J.

    2016-01-01

    High-mass planetary surface access is one of NASAs technical challenges involving entry, descent and landing (EDL). During the entry and descent phase, frictional interaction with the planetary atmosphere causes a heat build-up to occur on the spacecraft, which will rapidly destroy it if a heat shield is not used. However, the heat shield incurs a mass penalty because it must be launched from Earth with the spacecraft, thus consuming a lot of precious propellant. This NASA Innovative Advanced Concept (NIAC) project investigated an approach to provide heat shield protection to spacecraft after launch and prior to each EDL thus potentially realizing significant launch mass savings. Heat shields fabricated in situ can provide a thermal-protection system for spacecraft that routinely enter a planetary atmosphere. By fabricating the heat shield with space resources from materials available on moons and asteroids, it is possible to avoid launching the heat-shield mass from Earth. Regolith has extremely good insulating properties and the silicates it contains can be used in the fabrication and molding of thermal-protection materials. In this paper, we will describe three types of in situ fabrication methods for heat shields and the testing performed to determine feasibility of this approach.

  12. Carbon nanostructure composite for electromagnetic interference shielding

    Indian Academy of Sciences (India)

    Anupama Joshi; Suwarna Datar

    2015-06-01

    This communication reviews current developments in carbon nanostructure-based composite materials for electromagnetic interference (EMI) shielding. With more and more electronic gadgets being used at different frequencies, there is a need for shielding them from one another to avoid interference. Conventionally, metal-based shielding materials have been used. But due to the requirement of light weight, corrosion resistive materials, lot of work is being done on composite materials. In this research the forerunner is the nanocarbon-based composite material whose different forms add different characteristics to the composite. The article focusses on composites based on graphene, graphene oxide, carbon nanotubes, and several other novel forms of carbon.

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

    Science.gov (United States)

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

    2013-09-01

    It is postulated that up-flow anaerobic sludge blanket (UASB) reactor efficiency can be enhanced by a membrane immersed in the reactor to operate it as an anaerobic membrane bioreactor (AnMBR) for low-strength wastewater treatment. This postulate was tested by comparing the performance with and without a hollow fiber microfiltration membrane module immersed in UASB reactors operated at two specific organic loading rates (SOLR). Results showed that membrane filtration enhanced process performance and stability, with over 90% total organic carbon (TOC) removal consistently achieved. More than 91% of the TOC removal was achieved by suspended biomass, while less than 6% was removed by membrane filtration and digestion in the membrane attached biofilm during stable AnMBRs operation. Although the membrane and its biofilm played an important role in initial stage of the high SOLR test, linear increased TOC removal by bulk sludge mainly accounted for the enhanced process performance, implying that membrane led to enhanced biological activity of the suspended sludge. The high retention of active fine sludge particles in suspension was the main reason for this significant improvement of performance and biological activity, which led to decreased SOLR with time to a theoretical optimal level around 2  g COD/g MLVSS·d and the establishment of a microbial community dominated by Methanothrix-like microbes. It was concluded that UASB process performance can be enhanced by transforming such to AnMBR operation when the loading rate is too high for sufficient sludge retention, and/or when the effluent water quality demands are especially stringent. PMID:23578587

  14. Measurement of Velocity Profiles in a scaled Transparent Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Korea has developed two kinds of TBM for ITER; a Helium cooled solid breeder (HCSB) TBM and a Helium cooled molten lithium (HCML) TBM, respectively. Under the HCML TBM project, a 1/6 scaled mock-up of the TBM FW has been fabricated in Korea Atomic Energy Research Institute (KAERI). The size of the scaled mock-up is 260 mm height and 444 mm width. As coolant channels in the mock-up, there are rectangular shape of 10 channels with 10 mm height and 20 mm width. The scaled mock-up was manufactured by hot isostatic pressing bonding method using SS316L. Three components of the scaled mock-up were prepared; a front part of cooling channel 10 mm height with 20 mm width, a front cover plate, and a back plate. The front plate and the cover were bonded by welding, and the welded part and the back plate are attached by HIP process. A pair of manifolds, to distribute the coolant uniformly into 10 channels of the scaled mock-up, were designed and fabricated. The designed manifolds were then welded in inlet and outlet positions of the mock-up. To measure the flow distribution in each channel, the ultrasonic flowmeter (UFM) was used and the values were compared to a conventional flowmeter. Before the flow distribution test of the scaled mock-up, a calibration procedure was conducted with a single channel mock-up using the UFM and the flowmeter. The result showed a good agreement between the UFM and the flowmeter values in the single channel. The same test procedure conducted on the scaled mock-up; the velocity of each channel was measured by the UFM and total mass flow rate was measured with the flowmeter. The estimated velocities distributed from the manifold were simulated by ANSYS-CFX. However, there was a discrepancy between the measured and the simulated values. The current manifold could not provide uniform flow rate to the each channel or there would be a measurement error using the UFM in the specified mock-up. This means that the UFM measurement method should be validated with other analysis tools, such as particle image velocimetry (PIV) system and differential pressure (DP) transmitters. Therefore, a scaled transparent TBM facility was manufactured to validate the application of the UFM, by comparing the PIV system.

  15. Low activation steels welding with PWHT and coating for ITER test blanket modules and DEMO

    Science.gov (United States)

    Aubert, P.; Tavassoli, F.; Rieth, M.; Diegele, E.; Poitevin, Y.

    2011-02-01

    EUROFER weldability is investigated in support of the European material properties database and TBM manufacturing. Electron Beam, Hybrid, laser and narrow gap TIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel developed in Europe. These welding processes produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The fusion zones are typically composed of martensite laths, with small grain sizes. In the heat-affected zones, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. PWHT developments have driven to a one-step PWHT (750 °C/3 h), successfully applied to joints restoring good material performances. It will produce less distortion levels than a full austenitization PWHT process, not really applicable to a complex welded structure such as the TBM. Different tungsten coatings have been successfully processed on EUROFER material. It has shown no really effect on the EUROFER base material microstructure.

  16. Three-dimensional nuclear analysis for the US dual coolant lead lithium ITER test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Sawan, M.E., E-mail: sawan@engr.wisc.edu [Fusion Technology Institute, University of Wisconsin, 1500 Engineering Dr., Madison, WI 53706 (United States); Smith, B.; Marriott, E.P.; Wilson, P.P.H. [Fusion Technology Institute, University of Wisconsin, 1500 Engineering Dr., Madison, WI 53706 (United States)

    2010-12-15

    Detailed 3-D neutronics calculations have been performed for the US DCLL TBM. The neutronics calculations were performed directly in the CAD model using the DAG-MCNP code that allows preserving the geometrical details. Detailed high-resolution, high-fidelity profiles of the nuclear parameters were generated using fine mesh tallies. These included tritium production, nuclear heating, and radiation damage. The TBM heterogeneity, exact source profile, and inclusion of the surrounding frame and other in-vessel components result in lower TBM nuclear parameters compared to the previous 1-D predictions. This work clearly demonstrates the importance of preserving geometrical details in nuclear analyses of geometrically complex components in fusion systems.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  18. Six-Party Qualification Program of FW Fabrication Methods for ITER Blanket Module Procurement

    International Nuclear Information System (INIS)

    In December 2005, the new procurement allocation plan of the ITER components among the seven Parties was prepared. The need to qualify for procurement of the specific components was especially introduced in the document. The main features and milestones of the qualification program are described in '' Procurement Plan '' for each specific component. The management rules for cases of failure in the qualification are also documented in the Procurement Plan. Due to the complicated features of FW procurement (by 6 Parties: CN, EU, JA, KO, RF and US), the procurement document has to be developed precisely. To guarantee high quality of 1700 FW panels produced by 6 different Parties, a qualification program is essential. The qualification mock-up is 80 mm wide, 240 mm long and 81 mm thick with 3 beryllium tiles 10 mm thick. Three identical mock-ups will be fabricated by each of the 6 Parties in 2006-2007 with the same method as for the ITER first wall panels. Heat load tests will be performed on the qualification mock-ups in 2007 in EU and USA facilities. During the testing, the coolant velocity is reduced to 1-2 m/s instead of 4.5 m/s to simulate the nuclear heating. The cycle time of the test will be ∼ 4 minutes. When the heat flux is increased by ∼ 40 %, the cycle time can be ∼90 sec according to thermal and stress analysis. The maximum design heat load on the ITER FW is 0.5 MW/m2 (steady state) x 30,000 shots. The heat load due to NB shine-through to achieve H-mode plasma during the hydrogen phase will be 1 MW/m2 in a certain area of the outboard equatorial region (total 1,000 shots for 2.5 years). The maximum heat flux due to MARFE: 0.5-1.4 MW/m2 (up to 10 sec duration) also needs to be taken into account in the heat load test conditions. Mechanical tests of joints are also required using standardized methods. Only Parties which have satisfied the acceptance criteria of the qualification tests can proceed to the procurement stage of the ITER FW. Semi-prototypes (roughly 1000 mm x 200 mm) are also requested before the ITER FW manufacturing. As a part of the qualification program, Parties have proposed '' new materials '' to be qualified, such as new beryllium materials, cast CuCrZr, powder HIPed CuCrZr and 316 L(N). New beryllium materials shall be tested together with the reference material S65C for comparison. Irradiation tests of the new beryllium materials will be required (unless very similar materials are already irradiated). (author)

  19. Review of candidate welding processes of RAFM steels for ITER test blanket modules and DEMO

    Science.gov (United States)

    Aubert, P.; Tavassoli, F.; Rieth, M.; Diegele, E.; Poitevin, Y.

    2011-10-01

    EUROFER weldability is investigated in support of the European TBM manufacturing. Electron beam, hybrid, laser and NGTIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel. It is shown that the most promising processes are laser, electron beam and hybrid welding, depending on the section size and accessibility. They produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The FZ are typically composed of martensite laths, with small grain sizes. In the HAZ, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. A one step PWHT (750 °C/3 h) is successfully applied to joints restoring good material performance. Distortion levels, with and without PWHT, are controlled through adaptation of manufacturing steps and clamping devices, obtaining levels not exceeding 120 μm (+/-60 μm) on a full "one cell mock-up".

  20. Activation analyses for the Korea helium cooled ceramic reflector test blanket module

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    The activation analyses were performed to obtain activities, inventories and dose rates after shutdown for the Korea HCCR TBM. MonteBurns code coupled with MCNP and CINDER codes was used in the activation calculation and the irradiation history based on the SA2 scenario was applied. The total activity in TBM after shutdown was evaluated as a value of 4.12 × 10{sup 16} Bq. Dose rates after shutdown from the activated HCCR TBM was estimated. The decay gamma-ray spectra in each region inside TBM were estimated based on the calculated activities. The dose rate at 0 cm–3 m from the FW surface of HCCR TBM was evaluated according to the cooling time of 1 day, 1 week, 1 month, 6 months and 1 year. The dose rate at 0 cm from the FW surface was evaluated to be 575.84 Sv/h at 1 day after shutdown.

  1. 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

  2. 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.)

  3. Shielding design for better plant availability

    International Nuclear Information System (INIS)

    Design methods are described for providing a shield system for nuclear power plants that will facilitate maintenance and inspection, increase overall plant availability, and ensure that man-rem exposures are as low as practicable

  4. Shielded ADR Magnets For Space Applications Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Phase II program will concentrate on manufacturing of qualified low-current, light-weight, 10K ADR magnets for space application. Shielded ADR solenoidal...

  5. Long Duration Space Shelter Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Physical Sciences Inc. (PSI) has developed fiber reinforced ceramic composites for radiation shielding that can be used for external walls in long duration manned...

  6. Long Duration Space Shelter Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Physical Sciences Inc. (PSI) has developed a ceramic composite material system that is more effective for shielding both GCR and SPE than aluminum. The composite...

  7. Passive Magnetic Shielding in Gradient Fields

    CERN Document Server

    Bidinosti, C P

    2013-01-01

    The effect of passive magnetic shielding on dc magnetic field gradients imposed by both external and internal sources is studied. It is found that for concentric cylindrical or spherical shells of high permeability material, higher order multipoles in the magnetic field are shielded progressively better, by a factor related to the order of the multipole. In regard to the design of internal coil systems for the generation of uniform internal fields, we show how one can take advantage of the coupling of the coils to the innermost magnetic shield to further optimize the uniformity of the field. These results demonstrate quantitatively a phenomenon that was previously well-known qualitatively: that the resultant magnetic field within a passively magnetically shielded region can be much more uniform than the applied magnetic field itself. Furthermore we provide formulae relevant to active magnetic compensation systems which attempt to stabilize the interior fields by sensing and cancelling the exterior fields clos...

  8. Shield structure for a nuclear reactor

    International Nuclear Information System (INIS)

    An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

  9. Electromagnetic shielding with polypyrrole-coated fabrics

    OpenAIRE

    Avloni, J.; L. De Florio; Henn, A. R.; R. Lau; Ouyang, M.; Sparavigna, A.

    2006-01-01

    Several shielding applications, to protect human health and electronic devices against dangerous effects of electromagnetic radiation, require solutions that fabrics can suitably fulfill. Here, we will investigate the electromagnetic interference shielding effectiveness of polypyrrole-coated polyester textiles, in the frequency range 100-1000 MHz. Insertion losses for several conductive fabrics with different surface resistivity ranging from 40 Ohm till the very low value of 3 Ohm were evalua...

  10. Shielding Design for a Medical Cyclotron

    Institute of Scientific and Technical Information of China (English)

    WANG; Feng; SONG; Guo-fang; GUAN; Feng-ping; LV; Yin-long; ZHANG; Xing-zhi

    2012-01-01

    <正>A 10 MeV 100 μA medical cyclotron is constructed at CIAE which is used in the production of FDG. The energy of the cyclotron can reach 14 MeV by adjusting the magnetic field and RF system parameters, and the shielding design is in accordance with the 14 MeV beam energy. In this shielding design only neutron is considered, and the neutron source is produced by proton

  11. APOLLO2 code self-shielding formalism

    International Nuclear Information System (INIS)

    This report describes the various self-shielding methods used in the APOLLO2 code for treating one resonant nucleus or a mixture of resonant nuclei. The methods are expounded in chronological order. First of all, the methods dealing with one resonant isotope are explained. Then an original method dealing directly with a resonant mixture is detailed. This new method is also convenient for one resonant nucleus and leads, in that case, to interesting improvements in the self-shielding modeling. (author)

  12. Foamed Nanocomposites for EMI Shielding Applications

    OpenAIRE

    Molenberg, Isabel; Huynen, Isabelle; Baudouin, Anne-Christine; Bailly, Christian; Thomassin, Jean-Michel; Detrembleur, Christophe

    2010-01-01

    Throughout this chapter, we have seen that foamed polymer/CNTs nanocomposites make very good EMI shielding materials. They exhibit a high conductivity and a relatively low dielectric constant, leading to a high Shielding Effectiveness and a relatively low Reflectivity, although a compromise in CNT content must be found between a high SE and a low R. This was confirmed by measurement results and rheological measurements but also using a simple electrical equivalent model.

  13. Evaluation of tube shielding; Utvaerdering av tubskyddsmaterial

    Energy Technology Data Exchange (ETDEWEB)

    Hjoernhede, Anders; Westberg, Stig-Bjoern; Henderson, Pamela; Wetterstroem, Jonas; Jonasson, Anna

    2007-12-15

    Problems with soot-blowing have increased recently because of the poor fuel quality. Studies show that removing all the deposit by soot-blowing increases the metal loss of the superheaters, which drastically shortens component lifetimes. A simple, effective and common way of increasing the lifetime is to use tube shielding. Austenitic stainless steels seem to be the type of material most commonly used for tube shielding. It is thought that they give better protection against material removal than ferritic steels, but the cost of austenitics is several times greater than ferritic steels. It is clear that there is a significant economic advantage in choosing the right material for tube shielding, even though it might be expected that the cheaper materials do not perform as well as the more expensive ones. The reason for the study reported here is that very little material data exists in the literature. Few, if any tests have been performed to study the choice of material for tube shielding. The goal was to compare and evaluate a number of materials in a boiler to see if it is possible to replace the shielding material presently used with cheaper alternatives. About a dozen different shielding materials were installed and exposed for 4000 hours on primary- and secondary superheaters in a waste-fired boiler in Norrkoeping (Haendeloe Boiler 14.75MW). In total, 130 m of test material were installed and measured in several positions: a least 150 thickness measurements, before and after, were made on every tube shield. The results showed that the greatest attack was found on the secondary superheater shielding, where both the gas- and steam temperatures were higher. When considering cost and lifetime Sicromal 10 and 12 (however not Sicromal 8) and 15Mo3 are recommended as being better than 253 MA. The results should be of interest to most plants firing biomass or waste

  14. Influence of Shielding Arrangement on ECT Sensors

    Directory of Open Access Journals (Sweden)

    J. L. Fernandez Marron

    2006-09-01

    Full Text Available This paper presents a full 3D study of a shielded ECT sensor. The spatialresolution and effective sensing field are obtained by means of Finite Element Methodbased simulations and are the compared to a conventional sensor's characteristics. Aneffective improvement was found in the sensitivity in the pipe cross-section, resulting inenhanced quality of the reconstructed image. The sensing field along the axis of the sensoralso presents better behaviour for a shielded sensor.

  15. Measurement accuracy in shielded magnetic fields

    International Nuclear Information System (INIS)

    The measurement error due to both the probe size averaging effect and the coil arrangement is investigated when magnetic field measurements are performed in close proximity to different planar shields. The analysis is carried on through a hybrid FEM/BEM model which employs the 'thin shield' technique. Ferromagnetic, pure conductive and multilayer screens are taken into consideration and an estimation of the errors for concentric and non-concentric coil probes is given. The numerical results are validated by experiments

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Shielding Effectiveness of Composites Containing Flaky Inclusions

    Institute of Scientific and Technical Information of China (English)

    WANG Qingguo; QU Zhaoming; WANG Yilong

    2013-01-01

    To investigate the quantitative relationship between the electromagnetic-shielding property of composites and the distribution of inclusions,a scheme for predicting the shielding effectiveness of composites containing variously-distributed flaky inclusions is proposed.The scheme is based on equivalent parameters of homogeneous comparison materials and the plane-wave shielding theory.It leads to explicit formulas for the shielding effectiveness of multi-layered composites in terms of microstructural parameters that characterize the shape,distribution and orientation of the inclusions.For single layer composite that contains random and aligned flaky silver-coated carbonyl-iron particles with fractions of different volume,the predicted shielding effectiveness agrees well with the experimental data.As for composites containing aligned flaky particles,the shielding effectiveness obtained by the proposed scheme and experiment data is higher than that the random case,e.g.about 20 dB higher at 750 MHz.The proposed scheme is a straightforward method for optimizing future composite designs.

  19. Reactor shielding. Report of a panel

    International Nuclear Information System (INIS)

    Reactor shielding is necessary that people may work and live in the vicinity of reactors without receiving detrimental biological effects and that the necessary materials and instrumentation for reactor operation may function properly. Much of the necessary theoretical work and experimental measurement has been accomplished in recent years. Scientists have developed some very sophisticated methods which have contributed to a more thorough understanding of the problems involved and have produced some very reliable results leading to significant reductions in shield configurations. A panel of experts was convened from 9 to 13 March 1964 in Vienna at the Headquarters of the International Atomic Energy Agency to discuss the present status of reactor shielding. The participants were prominent shielding experts from most of the laboratories engaged in this field throughout the world. They presented status reports describing the past history and plans for further development of reactor shielding in their countries and much valuable discussion took place on some of the most relevant aspects of reactor shielding. All this material is presented in this report, together with abstracts of the supporting papers read to the Panel

  20. Corrugation Stuffed Shield for Spacecraft and Its Performance

    Institute of Scientific and Technical Information of China (English)

    LIU You-ying; WANG Hai-fu

    2006-01-01

    A corrugation stuffed shield system protecting spacecrafts against meteoroid and orbital debris (M/OD) is presented. The semi-empirical ballistic limit equations (BLEs)defining the protection capability of the shield system are given, an d the shielding performance is also discussed. The corrugation stuffed shield (CSS) is more effective than stuffed Whipple shield for M/OD protection,and its shielding performance will be improved significantly as increasing the impact angle. Orbital debris up to 1cm in diameter can be shielded effectively as increasing the impact angle to 25° at the corrugated angle of 30°. The results are significant to spacecraft design.

  1. 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)

  2. Thermal Analysis on Conceptual K-DEMO Breeding Blanket on parallel flow configuration

    International Nuclear Information System (INIS)

    The feasibility study consists the design guidelines and requirements for the K-DEMO(the Korean Fusion DEMOnstration reactor). It is possible to design flexible and realistic concepts of the demonstration fusion power plant. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the breeding blanket for the K-DEMO reactor. Recently, a new breeding blanket concept has been proposed for the K-DEMO reactor, and preliminary feasibility studies are actively ongoing. Design concept of parallel plate-type blanket of K-DEMO and thermal limits on components was identified, in this study. It was concluded that an acceptable thermal design was achieved in the proposed breeding blanket design. However, some design improvements in the geometry of the breeding blanket are ongoing. In aspect of pressure drop and outlet temperature, some analysis need to be conducted. Stage of study on the K-DEMO reactor is in the preliminary concept definition

  3. The integrated-blanket-coil concept applied to the spherical torus

    International Nuclear Information System (INIS)

    The purpose of this study is to investigate the potential of reactor embodiments based on the combination of two novel concepts, the integrated-blanket-coil (IBC) concept and the spherical torus (ST) concept. The IBC concept involves the combination of blanket and coil functions into a single component. The ST concept is based on the operation of the tokamak magnetic configuration at low aspect ratio. Two applications of the IBC concept to the ST are presented: (1) the IBC as the outer blanket and TF coil return legs (Application 1); and (2) the IBC as the inner blanket and OH solenoid coil set (Application 2). The application 1 IBC/ST yields reactor embodiments operating with high mass power density (≅500 kWe/MT) while, at the same time, yielding moderate neutron wall loading (≅5 MW/m/sup 2/) and modest electrical output (400-500 MWe). It is anticipated that fusion reactors operating in the above parameter space will exhibit attractive economic potential. When compared with a conventional ST design (copper TF coils and separate blanket), the Aplication 1 IBC/ST design exhibits a factor of two improvement in mass power density. The Application 2 IBC/ST provides an inductive current drive option which yields burn times in the range 3-4 minutes with no penalties in inboard breeding and energy recovery

  4. Retro Rocket Motor Self-Penetrating Scheme for Heat Shield Exhaust Ports

    Science.gov (United States)

    Marrese-Reading, Colleen; St.Vaughn, Josh; Zell, Peter; Hamm, Ken; Corliss, Jim; Gayle, Steve; Pain, Rob; Rooney, Dan; Ramos, Amadi; Lewis, Doug; Shepherd, Joe; Inaba, Kazuaki

    2009-01-01

    A preliminary scheme was developed for base-mounted solid-propellant retro rocket motors to self-penetrate the Orion Crew Module heat shield for configurations with the heat shield retained during landings on Earth. In this system the motors propel impactors into structural push plates, which in turn push through the heat shield ablator material. The push plates are sized such that the remaining port in the ablator material is large enough to provide adequate flow area for the motor exhaust plume. The push plate thickness is sized to assure structural integrity behind the ablative thermal protection material. The concept feasibility was demonstrated and the performance was characterized using a gas gun to launch representative impactors into heat shield targets with push plates. The tests were conducted using targets equipped with Fiberform(R) and PICA as the heat shield ablator material layer. The PICA penetration event times were estimated to be under 30 ms from the start of motor ignition. The mass of the system (not including motors) was estimated to be less than 2.3 kg (5 lbs) per motor. The configuration and demonstrations are discussed.

  5. Space Shielding Materials for Prometheus Application

    Energy Technology Data Exchange (ETDEWEB)

    R. Lewis

    2006-01-20

    At the time of Prometheus program restructuring, shield material and design screening efforts had progressed to the point where a down-selection from approximately eighty-eight materials to a set of five ''primary'' materials was in process. The primary materials were beryllium (Be), boron carbide (B{sub 4}C), tungsten (W), lithium hydride (LiH), and water (H{sub 2}O). The primary materials were judged to be sufficient to design a Prometheus shield--excluding structural and insulating materials, that had not been studied in detail. The foremost preconceptual shield concepts included: (1) a Be/B{sub 4}C/W/LiH shield; (2) a Be/B{sub 4}C/W shield; (3) and a Be/B{sub 4}C/H{sub 2}O shield. Since the shield design and materials studies were still preliminary, alternative materials (e.g., {sup nal}B or {sup 10}B metal) were still being screened, but at a low level of effort. Two competing low mass neutron shielding materials are included in the primary materials due to significant materials uncertainties in both. For LiH, irradiation-induced swelling was the key issue, whereas for H{sub 2}O, containment corrosion without active chemistry control was key, Although detailed design studies are required to accurately estimate the mass of shields based on either hydrogenous material, both are expected to be similar in mass, and lower mass than virtually any alternative. Unlike Be, W, and B{sub 4}C, which are not expected to have restrictive temperature limits, shield temperature limits and design accommodations are likely to be needed for either LiH or H{sub 2}O. The NRPCT focused efforts on understanding swelting of LiH, and observed, from approximately fifty prior irradiation tests, that either casting ar thorough out-gassing should reduce swelling. A potential contributor to LiH swelling appears to be LiOH contamination due to exposure to humid air, that can be eliminated by careful processing. To better understand LiH irradiation performance and

  6. Progress and prospects of calculation methods for radiation shielding

    International Nuclear Information System (INIS)

    Progress in calculation methods for radiation shielding are reviewed based on the activities of research committees related to radiation shielding fields established in the Atomic Energy Society of Japan. A technological roadmap for the field of radiation shielding; progress and prospects for specific shielding calculation methods such as the Monte Carlo, discrete ordinate Sn transport, and simplified methods; and shielding experiments used to validate calculation methods are presented in this paper. (author)

  7. Numerical Models for the Study of Electromagnetic Shielding

    Directory of Open Access Journals (Sweden)

    POPA Monica

    2012-10-01

    Full Text Available The paper presents 2D and 3D models for the study of electromagnetic shielding of a coil. The magnetic fields are computed for defining the shielding effectiveness. Parametrized numerical studies were performed in order to established the influence of shield thickness and height on magnetic field in certain points located in the exterior of coil – shield setup and on induced power within the shield.

  8. Shielding integrity testing of radioactive material transport packaging

    International Nuclear Information System (INIS)

    Although this Code of Practice is intended primarily to cover shielding integrity test requirements for off-site shielded radioactive material transport packaging, it may also be partly applicable to containers and specialised handling equipment (e.g. fuelling machines) used only on site, and to radiation shielding generally. The code is not concerned with proving adequacy of shielding design or with its absolute shielding value. (author)

  9. The vacuum vessel thermal shield of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, B.J. E-mail: bjyoon@kaeri.re.kr; In, S.R.; Cho, S.Y

    2003-09-01

    The Korea superconducting tokamak advanced research (KSTAR) tokamak has an all-superconductor magnet system and needs a thermal shield to cut off thermal radiation from the components of room temperature. The vacuum vessel thermal shield (VVTS) cooled to 70 K is placed in the narrow gap between the 5 K TF magnets and the 300 K vacuum vessel (VV). The VVTS is designed to be divided into 16 assembly modules of 22.5 deg. sector, each unit has an electrical insulation along the center line in the toroidal direction and four insulations in the poloidal direction to reduce eddy currents induced during plasma operations. All connections are bolted. The VVTS becomes consequently a rigid torus composed of 64 electrically insulated pieces. A key point of designing the VVTS is that supports of the VVTS are to be flexible enough to allow thermal constriction during cooling down to 70 K as well as sufficiently strong to withstand electromagnetic (EM) forces exerted on the VVTS during plasma disruptions. Leaf spring type supports devised to satisfy these requirements are to be installed along the mid plane of the VVTS. The cryopanel of the VVTS is of quilted plate type whose total thickness is 12 mm, cooled by 60 K, 20 bar GHe.

  10. High temperature blankets for non-electrical/electrical applications of fusion reactors: Annual report, [1983

    International Nuclear Information System (INIS)

    During FY '83 the Li2O solid-breeder, helium-cooled canister blanket emerged as the LLNL-UW choice for driving the low-temperature (2, high-temperature outer zone for driving the GA hydrogen synfuel process. Providing 3-dimensional neutronics analysis of power deposition and tritium breeding in both blankets was an important part of the UW-Rowe and Assoc. work. In both the LLNL-UW and MARS studies, the fusion driver as the Axi-Cell, A-cell version of the tandem mirror reactor (TMR). Physics parameters consistent with the synfuel interface were determined as part of the work. Defining and analyzing the thermal-electric interfaces between the TMR and the synfuel process continues to be of prime importance. The analysis of thermal transport and energy conversion in the interface, as well as thermal hydraulics analysis of the blanket, were part of the UW-Rowe Assoc. work

  11. Tritium recovery in Pb17Li-water cooled blanket systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  12. Analytic Ballistic Performance Model of Whipple Shields

    Science.gov (United States)

    Miller, J. E.; Bjorkman, M. D.; Christiansen, E. L.; Ryan, S. J.

    2015-01-01

    The dual-wall, Whipple shield is the shield of choice for lightweight, long-duration flight. The shield uses an initial sacrificial wall to initiate fragmentation and melt an impacting threat that expands over a void before hitting a subsequent shield wall of a critical component. The key parameters to this type of shield are the rear wall and its mass which stops the debris, as well as the minimum shock wave strength generated by the threat particle impact of the sacrificial wall and the amount of room that is available for expansion. Ensuring the shock wave strength is sufficiently high to achieve large scale fragmentation/melt of the threat particle enables the expansion of the threat and reduces the momentum flux of the debris on the rear wall. Three key factors in the shock wave strength achieved are the thickness of the sacrificial wall relative to the characteristic dimension of the impacting particle, the density and material cohesion contrast of the sacrificial wall relative to the threat particle and the impact speed. The mass of the rear wall and the sacrificial wall are desirable to minimize for launch costs making it important to have an understanding of the effects of density contrast and impact speed. An analytic model is developed here, to describe the influence of these three key factors. In addition this paper develops a description of a fourth key parameter related to fragmentation and its role in establishing the onset of projectile expansion.

  13. Shielding design to obtain compact marine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Akio; Sako, Kiyoshi (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1994-06-01

    The marine reactors equipped in previously constructed nuclear ships are in need of the secondary shield which is installed outside the containment vessel. Most of the weight and volume of the reactor plants are occupied by this secondary shield. An advanced marine reactor called MRX (Marine Reactor X) has been designed to obtain a more compact and lightweight marine reactor with enhanced safety. The MRX is a new type of marine reactor which is an integral PWR (The steam generator is installed in the pressure vessel.) with adopting a water-filled containment vessel and a new shielding design method of no installation of the secondary shield. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the Nuclear Ship MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor. The shielding design calculation was made using the ANISN, DOT3.5, QAD-CGGP2 and ORIGEN codes. The computational accuracy was confirmed by experimental analyses. (author).

  14. Asphalt as biological shielding against fusion neutrons

    International Nuclear Information System (INIS)

    For fusion experiments, thick biological radiation protection shields are necessary due to the deep penetration capability of the 14 MeV neutrons. A (D,T) neutron generator with a moderate output of around 1012 n/sec requires a concrete shielding of a wall thickness of 2 meters laterally and at the top of an experimental assembly. The cost for this biological shield may exceed the cost for most of the equipment for a fusion and/or hybrid experimental installation. Particularly, in Saudi Arabia, asphalt is very cheap and available in bulk quantities. As it is rich in hydrogen and carbon, it is worthwhile to investigate its shielding potential against fusion neutron. In the present work different biological shield configurations of asphalt at the wall of the experimental cavity for a research program being undertaken in Saudi Arabia, are investigated. The experimental cavity is approximated by a sphere of 5 meters radius. The yield of the neutron generator is taken as 1012 - 14 MeV - neutron/sec

  15. Shielding Development for Nuclear Thermal Propulsion

    Science.gov (United States)

    Caffrey, Jarvis A.; Gomez, Carlos F.; Scharber, Luke L.

    2015-01-01

    Radiation shielding analysis and development for the Nuclear Cryogenic Propulsion Stage (NCPS) effort is currently in progress and preliminary results have enabled consideration for critical interfaces in the reactor and propulsion stage systems. Early analyses have highlighted a number of engineering constraints, challenges, and possible mitigating solutions. Performance constraints include permissible crew dose rates (shared with expected cosmic ray dose), radiation heating flux into cryogenic propellant, and material radiation damage in critical components. Design strategies in staging can serve to reduce radiation scatter and enhance the effectiveness of inherent shielding within the spacecraft while minimizing the required mass of shielding in the reactor system. Within the reactor system, shield design is further constrained by the need for active cooling with minimal radiation streaming through flow channels. Material selection and thermal design must maximize the reliability of the shield to survive the extreme environment through a long duration mission with multiple engine restarts. A discussion of these challenges and relevant design strategies are provided for the mitigation of radiation in nuclear thermal propulsion.

  16. 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...

  17. Use of Ball Blanket in attention-deficit/hyperactivity disorder sleeping problems

    DEFF Research Database (Denmark)

    Hvolby, Allan; Bilenberg, Niels

    2011-01-01

    Objectives: Based on actigraphic surveillance, attention-deficit/hyperactivity disorder (ADHD) symptom rating and sleep diary, this study will evaluate the effect of Ball Blanket on sleep for a sample of 8-13-year-old children with ADHD. Design: Case-control study. Setting: A child and adolescent...... psychiatric department of a teaching hospital. Participants: 21 children aged 8-13 years with a diagnosis of ADHD and 21 healthy control subjects. Intervention: Sleep was monitored by parent-completed sleep diaries and 28 nights of actigraphy. For 14 of those days, the child slept with a Ball Blanket. Main...

  18. 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

  19. 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...

  20. 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)

  1. 77 FR 66597 - Chevron U.S.A. Inc.; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-11-06

    .... in Cameron Parish, Louisiana. Chevron states that DOE/FE has issued a number of blanket... U.S.A. Inc.; Application for Blanket Authorization To Export Previously Imported Liquefied Natural Gas on a Short-Term Basis AGENCY: Office of Fossil Energy, DOE. ACTION: Notice of application....

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

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

  4. Intercalated graphite fiber composites as EMI shields in aerospace structures

    Science.gov (United States)

    Gaier, James R.

    1992-01-01

    The requirements for electromagnetic interference (EMI) shielding in aerospace structures are more complicated than those for ground structures because of their weight limitations. As a result, the best EMI shielding materials must combine low density, high strength, and high elastic modulus with high shielding ability. EMI shielding characteristics were calculated for shields formed from pristine and intercalated graphite fiber/epoxy composites and compare to preliminary experimental results for these materials and to the characteristics of shields made from aluminum. Calculations indicate that effective EMI shields could be fabricated from intercalated graphite composites which would have less than 12 percent of the mass of conventional aluminum shields, based on mechanical properties and shielding characteristics alone.

  5. SAS1 and SAS4, two new shielding analysis sequences for spent fuel casks

    International Nuclear Information System (INIS)

    Two important Shielding Analysis Sequences (SAS) have recently been developed within the SCALE computational system. These sequences significantly enhance the existing SCALE system capabilities for evaluating radiation doses exterior to spent fuel casks. These new control module sequences (SAS1 and SAS4) and their capabilities are discussed and demonstrated. Particular attention is given to the new SAS4 sequence which provides an automated scheme for generating and using biasing parameters in a subsequent Monte Carlo analysis of a cask

  6. Accelerator shielding experts meet at CERN

    CERN Multimedia

    CERN Bulletin

    2010-01-01

    Fifteen years after its first CERN edition, the Shielding Aspects of Accelerator, Targets and Irradiation Facility (SATIF) conference was held again here from 2-4 June. Now at its 10th edition, SATIF10 brought together experts from all over the world to discuss issues related to the shielding techniques. They set out the scene for an improved collaboration and discussed novel shielding solutions.   This was the most attended meeting of the series with more than 65 participants from 34 institutions and 14 countries. “We welcomed experts from many different laboratories around the world. We come from different contexts but we face similar problems. In this year’s session, among other things, we discussed ways for improving the effectiveness of calculations versus real data, as well as experimental solutions to investigate the damage that radiation produces on various materials and the electronics”, says Marco Silari, Chair of the conference and member of the DGS/RP gro...

  7. Calculation system analysis for radiation shielding

    International Nuclear Information System (INIS)

    This work consists of the computational system implementation for nuclear reactor shielding analysis. The system has as objectives to facilitate the installation of the calculation framework, problem set-up, and results analysis. Several computational programmes commonly used for cross-section preparation and radiation transport were chosen for the system. This work represents the capacity necessary for nuclear reactor and particle accelerator shielding design, to aid in nuclear experiments and in the utilization of nuclear techniques that require the radiation field calculation. The system was implemented in PC-DOS environment and consists of the necessary and sufficient programs and data for generation of the cross sections, groups constants, self-shielding factors, activation sources, for the calculation of neutron and gamma-ray fluence, dose rates, and other types of response functions. (author). 11 refs., 8 figs

  8. Calculated shielding factors for selected European houses

    International Nuclear Information System (INIS)

    Shielding factors for gamma radiation from activity deposited on structures and ground surfaces have been calculated with the computer model DEPSHIELD for single-family and multi-storey buildings in France, United Kingdom and Denmark. For all three countries it was found that the shielding factors for single-family houses are approximately a factor of 2 - 10 higher that those for buildings with five or more storeys. Away from doors and windows the shielding factors for French, British, and Danish single-family houses are in the range 0.03 - 0.1, 0.06 - 0.4, and 0.07 - 0.3, respectively. The uncertainties of the calculations are discussed and DEPSHIELD-results are compared with other methods as well as with experimental results. (author)

  9. Radiation safety shield for a syringe

    International Nuclear Information System (INIS)

    Safety apparatus for use in administering radioactive serums by a syringe, without endangering the health and safety of the medical operators is described. The apparatus consists of a sheath and a shield which can be retracted into the sheath to assay the radioactive serum in an assay well. The shield can be moved from the retracted position into an extended position when the serum is to be injected into the patient. To protect the operator, the shield can be constructed of tantalum or any like high density substance to attenuate the radiation, emanating from the radioactive serums contained in the syringe, from passing to the atmosphere. A lead glass window is provided so that the operator can determine the exact quantity of the radioactive serum which is contained in the syringe

  10. Radiation shielding effectiveness of newly developed superconductors

    Science.gov (United States)

    Singh, Vishwanath P.; Medhat, M. E.; Badiger, N. M.; Saliqur Rahman, Abu Zayed Mohammad

    2015-01-01

    Gamma ray shielding effectiveness of superconductors with a high mass density has been investigated. We calculated the mass attenuation coefficients, the mean free path (mfp) and the exposure buildup factor (EBF). The gamma ray EBF was computed using the Geometric Progression (G-P) fitting method at energies 0.015-15 MeV, and for penetration depths up to 40 mfp. The fast-neutron shielding effectiveness has been characterized by the effective neutron removal cross-section of the superconductors. It is shown that CaPtSi3, CaIrSi3, and Bi2Sr2Ca1Cu2O8.2 are superior shielding materials for gamma rays and Tl0.6Rb0.4Fe1.67Se2 for fast neutrons. The present work should be useful in various applications of superconductors in fusion engineering and design.

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

    Directory of Open Access Journals (Sweden)

    P. A. Graniero

    1999-01-01

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

  12. On the use of double-walled tubes as a means to improve safety and availability of the EU DEMO Water-Cooled Pb-17Li blanket

    International Nuclear Information System (INIS)

    The impact on the blanket reliability and availability of both double-walled tube and welded joint failures in the Water-Cooled Pb-17Li Demo Single-Box blanket reference design is examined. The pertinence of employing a leak detection system is analysed and its contribution to the blanket safety and availability is evaluated. The contribution of welded joints to the blanket safety and availability is evaluated in normal operation. (orig.)

  13. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  14. Experimental analysis of an MIM capacitor with a concave shield

    Institute of Scientific and Technical Information of China (English)

    Liu Lintao; Yu Mingyan; Wang Jinxiang

    2009-01-01

    A novel shielding scheme is developed by inserting a concave shield between a metal-insulator-metal (MIM) capacitor and the silicon substrate. Chip measurements reveal that the concave shield improves the quality factor by 11 % at 11.8 GHz and 14% at 18.8 GHz compared with an unshielded MIM capacitor. It also alleviates the effect on shunt capacitance between the bottom plate of the MIM capacitor and the shield layer. Moreover, because the concave shields simplify substrate modeling, a simple circuit model of the MIM capacitor with concave shield is presented for radio frequency applications.

  15. Compilation of the computing methods in radiation shielding

    International Nuclear Information System (INIS)

    In order to update the KAERI radiation shielding technology, the calculational shielding methods were surveyed throughly. Computer codes and data libraries for radiation shielding calculation were collected and some model calculations were carried out with them. So far the following materials were ensured for our future use: 23 shielding codes, 7 data libraries, 7 data processing codes and 11 peripheral shielding codes. All of these were compiled again for the CYBER-73 computer system, and will be widely used in shielding analysis of accelerators, shipping casks as well as nuclear power plant. (author)

  16. Novel Concepts for Radiation Shielding Materials

    Science.gov (United States)

    Oliva-Buisson, Yvette J.

    2014-01-01

    It is critical that safety factors be maximized with respect to long duration, extraterrestrial space flight. Any significant improvement in radiation protection will be critical in ensuring the safety of crew and hardware on such missions. The project goal is to study novel concepts for radiation shielding materials that can be used for long-duration space missions. As part of this project we will investigate the use of thin films for the evaluation of a containment system that can retain liquid hydrogen and provide the necessary hydrogen density for effective shielding.

  17. [Electromagnetic Shielding Alters Behaviour of Rats].

    Science.gov (United States)

    Temuryants, N A; Kostyuk, A S; Tumanyants, K N

    2015-01-01

    It has been found that long-term electromagnetic shielding (19 hours per day for 10 days) leads to an increase in the duration of passive swimming time in male rats, decrease the duration of active swimming in the "forced swim" test as well as decrease of libido. On the other hand animals kept under the "open field" conditions do not show significant deviations from their normal behavior. Therefore, one could conclude that moderate electromagnetic shielding causes a depression-like state in rats. PMID:26080600

  18. Shielding effectiveness of rectangular cavity made of a new shielding material and resonance suppression

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    New shielding material has become an alternative to traditional metal to shield boxes from electromagnetic interferences. This article introduces the theory of transmission line method to study the shield boxes made of a new sort of material, and then expands the fundamental formulas to deal with the cases of multiple holes and polarization with arbitrary angle. By means of genetic algorithms with the aid of a three dimensional simulation tool, the damping of electromagnetic resonances in enclosures is researched.The computation indicates that under resonant frequency, electromagnetic resonance results in low, even negative shielding coefficient; whereas, for the same areas, shielding effectiveness of a single hole is worse than that of multiple holes. Shielding coefficient varies when polarization angle increases, and the coupled field through the rectangular aperture with the long side parallel to the thin wire is much weaker than that with the long side vertical to the thin wire. By using the metallic-loss dielectric layer of optimized calculation on the internal surface of the cavity, the best result of resonance suppression has been realized with the same thickness of coating. Finally, according to the calculation result, suggestions for shielding are proposed.

  19. 75 FR 33803 - Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-06-15

    ... Energy Regulatory Commission Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization June 8, 2010. Take notice that on June 1, 2010, Sabine Pipe Line LLC (Sabine), 4800 Fournace Place, Bellaire... L. Kirk, Regulatory Specialist, Chevron Pipe Line Company, 4800 Fournace Place, Bellaire,...

  20. 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)

  1. 77 FR 38622 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-06-28

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on June 4, 2012, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  2. 78 FR 68835 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-11-15

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on October 31, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  3. 78 FR 25264 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-04-30

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on April 16, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  4. 78 FR 53746 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-08-30

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on August 13, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  5. 77 FR 14517 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-03-12

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 21, 2012 Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 State Highway 56,...

  6. 78 FR 13663 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-02-28

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 11, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, P.O. Box...

  7. 75 FR 8053 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-02-23

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization February 16, 2010. Take notice that on January 29, 2010, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700...

  8. 32 CFR Appendix D to Part 505 - Exemptions; Exceptions; and DoD Blanket Routine Uses

    Science.gov (United States)

    2010-07-01

    ... 32 National Defense 3 2010-07-01 2010-07-01 true Exemptions; Exceptions; and DoD Blanket Routine Uses D Appendix D to Part 505 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS ARMY PRIVACY ACT PROGRAM Pt. 505, App. D Appendix D...

  9. 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.)

  10. Stochastic modeling to determine the economic effects of blanket, selective, and no dry cow therapy

    NARCIS (Netherlands)

    Huijps, K.; Hogeveen, H.

    2007-01-01

    In many countries, blanket dry cow therapy (DCT) is the standard way to dry off cows. Because of concerns about antibiotic resistance, selective DCT is proposed as an alternative. The economic consequences of different types of DCT were studied previously, but variation between input traits and diff

  11. 75 FR 38092 - The Dow Chemical Company; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-07-01

    ... The Dow Chemical Company (Dow), requesting blanket authorization to export liquefied natural gas (LNG... with its principal place of business in Midland, Michigan. Dow is an international chemical and...\\ The Dow Chemical Company, DOE/FE Order No. 2754 issued February 25, 2010. Current Application In...

  12. 77 FR 53874 - The Dow Chemical Company; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-09-04

    ... application (Application), filed on July 13, 2012, by The Dow Chemical Company (Dow), requesting blanket... comments are invited. \\1\\ The Dow Chemical Company, DOE/FE Order No. 2859 (October 5, 2010) extends through... chemical and plastics manufacturing company with operations in a number of U.S. states. Dow owns...

  13. 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.)

  14. 77 FR 31004 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-05-24

    ... Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on May 9, 2012, Southern Natural Gas Company (Southern), 569 Brookwood Village, Suite....210 of the Federal Energy Regulatory Commission's regulations under the Natural Gas Act (NGA),...

  15. 76 FR 18216 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2011-04-01

    ... Federal Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on March 16, 2011, Southern Natural Gas Company (Southern), Post Office Box 2563... and 157.216 of the Commission's Regulations under the Natural Gas Act (NGA) as amended, to abandon...

  16. 75 FR 13535 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-03-22

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization March 16, 2010. Take notice that on March 12, 2010, Northern Natural Gas Company (Northern), 1111 South... External Affairs, Northern Natural Gas Company, 1111 South 103rd Street, Omaha, Nebraska 68124, at...

  17. 75 FR 3232 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-01-20

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization January 8, 2010. Take notice that on December 30, 2009, Northern Natural Gas Company (Northern), 1111... sections 157.205 and 157.214 of the Commission's regulations under the Natural Gas Act for authorization...

  18. Two-dimensional cross-section sensitivity and uncertainty analysis for fusion reactor blankets

    International Nuclear Information System (INIS)

    A two-dimensional sensitivity and uncertainty analysis for the heating of the TF coil for the FED (fusion engineering device) blanket was performed. The uncertainties calculated are of the same order of magnitude as those resulting from a one-dimensional analysis. The largest uncertainties were caused by the cross section uncertainties for chromium

  19. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Science.gov (United States)

    2010-07-01

    ... Blanket Routine Uses (a) Routine Use—Law Enforcement. If a system of records maintained by a DoD Component, to carry out its functions, indicates a violation or potential violation of law, whether civil... decision concerning the hiring or retention of an employee, the issuance of a security clearance,...

  20. 78 FR 53737 - ConocoPhillips Company; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2013-08-30

    ...Phillips Company; Application for Blanket Authorization To Export Previously Imported Liquefied Natural Gas... application (Application), filed on August 7, 2013, by ConocoPhillips Company (ConocoPhillips), requesting... which trade is not prohibited by U.S. law or policy. The Application was filed under section 3 of...