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Sample records for blanket shield module

  1. Influence of the blanket shield modules geometry on the operation of the ITER ICRF antenna

    Energy Technology Data Exchange (ETDEWEB)

    Louche, F., E-mail: fabrice.louche@rma.ac.be [Laboratory for Plasma Physics, Royal Military Academy, 30, Avenue de la Renaissance, 1000 Brussels (Belgium); Dumortier, P.; Durodié, F.; Messiaen, A. [Laboratory for Plasma Physics, Royal Military Academy, 30, Avenue de la Renaissance, 1000 Brussels (Belgium)

    2013-10-15

    Highlights: ► The ITER ICRF antenna and its surrounding blanket shield modules have been modeled with the 3D electromagnetic software Microwave Studio. ► Unexpected resonances are detected in the ITER relevant range of frequencies. ► These resonances are caused by the geometry of the modules, and in particular by the cavity at the back, between the module rear face and the port plug outer face, present in the considered model. ► Simplified modeling and transmission line computations validate this interpretation. ► The resonance is strongly dependent on the detailed geometry of the modules, but large voids should be avoided. -- Abstract: Three-dimensional electromagnetic simulations of the ITER ICRF antenna have been recently performed with the commercial code CST Microwave Studio{sup ®} (MWS) [1]. A detailed model imported from the CATIA{sup ®} file has been considered: it includes the 24 straps array (CY3.1 geometry [2]) and the surrounding blanket shield modules. The transient solver in MWS has detected the presence of a very localized peak in the input impedance matrix at a frequency of approximately 51 MHz in vacuum conditions. The presence of such a resonance in the ITER operating range of frequency is of concern and should be understood as previous analysis reported in [3] concluded that TEM and non-TEM modes are not expected in this frequency band as long as the antenna is grounded to the port at 1 m back from the antenna front face. By using a simplified model of the geometry we demonstrate that the resonance is a consequence of the considered geometry of the blanket shield modules and in particular of the cavity at the back of the modules made of the module attachment and the port plug outer face. We show that the presence of such a cavity locally increases the coaxial line impedance and allows for a TEM mode in the band. This physical analysis is supported by a transmission line model where the system made of the antenna and its surrounding

  2. Fusion reactor blanket/shield design study

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Clemmer, R.G.; Harkness, S.D.

    1979-07-01

    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.

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

    NARCIS (Netherlands)

    Gessner, R.; Aiello, G.; Grossetti, G.; Meier, A.; Ronden, D.; Spaeh, P.; Scherer, T.; Schreck, S.; Strauss, D.; Vaccaro, A.

    2013-01-01

    The final design of the structural system for the ITER EC H&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 Mo

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi; Furuya, Kazuyuki; Hatano, Toshihisa; Kuroda, Toshimasa; Enoeda, Mikio; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Takatsu, Hideyuki [Japan Atomic Energy Research Inst., Office of ITER Project Promotion, Tokyo (Japan)

    2000-07-01

    A prototype ({approx}900{sup H} x 1700{sup W} x 350{sup T} mm) of the ITER shielding blanket module has been fabricated following the previous successful fabrication of a small-scale ({approx}500{sup H} x 400{sup W} x 150{sup T} mm) and mid-scale ({approx}800{sup H} x 500{sup W} x 350{sup T} 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 Al{sub 2}O{sub 3} 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

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

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

  8. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  9. Uranium self-shielding in fast reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kadiroglu, O.K.; Driscoll, M.J.

    1976-03-01

    The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.

  10. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  11. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

    2015-10-15

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. 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 BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  12. Neutronics optimization study for D-D fusion reactor blanket/shield

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, T.; Kanda, Y.; Nakashima, H.

    1985-12-01

    Position-dependent optimization calculations have been carried out on a D-D fusion reactor blanket/shield to maximize the energy gain in the blanket and to minimize the atomic displacement rate of the copper stabilizer in the superconducting magnet. The results obtained by using the optimization code SWAN indicate the advantage of D/sub 2/O coolant over H/sub 2/O coolant with respect to increasing the energy gain, and the difference in the optimal shield distributions between D-T and D-D neutron sources. The possibility of improving both the energy gain and radiation shielding characteristics is also discussed.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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.

  18. ITER (International Thermonuclear Experimental Reactor) shield and blanket work package report

    Energy Technology Data Exchange (ETDEWEB)

    1988-06-01

    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.

  19. Current status on the detailed design and development of fabrication techniques for the ITER blanket shield block in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Duck-Hoi [National Fusion Research Institute, 52 Yeoeun-dong, Yuseong-gu, Daejeon 305-333 (Korea, Republic of)], E-mail: kdwh@nfri.re.kr; Cho, Seungyon; Ahn, Mu-Young; Lee, Eun-Seok; Jung, Ki Jung [National Fusion Research Institute, 52 Yeoeun-dong, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Kim, Do-Hyeong [ANST, Inc., 222-7 Guro3-dong, Guro-gu, Seoul 152-848 (Korea, Republic of)

    2008-12-15

    Recent activities and progress on the design and fabrication of the ITER blanket shield block in Korea are described in this paper. Hydraulic analyses, using a flow driver model for determining the gap between the radial cooling passages and flow drivers inside the shield block, were performed. The thermo-hydraulic analysis of half of a shield block was also conducted to investigate the uniformity of the flow stream in cooling passages and to evaluate the temperature distribution in the structure. The maximum temperature is below the allowable value, although hot spots occurred in the corner edge in the shield block. A manufacturing feasibility study for the development of the blanket shield block was performed in cooperation with KO industries. It was found that specific techniques would be required for the successful fabrication of an ITER blanket shield block, specifically electron-beam welding at a thickness up to 110 mm. The development of joining and drilling technologies for the thick shield block and lid joints is in progress. In addition, a full scale mock-up fabrication and the development of NDT techniques are planned in the near future.

  20. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

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

  2. 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.

  3. 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.

  4. Nuclear analysis of ITER Test Blanket Module Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Villari, Rosaria, E-mail: rosaria.villari@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Kim, Byoung Yoon; Barabash, Vladimir; Giancarli, Luciano; Levesy, Bruno; Loughlin, Michael; Merola, Mario [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Moro, Fabio [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Pascal, Romain [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium); Polunovsky, Eduard; Van Der Laan, Jaap G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France)

    2015-10-15

    Highlights: • 3D nuclear analysis of the ITER TBM Port Plug (PP). • Calculations of neutron fluxes, nuclear heating, damage and He-production in TBM PP components. • Shutdown dose rate assessment with Advanced D1S method considering different configurations. • Potential design improvements to reduce the shutdown dose rate in the port interspace. - Abstract: Nuclear analyses have been performed for the ITER Test Blanket Module Port Plug (TBM PP) using the MCNP-5 Monte Carlo Code. A detailed 3D model of the TBM Port Plug with dummy TBM has been integrated into the ITER MCNP model (B-lite v.3). Neutron fluxes, nuclear heating, helium production and neutron damage have been calculated in all the TBM PP components. Global shutdown dose rate calculations have also been performed with Advanced D1S method for different configurations of the TBM PP system. This paper presents the results of these analyses and discusses potential design improvements aiming to further reduce the shutdown dose rate in the port interspace.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  10. 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.

  11. 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.

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

    NARCIS (Netherlands)

    Schaffer, M. J.; Snipes, J. A.; Gohil, P.; P. de Vries,; 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.

    2011-01-01

    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 m

  13. 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.

  14. Effect of thick blanket modules on neoclassical tearing mode locking in ITER

    Science.gov (United States)

    La Haye, R. J.; Paz-Soldan, C.; Liu, Y. Q.

    2017-01-01

    The rotation of m/n  =  2/1 tearing modes can be slowed and stopped (i.e. locked) by eddy currents induced in resistive walls in conjunction with residual error fields that provide a final ‘notch’ point. This is a particular issue in ITER with large inertia and low applied torque (m and n are poloidal and toroidal mode numbers respectively). Previous estimates of tolerable 2/1 island widths in ITER found that the ITER electron cyclotron current drive (ECCD) system could catch and subdue such islands before they persisted long enough and grew large enough to lock. These estimates were based on a forecast of initial island rotation using the n  =  1 resistive penetration time of the inner vacuum vessel wall and benchmarked to DIII-D high-rotation plasmas, However, rotating tearing modes in ITER will also induce eddy currents in the blanket as the effective first wall that can shield the inner vessel. The closer fitting blanket wall has a much shorter time constant and should allow several times smaller islands to lock several times faster in ITER than previously considered; this challenges the ECCD stabilization. Recent DIII-D ITER baseline scenario (IBS) plasmas with low rotation through small applied torque allow better modeling and scaling to ITER with the blanket as the first resistive wall.

  15. Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, H. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)], E-mail: tanigawa.hiroyasu@jaea.go.jp; Hirose, T.; Shiba, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kasada, R. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Wakai, E. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Serizawa, H.; Kawahito, Y. [Joining and Welding Research Institute, Osaka University, Ibaraki, Osaka 567-0047 (Japan); Jitsukawa, S. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kimura, A. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Kohno, Y. [Department of Materials Science and Engineering, Muroran Institute of Technology, Muroran, Hokkaido 050-8585 (Japan); Kohyama, A. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Katayama, S. [Joining and Welding Research Institute, Osaka University, Ibaraki, Osaka 567-0047 (Japan); Mori, H.; Nishimoto, K. [Division of Materials and Manufacturing Science, Osaka University, Ibaraki, Osaka 565-0871 (Japan); Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6132 (United States)

    2008-12-15

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed.

  16. Safety analysis of the US dual coolant liquid lead lithium ITER test blanket module

    Science.gov (United States)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2007-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  17. Irradiation behavior of Ti 4Al 2V (ΠT-3B) alloy for ITER blanket modules flexible attachment

    Science.gov (United States)

    Rodchenkov, B. S.; Kozlov, A. V.; Kuznetsov, Yu. G.; Kalinin, G. M.; Strebkov, Yu. S.

    2007-08-01

    Titanium alloys are recommended as a material to manufacture flexible attachments of the shield blanket modules in the ITER reactor owing to their advantageous combination of properties, i.e., high resistance to impact loading, strength, density and low thermal expansion coefficient. An additional factor for selecting Ti alloys is their fast induced radioactivity decay. The (α + β)-Ti alloys have higher strength than (α)-Ti alloys but are less developed. The data base on the irradiation behavior of these materials is limited. Neutron irradiation of (α)-Ti-4Al-2V (ΠT-3B) alloy has been performed in the framework of the ITER R&D programme. Specimens from a forging of Ti-4Al-2V alloy were irradiated in the IVV-2M reactor to doses of (0.32-0.43) dpa at temperatures of (240-260) °C. This paper describes the results of tensile, low cycle fatigue and fracture toughness tests of alloy in the unirradiated and neutron irradiated conditions. The results obtained are compared with those of the (α + β)-Ti-6Al-4V alloy.

  18. 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.

  19. 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.

  20. 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.

  1. Thermo-Mechanical Analyses of the High Heat Flux Component for ITER Dual Functional Lithium Lead Test Blanket Module

    Institute of Scientific and Technical Information of China (English)

    CHEN Hongli; BAI Yunqing

    2009-01-01

    The finite element code ANSYS is used to calculate the temperature and stress distributions for the first wall of DFLL-TBM (dual functional lithium lead-test blanket module),for testing in ITER. Preliminary analyses indicate that not only the low temperature design rules,the well-known 3Sm rules, are satisfied for the first wall, but the additional high temperature structural design criteria for the creep damage limits and creep-ratcheting limits are met as well.

  2. Experimental results and validation of a method to reconstruct forces on the ITER test blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Zeile, Christian, E-mail: christian.zeile@kit.edu; Maione, Ivan A.

    2015-10-15

    Highlights: • An in operation force measurement system for the ITER EU HCPB TBM has been developed. • The force reconstruction methods are based on strain measurements on the attachment system. • An experimental setup and a corresponding mock-up have been built. • A set of test cases representing ITER relevant excitations has been used for validation. • The influence of modeling errors on the force reconstruction has been investigated. - Abstract: In order to reconstruct forces on the test blanket modules in ITER, two force reconstruction methods, the augmented Kalman filter and a model predictive controller, have been selected and developed to estimate the forces based on strain measurements on the attachment system. A dedicated experimental setup with a corresponding mock-up has been designed and built to validate these methods. A set of test cases has been defined to represent possible excitation of the system. It has been shown that the errors in the estimated forces mainly depend on the accuracy of the identified model used by the algorithms. Furthermore, it has been found that a minimum of 10 strain gauges is necessary to allow for a low error in the reconstructed forces.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, G J; Ellis, R; Gorelenkova, M; Heidbrink, W W; Kurki-Suonio, T; Nazikian, R; Salmi, A; Schaffer, M J; Shinohara, K; Snipes, J A; Spong, D A; Koskela, T

    2011-06-03

    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.

  4. Fast-ion effects during test blanket module simulation experiments in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, G. [Princeton Plasma Physics Laboratory (PPPL); Budny, R. V. [Princeton Plasma Physics Laboratory (PPPL); Ellis, R. [Princeton Plasma Physics Laboratory (PPPL); Gorelenkova, M. [Princeton Plasma Physics Laboratory (PPPL); Heidbrink, W. [University of California, Irvine; Kurki-Suonio, T. [Aalto University, Finland; Nazikian, Raffi [Princeton Plasma Physics Laboratory (PPPL); Saimi, A. [Aalto University, Finland; Schaffer, M. J. [General Atomics, San Diego; Shinohara, K. [Japan Atomic Energy Agency (JAEA), Naka; Snipes, J. A. [ITER Organization, Cadarache, France; Spong, Donald A [ORNL; Koskela, T. [Aalto University, Finland; Van Zeeland, Michael [General Atomics

    2011-01-01

    Fast beam-ion losses were studied in DIII-D in the presence of a scaled mock-up 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.

  5. Shield fabrication development of ITER primary wall modules by powder HIP. ITER task T216-Subtask 3E1

    Energy Technology Data Exchange (ETDEWEB)

    Lind, A

    1997-12-01

    A research and development program for the blanket shield in the International Thermonuclear Experimental Reactor (ITER) has been implemented to provide input for the design and manufacture of full scale production components. It comprises fabrication and testing of mock-ups and prototype modules. The design, materials, manufacture, examination, testing and inspection of the mock-ups representing future full scale production modules. This work applies to the development of a shield block fabrication method by Hot Isostatic Pressing (HIP) starting from a gas atomised powder and pre-fabricated cooling tube galleries. The size of the block is 1250 x 650 x 250 mm and the weight is about 1400 kg. Examination and testing of the block was performed to determine properties, achieved fabrication tolerances, and quality of bonding. It is concluded that the today`s powder HIP route gives a 316 LN IG material with mechanical properties which fulfills the ITER material specification requirements and a fully dense block which is easy to examine with ultrasonic methods. The joints between tubes and matrix are excellent. In order to achieve and maintain accuracy in positioning of the tubes during fabrication improvements of the standard fabrication route have been identified, such as the positioning of tubes during welding, the powder particle distribution and the powder filling procedure. Modification of the actual HIP cycle may also be required

  6. A path to stable low-torque plasma operation in ITER with test blanket modules

    Science.gov (United States)

    Lanctot, M. J.; Snipes, J. A.; Reimerdes, H.; Paz-Soldan, C.; Logan, N.; Hanson, J. M.; Buttery, R. J.; deGrassie, J. S.; Garofalo, A. M.; Gray, T. K.; Grierson, B. A.; King, J. D.; Kramer, G. J.; La Haye, R. J.; Pace, D. C.; Park, J.-K.; Salmi, A.; Shiraki, D.; Strait, E. J.; Solomon, W. M.; Tala, T.; Van Zeeland, M. A.

    2017-03-01

    New experiments in the low-torque ITER Q  =  10 scenario on DIII-D demonstrate that n  =  1 magnetic fields from a single row of ex-vessel control coils enable operation at ITER performance metrics in the presence of applied non-axisymmetric magnetic fields from a test blanket module (TBM) mock-up coil. With n  =  1 compensation, operation below the ITER-equivalent injected torque is successful at three times the ITER equivalent toroidal magnetic field ripple for a pair of TBMs in one equatorial port, whereas the uncompensated TBM field leads to rotation collapse, loss of H-mode and plasma current disruption. In companion experiments at high plasma beta, where the n  =  1 plasma response is enhanced, uncorrected TBM fields degrade energy confinement and the plasma angular momentum while increasing fast ion losses; however, disruptions are not routinely encountered owing to increased levels of injected neutral beam torque. In this regime, n  =  1 field compensation leads to recovery of a dominant fraction of the TBM-induced plasma pressure and rotation degradation, and an 80% reduction in the heat load to the first wall. These results show that the n  =  1 plasma response plays a dominant role in determining plasma stability, and that n  =  1 field compensation alone not only recovers most of the impact on plasma performance of the TBM, but also protects the first wall from potentially damaging heat flux. Despite these benefits, plasma rotation braking from the TBM fields cannot be fully recovered using standard error field control. Given the uncertainty in extrapolation of these results to the ITER configuration, it is prudent to design the TBMs with as low a ferromagnetic mass as possible without jeopardizing the TBM mission.

  7. 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.

  8. ITER Test Blanket Module Error Field Simulation Experiments at DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Schaffer, M. J. [General Atomics, San Diego; Testa, D. [CRPP, Switzerland; Snipes, J. A. [ITER Organization, Cadarache, France; Gohil, P. [General Atomics; De Vries, P. [Culham Centre for Fusion Energy, Culham, UK; Evans, T. E. [General Atomics, San Diego; Fenstermacher, M. E. [Lawrence Livermore National Laboratory (LLNL); Gao, X. [Academia Sinica, Institute of Plasma Physics, Hefei, China; Garofalo, A. [General Atomics, San Diego; Gates, D.A. [Princeton Plasma Physics Laboratory (PPPL); Greenfield, C. M. [General Atomics; Heidbrink, W. [University of California, Irvine; La Haye, R. [General Atomics, San Diego; Liu, S. [ASIPP, Hefei, China; Loarte, A. [ITER Organization, Cadarache, France; Nave, M. F. F. [Association EURATOM/IST, Lisbon, Portugal; Osborne, T.H. [General Atomics, San Diego; Oyama, N. [Japan Atomic Energy Agency (JAEA); Osakabe, M. [National Institute for Fusion Science, Toki, Japan; Park, J. K. [Princeton Plasma Physics Laboratory (PPPL); Ramasubramanian, N. [Institute for Plasma Research, Gandhinagar, India; Reimerdes, H. [Columbia University; Saibene, G. [Fusion for Energy (F4E), Barcelona, Spain; Saimi, A. [Aalto University, Finland; Shinohara, K. [Japan Atomic Energy Agency (JAEA), Naka; Spong, Donald A [ORNL; Solomon, W. M. [Princeton Plasma Physics Laboratory (PPPL); Tala, T. [Association Euratom-Tekes, Finland; Zhu, Y. B. [University of California, Irvine; Zhai, K. [University of Wisconsin, Madison; Boedo, J. [University of California, San Diego; Chuyanov, V. [ITER Organization, Cadarache, France; Doyle, E. J. [University of California, Los Angeles; Jakubowski, M. W. [Max-Planck-Institute for Plasmaphysik, EURATOM-Association, Greifswald, Germany; Jhang, H. [National Fusion Research Institute, Daejon, South Korea; Nazikian, Raffi [Princeton Plasma Physics Laboratory (PPPL); Pustovitov, V. D. [Russian Research Center, Kurchatov Institute, Moscow, Russia; Schmitz, O. [Forschungszentrum Julich, Julich, Germany; Sanchez, Raul [ORNL; Srinivasan, R. [Institute for Plasma Research, Gandhinagar, India; Taylor, T. S. [General Atomics, San Diego; Wade, M. [General Atomics, San Diego; You, K. I. [National Fusion Research Institute, Daejon, South Korea; Zeng, L. [University of California, Los Angeles

    2011-01-01

    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 Delta upsilon/upsilon similar to 60% via non-resonant braking. Changes to global Delta n/n, Delta beta/beta and Delta H(98)/H(98) were similar to 3 times smaller. These effects are stronger at higher beta. 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.

  9. Technical issues of RAFMs for the fabrication of ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki [Japan Atomic Energy Agency (JP)] (and others)

    2007-07-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems, as it has they have been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. F82H and JLF-1 are RAFMs, which have been developed and studied in Japan and the various effects of irradiation were reported. F82H is designed with emphasis on high temperature property and weldability, and was provided and evaluated in various countries as a part of the IEA fusion materials development collaboration. The JAEA/US collaboration program also has been conducted with the emphasis on irradiation effects of F82H. Now, among the existing database for RAFMs the most extensive one is that for F82H. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of ITER Test Blanket Module (TBM) suggested from the recent achievements in Japan. It is desirable to make the status of RAFMs equivalent to commercial steels to use RAFMs as the ITER-TBM structural material. This would require demonstrating the reproducibility and weldability as well as providing the database. The excellent reproducibility of F82H has been demonstrated with four 5-ton-heats, and two of them were provided as F82H-IEA heats. It has been also proved that F82H could be provided as plates (thickness of 1.5 to 55 mm), pipes and rectangular tubes. It is also important to have the excellent weldability as the TBM has about 300m length of weld line, and it was proved through TIG, EB and YAG weld test performed in air atmosphere. Various mechanical and microstructural data have been accumulated including long-term tests such as creep rupture tests and aging tests. Although F82H is a well-perceived RAFM as the ITER-TBM structural material, some issues are

  10. Analysis of the thermo-mechanical behaviour of the DEMO Water-Cooled Lithium Lead breeding blanket module under normal operation steady state conditions

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A.; Arena, P. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Aubert, J. [CEA Saclay, DEN/DANS/DM2S/SEMT, 91191 Gif sur Yvette Cedex (France); Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, R. [ENEA – C.R. Brasimone, 40032 Camugnano (Italy); Li Puma, A. [CEA Saclay, DEN/DANS/DM2S/SEMT, 91191 Gif sur Yvette Cedex (France); Tincani, A. [ENEA – C.R. Brasimone, 40032 Camugnano (Italy)

    2015-10-15

    Highlights: • A DEMO WCLL blanket module thermo-mechanical behaviour has been investigated. • Two models of the WCLL blanket module have been set-up adopting a code based on FEM. • The water flow domain in the module has been considered. • A set of uncoupled steady state thermo-mechanical analyses has been carried out. • Critical temperature is not overcome. Safety verifications are generally satisfied. - Abstract: Within the framework of DEMO R&D activities, a research cooperation has been launched between ENEA, the University of Palermo and CEA to investigate the thermo-mechanical behaviour of the outboard equatorial module of the DEMO1 Water-Cooled Lithium Lead (WCLL) blanket under normal operation steady state scenario. The research campaign has been carried out following a theoretical–computational approach based on the Finite Element Method (FEM) and adopting a qualified commercial FEM code. In particular, two different 3D FEM models (Model 1 and Model 2), reproducing respectively the central and the lateral poloidal–radial slices of the WCLL blanket module, have been set up. A particular attention has been paid to the modelling of water flow domain, within both the segment box channels and the breeder zone tubes, to simulate realistically the coolant-box thermal coupling. Results obtained are herewith reported and critically discussed.

  11. 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.

  12. 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.

  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. Thermal-mechanical and thermal-hydraulic integrated study of the Helium-Cooled Lithium Lead Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Chiovaro, P., E-mail: pg.chiovaro@din.unipa.it [Dipartimento di Ingegneria Nucleare, Universita di Palermo, Palermo (Italy); Di Maio, P.A.; Giammusso, R.; Lupo, Q.; Vella, G. [Dipartimento di Ingegneria Nucleare, Universita di Palermo, Palermo (Italy)

    2010-12-15

    The Helium-Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two TBM to be installed in an ITER equatorial port since day 1 of operation, with the specific aim to investigate the main concept functionalities and issues such as high efficiency helium cooling, resistance to thermo-mechanical stresses, manufacturing techniques, as well as tritium transport, magneto-hydrodynamics effects and corrosion. In particular, in order to show a DEMO-relevant thermo-mechanical and thermal-hydraulic behavior, the HCLL-TBM has to meet several requirements especially as far as its coolant thermofluid-dynamic conditions and its thermal-mechanical field are concerned. The present paper is focused on the assessment of the HCLL-TBM thermal-mechanical performances under both nominal and accidental load conditions, by adopting a computational approach based on the Finite Element Method. A realistic 3D finite element model of the whole HCLL-TBM, in the horizontal first wall design has been set up, consisting of about 597,000 elements and 767,000 nodes. In particular, since the thermal fields of both the module and the coolant are strictly coupled, the helium flow domain has been modeled too and a thermal contact model has been set up to properly simulate the convective heat transfer between the structure wall and the coolant. Pure conductive heat transfer has been assumed within the Pb-Li eutectic alloy of the breeder units. The volumetric density of the nuclear deposited power, recently calculated at Department of Nuclear Engineering of the University of Palermo by the MCNP 4C code, has been applied as distributed thermal load in order to assess the potential influence on the module thermo-mechanical performances of the markedly non-uniform poloidal and toroidal distributions that have been predicted within the Segment Box. Different loading scenarios have been considered as to the heat flux onto the module First Wall. Steady state and transient thermal-mechanical analyses

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  16. 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.

  17. Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

    2016-02-15

    Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-09-01

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

  19. Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

  20. Design of the helium cooled lithium lead breeding blanket in CEA: from TBM to DEMO

    Science.gov (United States)

    Aiello, G.; Aubert, J.; Forest, L.; Jaboulay, J.-C.; Li Puma, A.; Boccaccini, L. V.

    2017-04-01

    The helium cooled lithium lead (HCLL) blanket concept was originally developed in CEA at the beginning of 2000: it is one of the two European blanket concepts to be tested in ITER in the form of a test blanket module (TBM) and one of the four blanket concepts currently being considered for the DEMOnstration reactor that will follow ITER. The TBM is a highly optimized component for the ITER environment that will provide crucial information for the development of the DEMO blanket, but its design needs to be adapted to the DEMO reactor. With respect to the TBM design, reduction of the steel content in the breeding zone (BZ) is sought in order to maximize tritium breeding reactions. Different options are being studied, with the potential of reaching tritium breeding ratio (TBR) values up to 1.21. At the same time, the design of the back supporting structure (BSS), which is a DEMO specific component that has to support the blanket modules inside the vacuum vessel (VV), is ongoing with the aim of maximizing the shielding power and minimizing pumping power. This implies a re-engineering of the modules’ attachment system. Design changes however, will have an impact on the manufacturing and assembly sequences that are being developed for the HCLL-TBM. Due to the differences in joint configurations, thicknesses to be welded, heat dissipation and the various technical constraints related to the accessibility of the welding tools and implementation of non-destructive examination (NDE), the manufacturing procedure should be adapted and optimized for DEMO design. Laser welding instead of TIG could be an option to reduce distortions. The time-of-flight diffraction (TOFD) technique is being investigated for NDE. Finally, essential information expected from the HCLL-TBM program that will be needed to finalize the DEMO design is discussed.

  1. 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.)

  2. 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.

  3. 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.

  4. 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.

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  7. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  8. Pre-conceptual design study on K-DEMO ceramic breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Sung, E-mail: jspark@nfri.re.kr [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kwon, Sungjin; Im, Kihak; Kim, Keeman [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Brown, Thomas; Neilson, George [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2015-11-15

    A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field (B{sub T0} = 7.4 T), R = 6.8 m, a = 2.1 m, and a steady-state operation. The design concepts of the K-DEMO blanket system considering the cooling in-vessel components with pressurized water and a solid pebble breeder are described herein. The structure of the K-DEMO blanket is toroidally subdivided into 16 inboard and 32 outboard sectors, in order to allow the vertical maintenance. Each blanket module is composed of plasma-facing first wall, layers of breeding parts, shielding and manifolds. A ceramic breeder using Li{sub 4}SiO{sub 4} pebbles with Be{sub 12}Ti as neuron multiplier is employed for study. MCNP neutronic simulations and thermo-hydraulic analyses are interactively performed in order to satisfy two key aspects: achieving a global Tritium Breeding Ratio (TBR) >1.05 and operating within the maximum allowable temperature ranges of materials.

  9. 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.

  10. Design analyses of self-cooled liquid metal blankets

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.

    1986-12-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations.

  11. Heat transfer problems in gas-cooled solid blankets

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-01-01

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

  12. 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.

  13. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

  14. The Haida Button Blanket.

    Science.gov (United States)

    Johnson, Vesta

    In the Haida nation, there are two phratries, Eagle and Raven, divided into a number of clans sharing one or more emblems. These emblems, inherited from the mother's line, adorn the button blankets which are the traditional ceremonial robes that serve to identify the family of the wearer. Written instructions and diagrams guide students in…

  15. MHD pressure drop in ferritic pipes of fusion blankets

    Energy Technology Data Exchange (ETDEWEB)

    Reimann, J.; Buehler, Leo E-mail: leo.buehler@iket.fzk.de; Messadek, K.; Stieglitz, R

    2003-09-01

    Magnetohydrodynamic flows in pipes of ferromagnetic material is an important issue for liquid metal blanket concepts using MANET as wall material. Fusion relevant magnetic fields of 4-8 T cause high pressure drop in the blanket header where a massive structure of ferromagnetic material exists. It is briefly outlined that in the blanket the reduction of pressure drop due to magnetic shielding is limited to about 10%. Remarkable reduction of pressure drop is possible by means of electrical insulation that prevents currents from short-circuiting through the very thick walls of the headers. Direct contact of the insulating material with the liquid metal is excluded by using metallic liners. Results are reported on the fabrication of such a test section and corresponding pressure drop measurements confirm the effective contribution of the electrical decoupling.

  16. 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.

  17. 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.

  18. 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.

  19. Tailorable Advanced Blanket Insulation (TABI)

    Science.gov (United States)

    Sawko, Paul M.; Goldstein, Howard E.

    1987-01-01

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

  20. SU-E-T-523: Investigation of Various MR-Compatible Shielding Materials for Direction Modulated Brachytherapy (DMBT) Tandem Applicator for Cervical Cancer Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Safigholi, H; Soliman, A; Song, W [Sunnybrook Research Institute, Sunnybrook Health Sciences Centre, U of T, Toronto, Ontario (Canada); Han, D [Sunnybrook Research Institute, Sunnybrook Health Sciences Centre, U of T, Toronto, Ontario (Canada); University of California, San Diego, La Jolla, CA (United States); Meigooni, A Soleimani [Comprehensive Cancer Center of Nevada, Las Vegas, NV (United States); Scanderbeg, D [UCSD Medical Center, La Jolla, CA (United States)

    2015-06-15

    Purpose: To evaluate various shielding materials such as Gold (Au), Osmium (Os), Tantalum (Ta), and Tungsten (W) based alloys for use with a novel intensity modulation capable direction modulated brachytherapy (DMBT) tandem applicator for image guided cervical cancer HDR brachytherapy. Methods: The novel MRI-compatible DMBT tandem, made from nonmagnetic tungsten-alloy rod with diameter of 5.4 mm, has 6 symmetric peripheral holes of 1.3 mm diameter with 2.05 mm distance from the center for a high degree intensity modulation capacity. The 0.3 mm thickness of bio-compatible plastic tubing wraps the tandem. MCNPX was used for Monte Carlo simulations of the shields and the mHDR Ir-192 V2 source. MC-generated 3D dose matrices of different shielding materials of Au, Os, Ta, and W with 1 mm3 resolution were imported into an in-house-coded inverse optimization planning system to evaluate 19 clinical patient plans. Prescription dose was 15Gy. All plans were normalized to receive the same HRCTV D90. Results: In general, the plan qualities for various shielding materials were similar. The OAR D2cc for bladder was very similar for Au, Os, and Ta with 11.64±2.30Gy. For W, it was very close 11.65±2.30Gy. The sigmoid D2cc was 9.82±2.46Gy for Au and Os while it was 9.84±2.48Gy for Ta and W. The rectum D2cc was 7.44±3.06Gy for Au, 7.43±3.07Gy for Os, 7.48±3.05Gy for Ta, and 7.47±3.05Gy for W. The HRCTV D98 and V100 were very close with 16.37±1.87 Gy and 97.37±1.93 Gy, on average, respectively. Conclusion: Various MRI-compatible shielding alloys were investigated for the DMBT tandem applicator. The clinical plan qualities were not significantly different among these various alloys, however. Therefore, the candidate metals (or in combination) can be used to select best alloys for MRI image guided cervical cancer brachytherapy using the novel DMBT applicator that is capable of unprecedented level of intensity modulation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  2. Design of 1-μm-pitch liquid crystal spatial light modulators having dielectric shield wall structure for holographic display with wide field of view

    Science.gov (United States)

    Isomae, Yoshitomo; Shibata, Yosei; Ishinabe, Takahiro; Fujikake, Hideo

    2017-03-01

    In the development of electronic holographic displays with a wide field of view, one issue is the realization of 1-μm-pitch spatial light modulators (SLMs) using liquid crystal on silicon (LCOS) techniques. We clarified that it is necessary to suppress not only the leakage of fringe electric fields from adjacent pixels but also the effect of elastic forces in the liquid crystal to achieve full-phase modulation (2π) in individual pixels. We proposed a novel LCOS-SLM with a dielectric shield wall structure, and achieved driving of individual 1-μm-pitch pixels. We also investigated the optimum values for width and dielectric constant of the wall structure when enlarging the area that can modulate light in the pixels. These results contribute to the design of 1-μm-pitch LCOS-SLM devices for wide-viewing-angle holographic displays.

  3. 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)

  4. 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 .

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

    Science.gov (United States)

    Sagara, A.; Imagawa, S.; Mitarai, O.; Dolan, T.; Tanaka, T.; Kubota, Y.; Yamazaki, K.; Watanabe, K. Y.; Mizuguchi, N.; Muroga, T.; Noda, N.; Kaneko, O.; Yamada, H.; Ohyabu, N.; Uda, T.; Komori, A.; Sudo, S.; Motojima, O.

    2005-04-01

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

  6. 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.

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, K.M.

    1994-04-01

    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/m{sup 2}. 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.

  9. Liquid immersion blanket design for use in a compact modular fusion reactor

    Science.gov (United States)

    Sorbom, Brandon; Ball, Justin; Barnard, Harold; Haakonsen, Christian; Hartwig, Zachary; Olynyk, Geoffrey; Sierchio, Jennifer; Whyte, Dennis

    2012-10-01

    Traditional tritium breeding blankets in fusion reactor designs include a large amount of structural material. This results in complex engineering requirements, complicated sector maintenance, and marginal tritium breeding ratios (TBR). We present a conceptual design of a fully liquid blanket. To maximize tritium breeding volume, the vacuum vessel is completely immersed in a continuously recycled FLiBe blanket, with the exception of small support posts. FLiBe has a wide liquid temperature window (459 C to 1430 C), low electrical conductivity to minimize MHD effects, similar thermal/fluid characteristics to water, and is chemically inert. While tritium breeding with FLiBe in traditional blankets is poor, we use MCNP neutronics analysis to show that the immersion blanket design coupled with a beryllium neutron multiplier results in TBR > 1. FLiBe is shown to be a sufficient radiation shield for the toroidal field magnets and can be used as a coolant for the vacuum vessel and divertor, allowing for a simplified single-phase, low-pressure, single-fluid cooling scheme. When coupled with a high-field compact reactor design, the immersion blanket eliminates the need for complex sector maintenance, allows the vacuum vessel to be a replaceable component, and reduces financial cost.

  10. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    DEFF Research Database (Denmark)

    Singh, Vishvanath P.; Badiger, Nagappa M.; Gerward, Leif

    2016-01-01

    Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fittin...... combination of low-and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors....

  11. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

  12. Activation calculation analysis for the China 2×6 solid breeder test blanket module%中国2×6固态实验包层模块活化计算分析

    Institute of Scientific and Technical Information of China (English)

    韩静茹; 陈义学; 张国书; 曹启祥

    2011-01-01

    Based on the new design of China 2x6 helium cooled solid breeder (CH-HCSB) test blanket module (TBM), three-dimensional activation calculation analysis was performed by using the Monte Carlo particle transport code MCNP and European activation code FISPACT. The results show that, at the beginning moment after shutdown, the total activity is 1.78×1016Bq, and the total afterheat is 3.01kW. They are both dominated by the structural material CLF-1. Meanwhile, the dominant radioactivity nuclides and reaction pathways have been identified. The results will provide useful indications for materials selection and further optimization design of the TBM. On basis of the calculated contact dose rate, the activated materials can be reused with the remote handling recycling options. That is effective for preventing from the radiation environmental hazard.%基于中国氦冷固态实验包层模块(CH-HCSB-TBM)的新设计方案,采用蒙特卡罗粒子输运程序MCNP和欧洲活化计算程序FISPACT,对CH-HCSB-TBM进行了三维活化计算分析.计算结果表明,停堆初期TBM总的放射性活度、衰变余热分别为1.78× 1016Bq和3.01kW,主要受结构材料CLF-1影响.同时给出了影响TBM材料活化特性的主要核素及其生成途径,为TBM设计的材料选取和优化提供依据.根据计算的停堆剂量率可知,TBM中的活化材料都能采取远程操作实现再循环利用,可有效防止放射性环境危害问题.

  13. 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.

  14. 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.).

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

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

  16. 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)

  17. Packed fluidized bed blanket for fusion reactor

    Science.gov (United States)

    Chi, John W. H.

    1984-01-01

    A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

  18. 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)

  19. RAMI analysis for DEMO HCPB blanket concept cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

    2015-10-15

    Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

  20. Rapid modulation of ultraviolet shielding in plants is influenced by solar ultraviolet radiation and linked to alterations in flavonoids.

    Science.gov (United States)

    Barnes, Paul W; Tobler, Mark A; Keefover-Ring, Ken; Flint, Stephan D; Barkley, Anne E; Ryel, Ronald J; Lindroth, Richard L

    2016-01-01

    The accumulation of ultraviolet (UV)-absorbing compounds (flavonoids and related phenylpropanoids) and the resultant decrease in epidermal UV transmittance (TUV ) are primary protective mechanisms employed by plants against potentially damaging solar UV radiation and are critical components of the overall acclimation response of plants to changing solar UV environments. Whether plants can adjust this UV sunscreen protection in response to rapid changes in UV, as occurs on a diurnal basis, is largely unexplored. Here, we use a combination of approaches to demonstrate that plants can modulate their UV-screening properties within minutes to hours, and these changes are driven, in part, by UV radiation. For the cultivated species Abelmoschus esculentus, large (30-50%) and reversible changes in TUV occurred on a diurnal basis, and these adjustments were associated with changes in the concentrations of whole-leaf UV-absorbing compounds and several quercetin glycosides. Similar results were found for two other species (Vicia faba and Solanum lycopersicum), but no such changes were detected in Zea mays. These findings reveal a much more dynamic UV-protection mechanism than previously recognized, raise important questions concerning the costs and benefits of UV-protection strategies in plants and have practical implications for employing UV to enhance crop vigor and quality in controlled environments.

  1. Lithium as a blanket coolant

    Energy Technology Data Exchange (ETDEWEB)

    Wells, W.M.

    1977-01-01

    Recent re-assessment of tokamak reactors which move towards smaller size and lower required field strength (higher beta)/sup 2/ change the picture as regards the magnitude of MHD effects on flow resistance for lithium coolant. Perhaps the most important consequence of this as regards use of this coolant is that of clear acceptability of such effects when the flow is predominantly transverse to the magnetic field. This permits defining a blanket that consists entirely of round tubes containing the circulated lithium with voids between the tubes. Required thermal-hydraulic calculations are then on bases which are well established, especially in view of recent results dealing with perturbations of ducts and magnetic fields. Mitigation of MHD effects is feasible through tapering of tube wall thickness or use of insulated layers, but their use was not mandatory for the assumed conditions. Blanket configurations utilizing flowing lithium in round tubes immersed in static lithium may be suitable, but calculational methods do not now exist for this situation. Use of boiling potassium or cesium appears to be prohibitive in terms of vapor flow area when temperature levels are consistent with stainless steel. Liquid sodium, in addition to not being a breeding material, requires higher velocity than lithium for the same heat removal.

  2. 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.

  3. 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.

  4. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

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

  5. Shielding requirements in helical tomotherapy

    Science.gov (United States)

    Baechler, S.; Bochud, F. O.; Verellen, D.; Moeckli, R.

    2007-08-01

    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.

  6. Multifractal Framework Based on Blanket Method

    Science.gov (United States)

    Paskaš, Milorad P.; Reljin, Irini S.; Reljin, Branimir D.

    2014-01-01

    This paper proposes two local multifractal measures motivated by blanket method for calculation of fractal dimension. They cover both fractal approaches familiar in image processing. The first two measures (proposed Methods 1 and 3) support model of image with embedded dimension three, while the other supports model of image embedded in space of dimension three (proposed Method 2). While the classical blanket method provides only one value for an image (fractal dimension) multifractal spectrum obtained by any of the proposed measures gives a whole range of dimensional values. This means that proposed multifractal blanket model generalizes classical (monofractal) blanket method and other versions of this monofractal approach implemented locally. Proposed measures are validated on Brodatz image database through texture classification. All proposed methods give similar classification results, while average computation time of Method 3 is substantially longer. PMID:24578664

  7. Advanced Multifunctional MMOD Shield: Radiation Shielding Assessment

    Science.gov (United States)

    Rojdev, Kristina; Christiansen, Eric

    2013-01-01

    Deep space missions must contend with a harsh radiation environment Impacts to crew and electronics. Need to invest in multifunctionality for spacecraft optimization. MMOD shield. Goals: Increase radiation mitigation potential. Retain overall MMOD shielding performance.

  8. 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.

  9. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    Directory of Open Access Journals (Sweden)

    Singh Vishvanath P.

    2016-01-01

    Full Text Available Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fitting at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neutron removal cross-section. It is shown that ZrH2 and VH2 are very good shielding materials for gamma rays and fast neutrons due to their suitable combination of low- and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rozov, Vladimir, E-mail: vladimir.rozov@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Belyakov, V.; Kukhtin, V.; Lamzin, E.; Mazul, I.; Sytchevsky, S. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation)

    2014-11-15

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

  11. 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.

  12. 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)

  13. FEPI-MB: identifying SNPs-disease association using a Markov Blanket-based approach

    Directory of Open Access Journals (Sweden)

    Han Bing

    2011-11-01

    Full Text Available Abstract Background The interactions among genetic factors related to diseases are called epistasis. With the availability of genotyped data from genome-wide association studies, it is now possible to computationally unravel epistasis related to the susceptibility to common complex human diseases such as asthma, diabetes, and hypertension. However, the difficulties of detecting epistatic interaction arose from the large number of genetic factors and the enormous size of possible combinations of genetic factors. Most computational methods to detect epistatic interactions are predictor-based methods and can not find true causal factor elements. Moreover, they are both time-consuming and sample-consuming. Results We propose a new and fast Markov Blanket-based method, FEPI-MB (Fast EPistatic Interactions detection using Markov Blanket, for epistatic interactions detection. The Markov Blanket is a minimal set of variables that can completely shield the target variable from all other variables. Learning of Markov blankets can be used to detect epistatic interactions by a heuristic search for a minimal set of SNPs, which may cause the disease. Experimental results on both simulated data sets and a real data set demonstrate that FEPI-MB significantly outperforms other existing methods and is capable of finding SNPs that have a strong association with common diseases. Conclusions FEPI-MB algorithm outperforms other computational methods for detection of epistatic interactions in terms of both the power and sample-efficiency. Moreover, compared to other Markov Blanket learning methods, FEPI-MB is more time-efficient and achieves a better performance.

  14. The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

    Energy Technology Data Exchange (ETDEWEB)

    Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse

    1988-03-01

    We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs.

  15. 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.

  16. 48 CFR 613.303 - Blanket purchase agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Blanket purchase agreements (BPAs). 613.303 Section 613.303 Federal Acquisition Regulations System DEPARTMENT OF STATE....303 Blanket purchase agreements (BPAs)....

  17. 48 CFR 1313.303 - Blanket Purchase Agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Blanket Purchase Agreements (BPAs). 1313.303 Section 1313.303 Federal Acquisition Regulations System DEPARTMENT OF COMMERCE....303 Blanket Purchase Agreements (BPAs)....

  18. 48 CFR 13.303 - Blanket purchase agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Blanket purchase agreements (BPAs). 13.303 Section 13.303 Federal Acquisition Regulations System FEDERAL ACQUISITION... Methods 13.303 Blanket purchase agreements (BPAs)....

  19. 48 CFR 313.303 - Blanket purchase agreements.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Blanket purchase agreements. 313.303 Section 313.303 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES....303 Blanket purchase agreements....

  20. Review: BNL graphite blanket design concepts

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-03-01

    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.

  1. The climatic impact of supervolcanic ash blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Morgan T.; Sparks, R.S.J. [University of Bristol, Department of Earth Sciences, Bristol (United Kingdom); Valdes, Paul J. [University of Bristol, School of Geographical Sciences, Bristol (United Kingdom)

    2007-11-15

    Supervolcanoes are large caldera systems that can expel vast quantities of ash, volcanic gases in a single eruption, far larger than any recorded in recent history. These super-eruptions have been suggested as possible catalysts for long-term climate change and may be responsible for bottlenecks in human and animal populations. Here, we consider the previously neglected climatic effects of a continent-sized ash deposit with a high albedo and show that a decadal climate forcing is expected. We use a coupled atmosphere-ocean General Circulation Model (GCM) to simulate the effect of an ash blanket from Yellowstone volcano, USA, covering much of North America. Reflectivity measurements of dry volcanic ash show albedo values as high as snow, implying that the effects of an ash blanket would be severe. The modeling results indicate major disturbances to the climate, particularly to oscillatory patterns such as the El Nino Southern Oscillation (ENSO). Atmospheric disruptions would continue for decades after the eruption due to extended ash blanket longevity. The climatic response to an ash blanket is not significant enough to investigate a change to stadial periods at present day boundary conditions, though this is one of several impacts associated with a super-eruption which may induce long-term climatic change. (orig.)

  2. Fidget Blankets: A Sensory Stimulation Outreach Program.

    Science.gov (United States)

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

    2016-01-01

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

  3. 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

  4. Multi-Sensor Data Fusion Technologies for Blanket Jamming Localization

    Institute of Scientific and Technical Information of China (English)

    WANG Ju; WU Si-liang; ZENG Tao

    2005-01-01

    The localization of the blanket jamming is studied and a new method of solving the localization ambiguity is proposed. Radars only can acquire angle information without range information when encountering the blanket jamming. Netted radars could get position information of the blanket jamming by make use of radars' relative position and the angle information, when there is one blanket jamming. In the presence of error, the localization method and the accuracy analysis of one blanket jamming are given. However, if there are more than one blanket jamming, and the two blanket jamming and two radars are coplanar, the localization of jamming could be error due to localization ambiguity. To solve this confusion, the Kalman filter model is established for all intersections, and through the initiation and association algorithm of multi-target, the false intersection can be eliminated. Simulations show that the presented method is valid.

  5. The Active Muon Shield

    CERN Document Server

    Bezshyiko, Iaroslava

    2016-01-01

    In the SHiP beam-dump of the order of 1011 muons will be produced per second. An active muon-shield is used to magnetically deflect these muons out of the acceptance of the spectrom- eter. This note describes how this shield is modelled and optimized. The SHiP spectrometer is being re-optimized using a conical decay-vessel, and utilizing the possibility to magnetize part of the beam-dump shielding iron. A shield adapted to these new conditions is presented which is significantly shorter and lighter than the shield used in the Technical Proposal (TP), while showing a similar performance.

  6. Reduced activation martensitic steels as a structural material for ITER test blanket

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, K. E-mail: shiba@realab01.tokai.jaeri.go.jp; Enoeda, M.; Jitsukawa, S

    2004-08-01

    A Japanese ITER test blanket module (TBM) is planed to use reduced-activation martensitic steel F82H. Feasibility of F82H for ITER test blanket module is discussed in this paper. Several kinds of property data, including physical properties, magnetic properties, mechanical properties and neutron-irradiation data on F82H have been obtained, and these data are complied into a database to be used for the designing of the ITER TBM. Currently obtained data suggests F82H will not have serious problems for ITER TBM. Optimization of F82H improves the induced activity, toughness and HIP resistance. Furthermore, modified F82H is resistant to temperature instability during material production.

  7. 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.).

  8. International Space Station Radiation Shielding Model Development

    Science.gov (United States)

    Qualls, G. D.; Wilson, J. W.; Sandridge, C.; Cucinotta, F. A.; Nealy, J. E.; Heinbockel, J. H.; Hugger, C. P.; Verhage, J.; Anderson, B. M.; Atwell, W.

    2001-01-01

    The projected radiation levels within the International Space Station (ISS) have been criticized by the Aerospace Safety Advisory Panel in their report to the NASA Administrator. Methods for optimal reconfiguration and augmentation of the ISS shielding are now being developed. The initial steps are to develop reconfigurable and realistic radiation shield models of the ISS modules, develop computational procedures for the highly anisotropic radiation environment, and implement parametric and organizational optimization procedures. The targets of the redesign process are the crew quarters where the astronauts sleep and determining the effects of ISS shadow shielding of an astronaut in a spacesuit. The ISS model as developed will be reconfigurable to follow the ISS. Swapping internal equipment rack assemblies via location mapping tables will be one option for shield optimization. Lightweight shield augmentation materials will be optimally fit to crew quarter areas using parametric optimization procedures to minimize the augmentation shield mass. The optimization process is being integrated into the Intelligence Synthesis Environment s (ISE s) immersive simulation facility at the Langley Research Center and will rely on High Performance Computing and Communication (HPCC) for rapid evaluation of shield parameter gradients.

  9. 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.

  10. Stellar model atmospheres with magnetic line blanketing

    CERN Document Server

    Kochukhov, O; Shulyak, D

    2004-01-01

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

  11. Chicxulub Ejecta Blanket Deposits From Belize

    Science.gov (United States)

    Ocampo, A.

    1995-01-01

    The Chicxulub impact into a thick sequence of carbonates and sulfates released over a trillion tons of volatiles. The importance of the explosive release of such a large mass of volatiles has been greatly underestimated in studies of ejecta depositional processes. Proximal Chicxulub ejecta blanket deposits recent discovered on Albion Island in Belize provide a key to understanding the role of volatile-rich target material during large impact events.

  12. 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.

  13. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Balaraman Kumar

    2010-06-01

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

  14. 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.

  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. Monte Carlo radiation shielding and activation analyses for the Diagnostic Equatorial Port Plug in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, A., E-mail: arkady.serikov@kit.edu [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Fischer, U.; Leichtle, D. [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pitcher, C.S. [ITER Organization, Route de Vinon sur Verdon, 13115, St. Paul lez Durance (France)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Systematic neutronics analyses were conducted to assess the ITER Equatorial Port Plug radiation shielding performance. Black-Right-Pointing-Pointer Shielding optimization was achieved by parametric analyses of several design variants using the MCNP5, FISPACT-2007, and R2Smesh codes. Black-Right-Pointing-Pointer Dominant effect of radiation streaming along the port plug gaps was recognized. Black-Right-Pointing-Pointer Combination of the gap labyrinths and streaming stoppers or rails reduces shutdown doses by 2 orders of magnitude. Black-Right-Pointing-Pointer Using the proposed shielding, the shutdown dose in the ITER port interspace is less than the personnel access limit of 100 {mu}Sv/h. - Abstract: This paper addresses neutronics aspects of the design development of the Diagnostic Generic Equatorial Port Plug (EPP) in ITER. To secure the personnel access at the EPP back-end interspace, parametric neutronics analyses of the EPP radiation environment have been performed and practical shielding solutions have been found. Radiation transport was performed with the Monte Carlo MCNP5 code. Activation calculations were conducted with the FISPACT-2007 inventory code. The R2Smesh approach was applied to couple transport and activation calculations. Newly created EPP local MCNP5 model was devised by extracting the EPP and adjacent blanket modules from the ITER Alite-4.1 model with proper modification of the EPP geometry in accordance with recent 3D CAD CATIA model. The EPP local model reproduces the EPP neutronically important features and allows investigation of the EPP neutronics effects in isolation from all other ITER components. Thorough EPP parametric analyses revealed dominant effect of gaps around EPP and several EPP design improvements were implemented as the outcomes of the analyses. Gap labyrinths and streaming stoppers inserted into the gaps were shown are capable to reduce the shutdown dose rate which is below the 100

  17. 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...

  18. Overview on ITER and DEMO blanket fabrication activities of the KIT INR and related frameworks

    Energy Technology Data Exchange (ETDEWEB)

    Neuberger, Heiko, E-mail: heiko.neuberger@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Karlsruhe (Germany); Rey, Joerg; Weth, Axel von der; Hernandez, Francisco [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Karlsruhe (Germany); Martin, Tatiana [Karlsruhe Institute of Technology (KIT), Institute for Applied materials, Karlsruhe (Germany); Zmitko, Milan [Fusion for energy, ITER Department, Test Blanket Modules and Materials Development Project Team, Barcelona (Spain); Felde, Alexander [Institut für Umformtechnik (IFU), Universität Stuttgart (Germany); Niewöhner, Reinhard [Forschungsgesellschaft Umformtechnik (FGU), Stuttgart (Germany); Krüger, Friedhelm [Krüger Erodiertechnik, Biedenkopf (Germany)

    2015-10-15

    Highlights: • Recent achievements in fabricaition within different frameworks. • First Wall mockup with erosion technology. • Manufacturing of a HCPB TBM Cooling Plate Mockup (F4E) - Abstract: Fabrication experiments have been carried out in the KIT with the goal to qualify manufacturing technologies for the realization of fusion reactor components. The main focus of the activities managed by the fabrication team in the Institute of Neutron Physics and reactor technologies (INR) has been on the Test Blanket Module for ITER. Sets of fabrication and welding procedure specifications have been demonstrated and qualified in relevant scale for TBM structural and functional components. This paper presents interactions in between the different frameworks on domestic and European level to underline backgrounds of developments. It also summarizes results of development and their relevancy for DEMO and gives an outlook on the future development strategy for the DEMO blanket fabrication.

  19. 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.

  20. 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.

  1. Radiation Shielding Materials

    Science.gov (United States)

    Adams, James H., Jr.; Rose, M. Franklin (Technical Monitor)

    2001-01-01

    NASA has relied on the materials to provide radiation shielding for astronauts since the first manned flights. Until very recently existing materials in the structure of manned spacecraft as well as the equipment and consumables onboard have been taken advantage of for radiation shielding. With the advent of the International Space Station and the prospect of extended missions to the Moon or Mars, it has been found that the materials, which were included in the spacecraft for other reasons, do not provide adequate shielding. For the first time materials are being added to manned missions solely to improve the radiation shielding. It is now recognized that dual use materials must be identified/developed. These materials must serve a purpose as part of the spacecraft or its cargo and at the same time be good shielding. This paper will review methods for evaluating the radiation shielding effectiveness of materials and describe the character of materials that have high radiation shielding effectiveness. Some candidate materials will also be discussed.

  2. 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,...

  3. 48 CFR 213.303 - Blanket purchase agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 3 2010-10-01 2010-10-01 false Blanket purchase agreements (BPAs). 213.303 Section 213.303 Federal Acquisition Regulations System DEFENSE ACQUISITION... PROCEDURES Simplified Acquisition Methods 213.303 Blanket purchase agreements (BPAs)....

  4. 48 CFR 8.405-3 - Blanket purchase agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Blanket purchase... Blanket purchase agreements (BPAs). (a)(1) Establishment. Ordering activities may establish BPAs under any..., before placing an order exceeding the micro-purchase threshold, the ordering activity shall— (i)...

  5. 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.

  6. 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.

  7. Shielding high energy accelerators

    CERN Document Server

    Stevenson, Graham Roger

    2001-01-01

    After introducing the subject of shielding high energy accelerators, point source, line-of-sight models, and in particular the Moyer model. are discussed. Their use in the shielding of proton and electron accelerators is demonstrated and their limitations noted. especially in relation to shielding in the forward direction provided by large, flat walls. The limitations of reducing problems to those using it cylindrical geometry description are stressed. Finally the use of different estimators for predicting dose is discussed. It is suggested that dose calculated from track-length estimators will generally give the most satisfactory estimate. (9 refs).

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

    Energy Technology Data Exchange (ETDEWEB)

    Li, Min; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Lv, Zhongliang; Ye, Minyou

    2015-02-15

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

  9. 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.

  10. 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.

  11. Neutronic Reactor Shield

    Science.gov (United States)

    Fermi, Enrico; Zinn, Walter H.

    The argument of the present Patent is a radiation shield suitable for protection of personnel from both gamma rays and neutrons. Such a shield from dangerous radiations is achieved to the best by the combined action of a neutron slowing material (a moderator) and a neutron absorbing material. Hydrogen is particularly effective for this shield since it is a good absorber of slow neutrons and a good moderator of fast neutrons. The neutrons slowed down by hydrogen may, then, be absorbed by other materials such as boron, cadmium, gadolinium, samarium or steel. Steel is particularly convenient for the purpose, given its effectiveness in absorbing also the gamma rays from the reactor (both primary gamma rays and secondary ones produced by the moderation of neutrons). In particular, in the present Patent a shield is described, made of alternate layers of steel and Masonite (an hydrolized ligno-cellulose material). The object of the present Patent is not discussed in any other published paper.

  12. Adhesive particle shielding

    Science.gov (United States)

    Klebanoff, Leonard Elliott; Rader, Daniel John; Walton, Christopher; Folta, James

    2009-01-06

    An efficient device for capturing fast moving particles has an adhesive particle shield that includes (i) a mounting panel and (ii) a film that is attached to the mounting panel wherein the outer surface of the film has an adhesive coating disposed thereon to capture particles contacting the outer surface. The shield can be employed to maintain a substantially particle free environment such as in photolithographic systems having critical surfaces, such as wafers, masks, and optics and in the tools used to make these components, that are sensitive to particle contamination. The shield can be portable to be positioned in hard-to-reach areas of a photolithography machine. The adhesive particle shield can incorporate cooling means to attract particles via the thermophoresis effect.

  13. 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)

  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. 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.

  16. 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.

  17. The development of ferritic steels for DEMO blanket

    Energy Technology Data Exchange (ETDEWEB)

    Kohyama, A. [Kyoto Univ. (Japan). Inst. of Advanced Energy; Hishinuma, A.; Shiba, K. [Tokai Establishment, JAERI, Tokai, Ibaraki (Japan); Kohno, Y. [Department of Materials Science, University of Tokyo, Hongo, Tokyo 113 (Japan); Sagara, A. [National Institute for Fusion Science, Toki, Gifu (Japan)

    1998-09-01

    The development of low-activation ferritic/martensitic steels is a key to the achievement of nuclear fusion as a safe, environmentally attractive and economically competitive energy source. The Japanese and the European Fusion Materials programs have put low-activation ferritic and martensitic steels R and D at the highest priority for a demonstration reactor (DEMO) and the beyond. An international collaborative test program on low-activation ferritic/martensitic steels for fusion is in progress as an activity of the International Energy Agency (IEA) fusion materials working group to verify the feasibility of using ferritic/martensitic steels for fusion by an extensive test program covering the most relevant technical issues for the qualification of a material for a nuclear application. The development of a comprehensive data base on the representative industrially processed reduced-activation steels of type 8-9Cr-2WVTa is underway for providing designers a preliminary set of material data for the mechanical design of components, e.g. for DEMO relevant blanket modules. The current design status of FFHR and SSTR utilizing low-activation ferritic steels is reviewed and future prospects are defined. (orig.) 12 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

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

  19. 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...

  20. 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

  1. 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...

  2. Performance of uncoated AFRSI blankets during multiple Space Shuttle flights

    Science.gov (United States)

    Sawko, Paul M.; Goldstein, Howard E.

    1992-01-01

    Uncoated Advanced Flexible Reusable Surface Insulation (AFRSI) blankets were successfully flown on seven consecutive flights of the Space Shuttle Orbiter OV-099 (Challenger). In six of the eight locations monitored (forward windshield, forward canopy, mid-fuselage, upper wing, rudder/speed brake, and vertical tail) the AFRSI blankets performed well during the ascent and reentry exposure to the thermal and aeroacoustic environments. Several of the uncoated AFRSI blankets that sustained minor damage, such as fraying or broken threads, could be repaired by sewing or by patching with a surface coating called C-9. The chief reasons for replacing or completely coating a blanket were fabric embrittlement and fabric abrasion caused by wind erosion. This occurred in the orbiter maneuvering system (OMS) pod sidewall and the forward mid-fuselage locations.

  3. 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 performance...

  4. 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...

  5. 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...

  6. 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.

  7. 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)

  8. 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.

  9. [Calculation of radiation loads in a space station compartment with a secondary shielding].

    Science.gov (United States)

    Kartashov, D A; Tolochek, R V; Shurshakov, V A; Yarmanova, E N

    2013-01-01

    Doses from space ionizing radiation were estimated using a model of ISS cosmonaut's quarters (CQ) outfitted with secondary shielding ("Protective shutter" (PS) as part of experiment MATRYOSHKA-R). Protective shutter is a "blanket" of water-containing material with mass thickness of - 6 g/cm2 covering the CQ exterior wall. Calculation was performed specifically for locations of experimental dosimetry assemblies. Agreement of calculations and experimental data reaching accuracy - 15% proves model applicability to estimating protective effectiveness of secondary shielding in the present-day and future space vehicles. This shielding may reduce radiation loading onto crewmembers as an equivalent dose by more than 40% within a broad range of orbit altitudes equally during the solar minimum and maximum.

  10. 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.

  11. 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.

  12. Shielding design for a laser-accelerated proton therapy system.

    Science.gov (United States)

    Fan, J; Luo, W; Fourkal, E; Lin, T; Li, J; Veltchev, I; Ma, C-M

    2007-07-07

    In this paper, we present the shielding analysis to determine the necessary neutron and photon shielding for a laser-accelerated proton therapy system. Laser-accelerated protons coming out of a solid high-density target have broad energy and angular spectra leading to dose distributions that cannot be directly used for therapeutic applications. A special particle selection and collimation device is needed to generate desired proton beams for energy- and intensity-modulated proton therapy. A great number of unwanted protons and even more electrons as a side-product of laser acceleration have to be stopped by collimation devices and shielding walls, posing a challenge in radiation shielding. Parameters of primary particles resulting from the laser-target interaction have been investigated by particle-in-cell simulations, which predicted energy spectra with 300 MeV maximum energy for protons and 270 MeV for electrons at a laser intensity of 2 x 10(21) W cm(-2). Monte Carlo simulations using FLUKA have been performed to design the collimators and shielding walls inside the treatment gantry, which consist of stainless steel, tungsten, polyethylene and lead. A composite primary collimator was designed to effectively reduce high-energy neutron production since their highly penetrating nature makes shielding very difficult. The necessary shielding for the treatment gantry was carefully studied to meet the criteria of head leakage shield neutrons and an outer layer of lead was used to reduce photon dose from neutron capture and electron bremsstrahlung. It is shown that the two-layer shielding design with 10-12 cm thick polyethylene and 4 cm thick lead can effectively absorb the unwanted particles to meet the shielding requirements.

  13. Shielding design for a laser-accelerated proton therapy system

    Science.gov (United States)

    Fan, J.; Luo, W.; Fourkal, E.; Lin, T.; Li, J.; Veltchev, I.; Ma, C.-M.

    2007-07-01

    In this paper, we present the shielding analysis to determine the necessary neutron and photon shielding for a laser-accelerated proton therapy system. Laser-accelerated protons coming out of a solid high-density target have broad energy and angular spectra leading to dose distributions that cannot be directly used for therapeutic applications. A special particle selection and collimation device is needed to generate desired proton beams for energy- and intensity-modulated proton therapy. A great number of unwanted protons and even more electrons as a side-product of laser acceleration have to be stopped by collimation devices and shielding walls, posing a challenge in radiation shielding. Parameters of primary particles resulting from the laser-target interaction have been investigated by particle-in-cell simulations, which predicted energy spectra with 300 MeV maximum energy for protons and 270 MeV for electrons at a laser intensity of 2 × 1021 W cm-2. Monte Carlo simulations using FLUKA have been performed to design the collimators and shielding walls inside the treatment gantry, which consist of stainless steel, tungsten, polyethylene and lead. A composite primary collimator was designed to effectively reduce high-energy neutron production since their highly penetrating nature makes shielding very difficult. The necessary shielding for the treatment gantry was carefully studied to meet the criteria of head leakage shield neutrons and an outer layer of lead was used to reduce photon dose from neutron capture and electron bremsstrahlung. It is shown that the two-layer shielding design with 10-12 cm thick polyethylene and 4 cm thick lead can effectively absorb the unwanted particles to meet the shielding requirements.

  14. 75 FR 38459 - Certain Woven Electric Blankets From the People's Republic of China: Final Determination of Sales...

    Science.gov (United States)

    2010-07-02

    ... Antidumping Investigations involving Non-Market Economy Countries,'' which states: \\23\\ See Certain Woven... International Trade Administration Certain Woven Electric Blankets From the People's Republic of China: Final... Department'') has determined that certain woven electric blankets (``woven electric blankets'') from...

  15. 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

  16. Numerical simulation of the transient thermal-hydraulic behaviour of the ITER blanket cooling system under the draining operational procedure

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • ITER blanket cooling system hydraulic behaviour is studied under draining transient. • A computational approach based on the finite volume method has been followed. • Draining efficiency has been assessed in term of transient duration and residual water. • Transient duration ranges from ∼40 to 50 s, under the reference draining scenario. • Residual water is predicted to range from few tens of gram up to few kilograms. - Abstract: Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed.

  17. 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.

  18. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  19. 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.

  20. 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. 

  1. 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.

  2. Shielding in Mental Health Hospitals

    Directory of Open Access Journals (Sweden)

    Espen W. Haugom

    2016-02-01

    Full Text Available Shielding is defined as the confinement of patients to a single room or a separate unit/area inside the ward, accompanied by a member of staff. It is understood as both a treatment and a control. The purpose of this study is to examine how staff in psychiatric hospitals describe and assess shielding. This qualitative study uses a descriptive and exploratory design with an inductive approach. The material was acquired through the Acute Network (in Psychiatry nationwide shielding project. Data collection was carried out by the staff, who described the shielding procedure on a semi-structured form. The analysis was inspired by Graneheim and Lundman’s qualitative content analysis. Shielding has been described as an ambiguous practice, that is, shielding can be understood in several ways. There is a clear tension between shielding as a control and shielding as a treatment, with control being described as more important. The important therapeutic elements of shielding have also been mentioned, and shielding involves isolation to different degrees.

  3. 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.

  4. 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.

  5. 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.

  6. Justification for Shielded Receiver Tube Additional Lead Shielding

    Energy Technology Data Exchange (ETDEWEB)

    BOGER, R.M.

    2000-04-11

    In order to reduce high radiation dose rates encountered when core sampling some radioactive waste tanks the addition of 240 lbs. of lead shielding is being considered to the shielded receiver tube on core sample trucks No.1, No.3 and No.4. The lead shielding is 4 inch diameter x 1/2 inch thick half rounds that have been installed around the SR tube over its' full length. Using three unreleased but independently reviewed structural analyses HNF-6018 justifies the addition of the lead shielding.

  7. The axion shield

    CERN Document Server

    Andrianov, A A; Mescia, F; Renau, A

    2010-01-01

    We investigate the propagation of a charged particle in a spatially constant, but time dependent, pseudoscalar background. Physically this pseudoscalar background could be provided by a relic axion density. The background leads to an explicit breaking of Lorentz invariance; as a consequence the process p-> p gamma is possible and the background acts as a shield against extremely energetic cosmic rays, an effect somewhat similar to the GZK cut-off effect. The effect is model independent and can be computed exactly. The hypothetical detection of the photons radiated via this mechanism would provide an indirect way of verifying the cosmological relevance of axions.

  8. Watching a disappearing shield

    Science.gov (United States)

    Stolarski, Richard S.

    1988-10-01

    The remote-sensing techniques used to monitor atmospheric ozone levels are reviewed, and recent results are discussed. The importance of the ozone layer as a shield for UV radiation is stressed, and the impact of human activities generating ozone-destroying compounds is considered. Ground-based, airborne, balloon-borne, and satellite remote-sensing methods are shown to complement each other to provide both global coverage and detailed structural information. Data obtained with the Nimbus-7 TOMS and solar-backscatter UV instruments are presented in graphs and briefly characterized.

  9. 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.

  10. 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 ...

  11. 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....

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

  13. Technical issues for beryllium use in fusion blanket applications

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-01-01

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

  14. 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

  15. Assessment of alkali metal coolants for the ITER blanket

    Science.gov (United States)

    Natesan, K.; Reed, C. B.; Mattas, R. F.

    1994-06-01

    The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The blanket comparison and selection study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper addresses the thermodynamics of interactions between the liquid metals (e.g., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data are used to assess the long-term performance of the first wall in a liquid metal environment. Other key issues include development of electrical insulator coatings on the first-wall structural material to MHD pressure drop, and tritium permeation/inventory in self-cooled and indirectly cooled concepts. Acceptable types of coatings (based on their chemical compatibility and physical properties) are identified, and surface-modification avenues to achieve these coatings on the first wall are discussed. The assessment examines the extent of our knowledge on structural materials performance in liquid metals and identifies needed research and development in several of the areas in order to establish performance envelopes for the first wall in a liquid-metal environment.

  16. Radiation Shielding Optimization on Mars

    Science.gov (United States)

    Slaba, Tony C.; Mertens, Chris J.; Blattnig, Steve R.

    2013-01-01

    Future space missions to Mars will require radiation shielding to be optimized for deep space transit and an extended stay on the surface. In deep space, increased shielding levels and material optimization will reduce the exposure from most solar particle events (SPE) but are less effective at shielding against galactic cosmic rays (GCR). On the surface, the shielding provided by the Martian atmosphere greatly reduces the exposure from most SPE, and long-term GCR exposure is a primary concern. Previous work has shown that in deep space, additional shielding of common materials such as aluminum or polyethylene does not significantly reduce the GCR exposure. In this work, it is shown that on the Martian surface, almost any amount of aluminum shielding increases exposure levels for humans. The increased exposure levels are attributed to neutron production in the shield and Martian regolith as well as the electromagnetic cascade induced in the Martian atmosphere. This result is significant for optimization of vehicle and shield designs intended for the surface of Mars.

  17. 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)

  18. 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.

  19. 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.

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-15

    -up fabrication was started; Cu/SS joints were fabricated and purchase of Be tiles was prepared. Fabrication manual and test manual such as mechanical tests and NDE were documented in the form of the TSD. Based on the design by the ITER-O, 3D modeling of the module no. 4 for ITER blanket FW was produced, thermal-hydraulic and thermo-mechanical analysis were performed. The developed NDE methods were applied to all fabricated mock-ups before HHF test and the UT results were compared with the IR images, which were generated when screening test during HHF test. ECT probes were prepared according to the previous simulation results and they were evaluated experimentally with the NDT mock-up, which has artificial defects. The developed NDE methods and their application were documented as an inspection manual and a QC document, and they were included in the TS000.

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.L. [Design and Accelerator Operations Consulting, 568 Wintergreen Ct Ridge, NY 11961 (United States); Ghosh, V.J.; Breitfeller, M. [NSLS-II, Brookhaven National Laboratory, Upton, NY 11973 (United States)

    2016-08-11

    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.

  4. 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.

  5. Laboratory experiments on drought and runoff in blanket peat

    OpenAIRE

    Holden, J; Burt, T. P.

    2002-01-01

    Global warming might change the hydrology of upland blanket peats in Britain. We have therefore studied in laboratory experiments the impact of drought on peat from the North Pennines of the UK. Runoff was dominated by surface and near-surface flow; flow decreased rapidly with depth and differed from one type of cover to another. Infiltration depended on the intensity of rain, and runoff responded rapidly to rain, with around 50% of rainwater emerging as overland flow. Drought changed the str...

  6. Development of insulating coatings for liquid metal blankets

    Energy Technology Data Exchange (ETDEWEB)

    Malang, S.; Borgstedt, H.U. [Kernforschungszentrum Karlsruhe GmbH (Germany); Farnum, E.H. [Los Alamos National Lab., NM (United States); Natesan, K. [Argonne National Lab., IL (United States); Vitkovski, I.V. [Efremov Inst., St. Petersburg (Russian Federation). MHD-Machines Lab.

    1994-07-01

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

  7. MFTF-B Upgrade for blanket-technology testing

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I.; Doggett, J.N.; Logan, B.G.

    1982-10-22

    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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Hybrid Shielding for Magnetic Fields

    Science.gov (United States)

    Mullins, David; Royal, Kevin

    2017-01-01

    Precision symmetry measurements such as the search for the electric dipole moment of the neutron require magnetic shielding rooms to reduce the ambient field to the pT scale. The massive mu-metal sheets and large separation between layers make these shield rooms bulky and expensive. Active field cancellation systems used to reduce the surrounding field are limited in uniformity of cancellation. A novel approach to reducing the space between shield layers and increasing the effectiveness of active cancellation is to combine the two systems into a hybrid system, with active and passive layers interspersed. We demonstrate this idea in a prototype with an active layer sandwiched between two passive layers of shielding.

  13. Radiation shielding for neutron guides

    Science.gov (United States)

    Ersez, T.; Braoudakis, G.; Osborn, J. C.

    2006-11-01

    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.

  14. Structural/Radiation-Shielding Epoxies

    Science.gov (United States)

    Connell, John W.; Smith, Joseph G.; Hinkley, Jeffrey; Blattnig, Steve; Delozier, Donavon M.; Watson, Kent A.; Ghose, Sayata

    2009-01-01

    A development effort was directed toward formulating epoxy resins that are useful both as structural materials and as shielding against heavy-ion radiation. Hydrogen is recognized as the best element for absorbing heavy-ion radiation, and high-hydrogen-content polymers are now in use as shielding materials. However, high-hydrogen-content polymers (e.g. polyethylene) are typically not good structural materials. In contrast, aromatic polymers, which contain smaller amounts of hydrogen, often have the strength necessary for structural materials. Accordingly, the present development effort is based on the concept that an ideal structural/ heavy-ion-radiation-shielding material would be a polymer that contains sufficient hydrogen (e.g., in the form of aliphatic molecular groups) for radiation shielding and has sufficient aromatic content for structural integrity.

  15. Neutron Shielding Effectiveness of Multifunctional Composite Materials

    Science.gov (United States)

    2013-03-01

    shielded fast neutrons more effectively than the other materials overall, but the sample with boron shielded ...the materials will shield against fast neutrons . 3.2 Assumptions With the information and specifications originally provided by the manufacturer on...to conduct fast foil activation experiments to determine the relative difference in the amount of neutrons shielded by the materials . This

  16. Thermal neutron shield and method of manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Metzger, Bert Clayton; Brindza, Paul Daniel

    2014-03-04

    A thermal neutron shield comprising boron shielding panels with a high percentage of the element Boron. The panel is least 46% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of boron shielding panels which includes enriching the pre-cursor mixture with varying grit sizes of Boron Carbide.

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

    Science.gov (United States)

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

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  19. Prevalence of enterobiasis and its incidence after blanket chemotherapy in a male orphanage.

    Science.gov (United States)

    Sirivichayakul, C; Pojjaroen-anant, C; Wisetsing, P; Lalitphiphat, A; Chanthavanich, P; Kabkaew, K

    2000-03-01

    A prospective observational study was conducted in a male orphanage to find out the prevalence of enterobiasis and its incidence after blanket chemotherapy using mebendazole. We found that the prevalence of enterobiasis was 28.9%. The incidence density of enterobiasis after blanket chemotherapy was 379.82 per 1,000 person-years which was quite high. We suggest that blanket chemotherapy should be repeated at every 6 months interval to control enterobiasis in orphanages.

  20. Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

    2015-10-15

    Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

  1. 76 FR 44903 - Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2011-07-27

    ...-000] Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on June 30, 2011 Kinder Morgan Interstate Gas Transmission, LLC (KMIGT), Post...

  2. 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.

  3. Magnetic shielding for superconducting RF cavities

    Science.gov (United States)

    Masuzawa, M.; Terashima, A.; Tsuchiya, K.; Ueki, R.

    2017-03-01

    Magnetic shielding is a key technology for superconducting radio frequency (RF) cavities. There are basically two approaches for shielding: (1) surround the cavity of interest with high permeability material and divert magnetic flux around it (passive shielding); and (2) create a magnetic field using coils that cancels the ambient magnetic field in the area of interest (active shielding). The choice of approach depends on the magnitude of the ambient magnetic field, residual magnetic field tolerance, shape of the magnetic shield, usage, cost, etc. However, passive shielding is more commonly used for superconducting RF cavities. The issue with passive shielding is that as the volume to be shielded increases, the size of the shielding material increases, thereby leading to cost increase. A recent trend is to place a magnetic shield in a cryogenic environment inside a cryostat, very close to the cavities, reducing the size and volume of the magnetic shield. In this case, the shielding effectiveness at cryogenic temperatures becomes important. We measured the permeabilities of various shielding materials at both room temperature and cryogenic temperature (4 K) and studied shielding degradation at that cryogenic temperature.

  4. Shielding structure analysis for LSDS facility

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization.

  5. Gamma Imaging using Rotational Modulation Collimation

    Science.gov (United States)

    2014-01-01

    in detection probabilities over traditional gamma detection, especially when the radiation source was shielded . Within DSTO the gamma imaging program...container Figure A2: Experimental setup used to measure radiation source stored in its lead and steel shielding transport container 16 UNCLASSIFIED...characterise the performance of two rotating modulation collimator (RMC) gamma imagers built by DSTO. The ability of these devices to image shielded and

  6. Inhibition of Frying Oil Oxidation by Carbon Dioxide Blanketing.

    Science.gov (United States)

    Totani, Nagao; Inoue, Ryota; Yawata, Miho

    2016-06-01

    The oxidation of oil starts, in general, from the penetration of atmospheric oxygen into oil. Inhibition of the vigorous oxidation of oil at deep-frying temperature under carbon dioxide flow, by disrupting the contact between oil and air, was first demonstrated using oil in a round bottom flask. Next, the minimum carbon dioxide flow rate necessary to blanket 4 L of frying oil in an electric fryer (surface area 690 cm(2)) installed with nonwoven fabric cover, was found to be 40 L/h. Then deep-frying of potato was done accordingly; immediately after deep-frying, an aluminum cover was placed on top of the nonwoven fabric cover to prevent the loss of carbon dioxide and the carbon dioxide flow was shut off. In conclusion, the oxidation of oil both at deep-frying temperature and during standing was remarkably inhibited by carbon dioxide blanketing at a practical flow rate and volume. Under the deep-frying conditions employed in this study, the increase in polar compound content was reduced to half of that of the control.

  7. 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).

  8. Effect of reactor size on the breeding economics of LMFBR blankets

    Energy Technology Data Exchange (ETDEWEB)

    Tagishi, A.; Driscoll, M.J.

    1975-02-01

    The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's.

  9. 77 FR 31004 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-05-24

    ...] 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 501, Birmingham, Alabama 35209, filed... Commission's regulations under the Natural Gas Act (NGA), and Southern's blanket certificate issued in...

  10. 48 CFR 313.303-5 - Purchases under blanket purchase agreements.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Purchases under blanket purchase agreements. 313.303-5 Section 313.303-5 Federal Acquisition Regulations System HEALTH AND HUMAN... Methods 313.303-5 Purchases under blanket purchase agreements. (e)(5) HHS personnel that sign...

  11. 76 FR 58488 - Dominion Cove Point LNG, LP; Application for Blanket Authorization to Export Previously Imported...

    Science.gov (United States)

    2011-09-21

    ... Dominion Cove Point LNG, LP; Application for Blanket Authorization to Export Previously Imported Liquefied... (Application), filed on August 8, 2011, by Dominion Cove Point LNG, LP (DCP), requesting blanket authorization to export liquefied natural gas (LNG) that previously had been imported into the United States...

  12. 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 250 billion cubic feet (Bcf) of foreign sourced liquefied natural gas (LNG) for a two-year...

  13. 78 FR 35263 - Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Previously...

    Science.gov (United States)

    2013-06-12

    ... Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Previously Imported... receipt of an application (Application), filed on April 19, 2013, by Freeport LNG Development, L.P. (Freeport LNG), requesting blanket authorization to export liquefied natural gas (LNG) that previously...

  14. 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... application (Application), filed on October 26, 2012, by Sempra LNG Marketing, LLC (Sempra LNG Marketing), requesting blanket authorization to export liquefied natural gas (LNG) that previously had been imported...

  15. 75 FR 38092 - The Dow Chemical Company; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-07-01

    ... Chemical Company; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY: Office of... The Dow Chemical Company (Dow), requesting blanket authorization to export liquefied natural gas (LNG... equivalent of 390 billion cubic feet (Bcf) of natural gas on a short-term or spot market basis. The LNG...

  16. 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... 500 Billion cubic feet (Bcf) of natural gas on a short-term or spot market basis. The LNG would...

  17. 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...

  18. 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 an...

  19. Reliability-Based Electronics Shielding Design Tools

    Science.gov (United States)

    Wilson, J. W.; O'Neill, P. J.; Zang, T. A.; Pandolf, J. E.; Tripathi, R. K.; Koontz, Steven L.; Boeder, P.; Reddell, B.; Pankop, C.

    2007-01-01

    Shielding design on large human-rated systems allows minimization of radiation impact on electronic systems. Shielding design tools require adequate methods for evaluation of design layouts, guiding qualification testing, and adequate follow-up on final design evaluation.

  20. Material Effectiveness for Radiation Shielding

    Science.gov (United States)

    2003-01-01

    Materials with a smaller mean atomic mass, such as lithium (Li) hydride and polyethylene, make the best radiation shields for astronauts. The materials have a higher density of nuclei and are better able to block incoming radiation. Also, they tend to produce fewer and less dangerous secondary particles after impact with incoming radiation.

  1. 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.

  2. Predictions for Radiation Shielding Materials

    Science.gov (United States)

    Kiefer, Richard L.

    2002-01-01

    Radiation from galactic cosmic rays (GCR) and solar particle events (SPE) is a serious hazard to humans and electronic instruments during space travel, particularly on prolonged missions outside the Earth s magnetic fields. Galactic cosmic radiation (GCR) is composed of approx. 98% nucleons and approx. 2% electrons and positrons. Although cosmic ray heavy ions are 1-2% of the fluence, these energetic heavy nuclei (HZE) contribute 50% of the long-term dose. These unusually high specific ionizations pose a significant health hazard acting as carcinogens and also causing microelectronics damage inside spacecraft and high-flying aircraft. These HZE ions are of concern for radiation protection and radiation shielding technology, because gross rearrangements and mutations and deletions in DNA are expected. Calculations have shown that HZE particles have a strong preference for interaction with light nuclei. The best shield for this radiation would be liquid hydrogen, which is totally impractical. For this reason, hydrogen-containing polymers make the most effective practical shields. Shielding is required during missions in Earth orbit and possibly for frequent flying at high altitude because of the broad GCR spectrum and during a passage into deep space and LunarMars habitation because of the protracted exposure encountered on a long space mission. An additional hazard comes from solar particle events (SPEs) which are mostly energetic protons that can produce heavy ion secondaries as well as neutrons in materials. These events occur at unpredictable times and can deliver a potentially lethal dose within several hours to an unshielded human. Radiation protection for humans requires safety in short-term missions and maintaining career exposure limits within acceptable levels on future long-term exploration missions. The selection of shield materials can alter the protection of humans by an order of magnitude. If improperly selected, shielding materials can actually

  3. WAVS radiation shielding references and assumptions

    Energy Technology Data Exchange (ETDEWEB)

    McLean, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-07

    At ITER, the confluence of a high radiation environment and the requirement for high performance imaging for plasma and plasma-facing surface diagnosis will necessitate extensive application of radiation shielding. Recommended here is a dual-layer shield design composed of lead for gamma attenuation, surrounded by a fire-resistant polyehtylene doped with a thermal neutron absorber for neutron shielding.

  4. 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.

  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. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Radiation shielding for diagnostic radiology.

    Science.gov (United States)

    Martin, Colin J

    2015-07-01

    Scattered radiation makes up the majority of the stray radiation field around an X-ray unit. The scatter is linked to the amount of radiation incident on the patient. It can be estimated from quantities used to assess patient dose such as the kerma-area product, and factors have been established linking this to levels of scattered radiation for radiography and fluoroscopy. In radiography shielding against primary radiation is also needed, but in other modalities this is negligible, as the beam is intercepted by the image receptor. In the same way scatter from CT can be quantified in terms of dose-length product, but because of higher radiation levels, exposure to tertiary scatter from ceilings needs to be considered. Transmission requirements are determined from comparisons between calculated radiation levels and agreed dose criteria, taking into account the occupancy of adjacent areas. Thicknesses of shielding material required can then be calculated from simple equations.

  18. 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.

  19. Light shield for solar concentrators

    Energy Technology Data Exchange (ETDEWEB)

    Plesniak, Adam P.; Martins, Guy L.

    2014-08-26

    A solar receiver unit including a housing defining a recess, a cell assembly received in the recess, the cell assembly including a solar cell, and a light shield received in the recess and including a body and at least two tabs, the body defining a window therein, the tabs extending outward from the body and being engaged with the recess, wherein the window is aligned with the solar cell.

  20. Axial Neutron Flux Evaluation in a Tokamak System: a Possible Transmutation Blanket Position for a Fusion-Fission Transmutation System

    Science.gov (United States)

    Velasquez, Carlos E.; de P. Barros, Graiciany; Pereira, Claubia; Fortini Veloso, Maria A.; Costa, Antonella L.

    2012-08-01

    A sub-critical advanced reactor based on Tokamak technology with a D-T fusion neutron source is an innovative type of nuclear system. Due to the large number of neutrons produced by fusion reactions, such a system could be useful in the transmutation process of transuranic elements (Pu and minor actinides (MAs)). However, to enhance the MA transmutation efficiency, it is necessary to have a large neutron wall loading (high neutron fluence) with a broad energy spectrum in the fast neutron energy region. Therefore, it is necessary to know and define the neutron fluence along the radial axis and its characteristics. In this work, the neutron flux and the interaction frequency along the radial axis are evaluated for various materials used to build the first wall. W alloy, beryllium, and the combination of both were studied, and the regions more suitable to transmutation were determined. The results demonstrated that the best zone in which to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements of W alloy/W alloy and W alloy/beryllium would be able to meet the requirements of the high fluence and hard spectrum that are needed for transuranic transmutation. The system was simulated using the MCNP code, data from the ITER Final Design Report, 2001, and the Fusion Evaluated Nuclear Data Library/MC-2.1 nuclear data library.

  1. 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...

  2. 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.

  3. The Feasibility of Multipole Electrostatic Radiation Shielding

    Science.gov (United States)

    Metzger, Philip T.; Lane, John E.; Youngquist, Robert C.

    2004-01-01

    Although passive shielding appears to be the only workable solution for galactic cosmic radiation (GCR), active shielding may play an important augmenting role to control the dose from solar particle events (SPEs). It has been noted that, to meet the guidelines of NCRP Report No. 98 through the six SPEs of 1989, a crew member would need roughly double the passive shielding that is necessary to control the GCR dose . This would dramatically increase spacecraft mass, and so it has been proposed that a small but more heavily shielded storm shelter may be used to protect the crew during SPEs. Since a gradual SPE may last 5 or more days, staying in a storm shelter may be psychologically and physiologically distressing to the crew. Storm shelters do not provide shielding for the spacecraft itself against the SPE radiation, and radiation damage to critical electronics may result in loss of mission and life. Single-event effects during the radiation storm may require quick crew response to maintain the integrity of the spacecraft, and confining the crew to a storm shelter prohibits their attending to the spacecraft at the precise time when that attention is needed the most. Active shielding cannot protect against GCR because the particle energies are too high. Although lower energy particles are easier to stop in a passive shield, such shielding is more satisfactory against GCR than against SPE radiation because of the tremendous difference in their initial fluences. Even a small fraction of the SPE fluence penetrating the passive shielding may result in an unacceptably high dose. Active shielding is more effective than passive shielding against SPE radiation because it offers 100% shielding effectiveness up to the cutoff energy, and significant shielding effectiveness beyond the cutoff as well.

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

    Directory of Open Access Journals (Sweden)

    D. L. Griscom

    2012-02-01

    Full Text Available The discovery nearly two decades ago of a 90 km-diameter impact crater below the lower Chesapeake Bay has gone unnoted by the general public because to date all published literature on the subject has described it as "buried". To the contrary, evidence is presented here that the so-called "upland deposits" that blanket ∼5000 km2 of the U.S. Middle-Atlantic Coastal Plain (M-ACP display morphologic, lithologic, and stratigraphic features consistent with their being ejecta from the 35.4 Ma Chesapeake Bay Impact Structure (CBIS and absolutely inconsistent with the prevailing belief that they are of fluvial origin. Specifically supporting impact origin are the facts that (i a 95 %-pure iron ore endemic to the upland deposits of southern Maryland, eastern Virginia, and the District of Columbia has previously been proven to be impactoclastic in origin, (ii this iron ore welds together a small percentage of well-rounded quartzite pebbles and cobbles of the upland deposits into brittle sheets interpretable as "spall plates" created in the interference-zone of the CBIS impact, (iii the predominantly non-welded upland gravels have long ago been shown to be size sorted with an extreme crater-centric gradient far too large to have been the work of rivers, but well explained as atmospheric size-sorted interference-zone ejecta, (iv new evidence is provided here that ~60 % of the non-welded quartzite pebbles and cobbles of the (lower lying gravel member of the upland deposits display planar fractures attributable to interference-zone tensile waves, (v the (overlying loam member of the upland deposits is attributable to base-surge-type deposition, (vi several exotic clasts found in a debris flow topographically below the upland deposits can only be explained as jetting-phase crater ejecta, and (vii an allogenic granite boulder found among the upland deposits is deduced to have been launched into space and sculpted by hypervelocity air friction

  5. Facilities, testing program and modeling needs for studying liquid metal magnetohydrodynamic flows in fusion blankets

    Energy Technology Data Exchange (ETDEWEB)

    Bühler, L., E-mail: leo.buehler@kit.edu [Karlsruhe Institute of Technology (KIT), Postfach 3640, 76021 Karlsruhe (Germany); Mistrangelo, C.; Konys, J. [Karlsruhe Institute of Technology (KIT), Postfach 3640, 76021 Karlsruhe (Germany); Bhattacharyay, R. [Institute for Plasma Research, Gandhinagar, Gujarat 382428 (India); Huang, Q. [Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS) (China); Obukhov, D. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA) (Russian Federation); Smolentsev, S. [University of California Los Angeles (UCLA) (United States); Utili, M. [ENEA C.R. Brasimone, Camugnano 40032 (Italy)

    2015-11-15

    Since many years, liquid metal flows for applications in fusion blankets have been investigated worldwide. A review is given about modeling requirements and existing experimental facilities for investigations of liquid metal related issues in blankets with the focus on magnetohydrodynamics (MHD). Most of the performed theoretical and experimental works were dedicated to fundamental aspects of MHD flows under very strong magnetic fields as they may occur in generic elements of fusion blankets like pipes, ducts, bends, expansions and contractions. Those experiments are required to progressively validate numerical tools with the purpose of obtaining codes capable to predict MHD flows at fusion relevant parameters in complex blanket geometries, taking into account electrical and thermal coupling between fluid and structural materials. Scaled mock-up experiments support the theoretical activities and help deriving engineering correlations for cases which cannot be calculated with required accuracy up to now.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

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

  7. Properties of Ejecta Blanket Deposits Surrounding Morasko Meteorite Impact Craters (Poland)

    Science.gov (United States)

    Szokaluk, M.; Muszyński, A.; Jagodziński, R.; Szczuciński, W.

    2016-08-01

    Morasko impact craters are a record of the fall of a meteorite into the soft sediments. The presented results illustrate the geological structure of the area around the crater as well as providing evidence of the occurrence of ejecta blanket.

  8. 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 ...

  9. 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.)

  10. 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

  11. 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

  12. 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.

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

    OpenAIRE

    N. Lukwa; A. Makuwaza; T. Chiwade; Mutambu, S L; 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...

  14. 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.

  15. Dique seco, en South Shields

    Directory of Open Access Journals (Sweden)

    Frank Stott, Peter

    1958-10-01

    Full Text Available La conocida empresa Brigham & Cowan Ltd, de South Shields (Inglaterra, acaba de construir un dique de carena en la desembocadura del río Tyne, destinado a la reparación de tanques y cargas de gran tonelaje y de relativamente poco calado. El vaso tiene 217 m de longitud, 29 de anchura mínima en la entrada, 6,40 de a l tura de agua sobre el umbral de entrada y una compuerta metálica rebatible hacia adelante. En este trabajo se describen las partes que mejor caracterizan esta importante obra.

  16. 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.

  17. Lithium hydride - A space age shielding material

    Science.gov (United States)

    Welch, F. H.

    1974-01-01

    Men and materials performing in the environment of an operating nuclear reactor require shielding from the escaping neutron particles and gamma rays. For efficient shielding from gamma rays, dense, high atomic number elements such as iron, lead, or tungsten are required, whereas light, low atomic number elements such as hydrogen, lithium, or beryllium are required for efficient neutron shielding. The use of lithium hydride (LiH) as a highly efficient neutron-shielding material is considered. It contains, combined into a single, stable compound, two of the elements most effective in attenuating and absorbing neutrons.

  18. 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.

  19. 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

  20. Artificial Dielectric Shields for Integrated Transmission Lines

    NARCIS (Netherlands)

    Ma, Y.; Rejaei, B.; Zhuang, Y.

    2008-01-01

    We present a novel shielding method for on-chip transmission lines built on conductive silicon substrates. The shield consists of an artificial dielectric with a very high in-plane dielectric constant, built from two patterned metal layers isolated by a very thin dielectric film. Inserted below an i

  1. Alignment modification for pencil eye shields

    Energy Technology Data Exchange (ETDEWEB)

    Evans, M.D.; Pla, M.; Podgorsak, E.B. (McGill Univ., Quebec (Canada))

    1989-01-01

    Accurate alignment of pencil beam eye shields to protect the lens of the eye may be made easier by means of a simple modification of existing apparatus. This involves drilling a small hole through the center of the shield to isolate the rayline directed to the lens and fabricating a suitable plug for this hole.

  2. Shielding for beta-gamma radiation.

    Science.gov (United States)

    Fletcher, J J

    1993-06-01

    The build-up factor, B, for lead was expressed as a polynominal cubic function of the relaxation length, mu x, and incorporated in a "general beta-gamma shielding equation." A computer program was written to determine shielding thickness for polyenergetic beta-gamma sources without resorting to the conventional "add-one-HVL" method.

  3. Thermal neutron shield and method of manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Brindza, Paul Daniel; Metzger, Bert Clayton

    2013-05-28

    A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

  4. 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.

  5. 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)

  6. Mars Exploration Rover Heat Shield Recontact Analysis

    Science.gov (United States)

    Raiszadeh, Behzad; Desai, Prasun N.; Michelltree, Robert

    2011-01-01

    The twin Mars Exploration Rover missions landed successfully on Mars surface in January of 2004. Both missions used a parachute system to slow the rover s descent rate from supersonic to subsonic speeds. Shortly after parachute deployment, the heat shield, which protected the rover during the hypersonic entry phase of the mission, was jettisoned using push-off springs. Mission designers were concerned about the heat shield recontacting the lander after separation, so a separation analysis was conducted to quantify risks. This analysis was used to choose a proper heat shield ballast mass to ensure successful separation with low probability of recontact. This paper presents the details of such an analysis, its assumptions, and the results. During both landings, the radar was able to lock on to the heat shield, measuring its distance, as it descended away from the lander. This data is presented and is used to validate the heat shield separation/recontact analysis.

  7. 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.

  8. Spacesuit Radiation Shield Design Methods

    Science.gov (United States)

    Wilson, John W.; Anderson, Brooke M.; Cucinotta, Francis A.; Ware, J.; Zeitlin, Cary J.

    2006-01-01

    Meeting radiation protection requirements during EVA is predominantly an operational issue with some potential considerations for temporary shelter. The issue of spacesuit shielding is mainly guided by the potential of accidental exposure when operational and temporary shelter considerations fail to maintain exposures within operational limits. In this case, very high exposure levels are possible which could result in observable health effects and even be life threatening. Under these assumptions, potential spacesuit radiation exposures have been studied using known historical solar particle events to gain insight on the usefulness of modification of spacesuit design in which the control of skin exposure is a critical design issue and reduction of blood forming organ exposure is desirable. Transition to a new spacesuit design including soft upper-torso and reconfigured life support hardware gives an opportunity to optimize the next generation spacesuit for reduced potential health effects during an accidental exposure.

  9. 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.

  10. Background simulations and shielding calculations

    Science.gov (United States)

    Kudryavtsev, Vitaly A.

    2011-04-01

    Key improvements in the sensitivity of the underground particle astrophysics experiments can only be achieved if the radiation causing background events in detectors is well understood and proper measures are taken to suppress it. The background radiation arising from radioactivity and cosmic-ray muons is discussed here together with the methods of its suppression. Different shielding designs are considered to attenuate gamma-rays and neutrons coming from radioactivity in rock and lab walls. Purity of materials used in detector construction is analysed and the background event rates due to the presence of radioactive isotopes in detector components are discussed. Event rates in detectors caused by muon-induced neutrons with and without active veto systems are presented leading to the requirements for the depth of an underground laboratory and the efficiency of the veto system.

  11. 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.

  12. The use of nipple shields: A review

    Directory of Open Access Journals (Sweden)

    Selina Chow

    2016-11-01

    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.

  13. On the morphometry of terrestrial shield volcanoes

    Science.gov (United States)

    Grosse, Pablo; Kervyn, Matthieu

    2016-04-01

    Shield volcanoes are described as low angle edifices that have convex up topographic profiles and are built primarily by the accumulation of lava flows. This generic view of shields' morphology is based on a limited number of monogenetic shields from Iceland and Mexico, and a small set of large oceanic islands (Hawaii, Galapagos). Here, the morphometry of over 150 monogenetic and polygenetic shield volcanoes, identified inthe Global Volcanism Network database, are analysed quantitatively from 90-meter resolution DEMs using the MORVOLC algorithm. An additional set of 20 volcanoes identified as stratovolcanoes but having low slopes and being dominantly built up by accumulation of lava flows are documented for comparison. Results show that there is a large variation in shield size (volumes range from 0.1 to >1000 km3), profile shape (height/basal width ratios range from 0.01 to 0.1), flank slope gradients, elongation and summit truncation. Correlation and principal component analysis of the obtained quantitative database enables to identify 4 key morphometric descriptors: size, steepness, plan shape and truncation. Using these descriptors through clustering analysis, a new classification scheme is proposed. It highlights the control of the magma feeding system - either central, along a linear structure, or spatially diffuse - on the resulting shield volcano morphology. Genetic relationships and evolutionary trends between contrasted morphological end-members can be highlighted within this new scheme. Additional findings are that the Galapagos-type morphology with a central deep caldera and steep upper flanks are characteristic of other shields. A series of large oceanic shields have slopes systematically much steeper than the low gradients (<4-8°) generally attributed to large Hawaiian-type shields. Finally, the continuum of morphologies from flat shields to steeper complex volcanic constructs considered as stratovolcanoes calls for a revision of this oversimplified

  14. 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 will...

  15. 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...

  16. 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.

  17. The European ITER Test Blanket Modules: Current status of fabrication technologies development and a way forward

    Energy Technology Data Exchange (ETDEWEB)

    Zmitko, Milan, E-mail: milan.zmitko@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain); Galabert, Jose [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain); Thomas, Noël [ATMOSTAT, F-94815 Villejuif (France); Forest, Laurent [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bucci, Philippe; Cogneau, Laurence [CEA-DRT, 38000 Grenoble (France); Rey, Jörg; Neuberger, Heiko [Karlsruhe Institute of Technology (KIT), Postfach 3640, Karlsruhe (Germany); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain)

    2015-10-15

    Highlights: • Significant progress on development of welding procedures for European TBM achieved. • Fabrication processes feasibility based on diffusion and fusion welding demonstrated. • TBM box assembly welding scenarios investigated and welding scenarios identified. • Future qualification of pF/WPS proposed through realization of a number of QMUs. - Abstract: The paper reviews fabrication technologies and procedures applied for manufacturing of the TBM sub-components, like, HCLL and HCPB cooling plates, HCLL/HCPB stiffening plates, and HCLL/HCPB first wall and side caps. The used technologies are based on fusion and diffusion welding techniques taking into account specificities of the EUROFER-97 steel. Development of a standardized procedure complying with professional codes and standards (RCC-MRx), a preliminary fabrication/welding procedure specification (pF/WPS), is described as well as a fabrication and characterization of feasibility mock-ups (FMU) aimed at assessing the suitability of a fabrication process for fulfilling the design and fabrication specifications. Also, fabrication procedures for the TBM box assembly are presently under development through collaboration between European Fusion Laboratories and Industry for the establishment of an optimized assembly sequence/scenario and development of standardized welding procedure specifications. Selection of optimized assembly scenario takes into accounts not only the design requirements and fabrication possibilities/constraints but also maximum accessibility to the welds for sound non-destructive examination in compliance with welds classification. A future approach towards qualification of the developed fabrication technologies and procedures, through a number of medium to full-size qualification mock-ups according to European standards, is outlined before construction of the first TBMs.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Reduction in stray radiation dose using a body-shielding device during external radiation therapy.

    Science.gov (United States)

    Zhang, Shuxu; Jiang, Shaohui; Zhang, Quanbin; Lin, Shengqu; Wang, Ruihao; Zhou, Xiang; Zhang, Guoqian; Lei, Huaiyu; Yu, Hui

    2017-03-01

    With the purpose of reducing stray radiation dose (SRD) in out-of-field region (OFR) during radiotherapy with 6 MV intensity-modulated radiation therapy (IMRT), a body-shielding device (BSD) was prepared according to the measurements obtained in experimental testing. In experimental testing, optimal shielding conditions, such as 1 mm lead, 2 mm lead, and 1 mm lead plus 10 mm bolus, were investigated along the medial axis of a phantom using thermoluminescent dosimeters (TLDs). The SRDs at distances from field edge were then measured and analyzed for a clinical IMRT treatment plan for nasopharyngeal carcinoma before and after shielding using the BSD. In addition, SRDs in anterior, posterior, left and right directions of phantom were investigated with and without shielding, respectively. Also, the SRD at the bottom of treatment couch was measured. SRD decreased exponentially to a constant value with increasing distance from field edge. The shielding rate was 50%-80%; however, there were no significant differences in SRDs when shielded by 1 mm lead, 2 mm lead, or 1 mm lead plus 10 mm bolus (P>0.05). Importantly, the 10 mm bolus absorbed back-scattering radiation due to the interaction between photons and lead. As a result, 1 mm lead plus 10 mm bolus was selected to prepare the BSD. After shielding with BSD, total SRDs in the OFR decreased to almost 50% of those without shielding when irradiated with IMRT beams. Due to the effects of treatment couch and gantry angle, SRDs at distances were not identical in anterior, posterior, left and right direction of phantom without BSD. As higher dose in anterior and lower dose in posterior, SRDs were substantial similarities after shielding. There was no significant difference in SRDs for left and right directions with or without shielding. Interestingly, SRDs in the four directions were similar after shielding. From these results, the BSD developed in this study may significantly reduce SRD in the OFR during

  3. 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.

  4. Optimization design of electromagnetic shielding composites

    Science.gov (United States)

    Qu, Zhaoming; Wang, Qingguo; Qin, Siliang; Hu, Xiaofeng

    2013-03-01

    The effective electromagnetic parameters physical model of composites and prediction formulas of composites' shielding effectiveness and reflectivity were derived based on micromechanics, variational principle and electromagnetic wave transmission theory. The multi-objective optimization design of multilayer composites was carried out using genetic algorithm. The optimized results indicate that material parameter proportioning of biggest absorption ability can be acquired under the condition of the minimum shielding effectiveness can be satisfied in certain frequency band. The validity of optimization design model was verified and the scheme has certain theoretical value and directive significance to the design of high efficiency shielding composites.

  5. 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.

  6. Development of Joining Technologies for the ITER Blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byoung Kwon; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan; Lee, Dong Won; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of)

    2011-01-15

    The design of the ITER blanket first wall includes the Beryllium amour tiles joined to CuCrZr heat sink with stainless steel cooling tubes. For the ITER application, the Be/CuCrZr/SS joint was proposed as a first wall material. The joining of Be/CuCrZr as well as CuCrZr/SS was generally carried out by using a hot isostatic pressing (CuC) in many countries. The joining strength for Be/CuCrZr is relatively lower than that for CuCrZr/SS, since we usually forms surface oxides (BeO) and brittle a metallics with Cu. Therefore, the joining technology for the Be/CuCrZr joint has been investigated. Be is apt to adsorb oxygen in an air atmosphere, so we should be etched to eliminate the surface pre-oxide using a chemical solution and Ar ions in a vacuum chamber. Then we is coated with a first was to prevent further oxidation. The kinds of a first we are chosen to be able to enhance the joining strength as inhibiting excessive be diffusion. The performance of the Be/CuCrZr/SS joint used for the ITER first wall is primarily dependent on the joining strength of the Be/CuCrZr interface. The Cr/Cu and Ti/Cr/Cu interlayers enabled the successful joining of be tile to CuCrZr plate. Moreover, ion-beam assisted deposition (IBAD) increased joining strength of the Be/CuCrZr joint mock-ups. IBAD induced the increased packing of depositing atoms, which resulted in denser and more adhesive interlayers. The interlayers formed by IBAD process revealed about 40% improved resistance to the scratch test. It is suggested that the improved adhesion of coating interlayers enabled tight joining of Be and CuCrZr blocks. As compared to without IBAD coating, the shear strength as well as the 4-point bend strength were increased more than 20% depending on interlayer types and coating conditions

  7. Numerical design of the Seed-Blanket Unit for the thorium nuclear fuel cycle

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the Monte Carlo modelling by the means of the Monte Carlo Continuous Energy Burn-up Code of the 17x17 Pressurized Water Reactor fuel assembly designed according to the Radkowsky Thorium Fuel concept. The design incorporates the UO2 seed fuel located in the centre and (Th,UO2 blanket fuel located in the peripheries of fuel assembly. The high power seed region supplies neutrons for the low power blanket region and thus induces breeding of fissile 233U from fertile 232Th. The both regions are physically separated and thus this approach is also known as either the heterogonous approach or the Seed-Blanket Unit. In the numerical analysis we consider the time evolutions of infinite neutron multiplication factor, axial/radial power density profile, 233U, 235U and 232Th.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-17

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

  9. 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.

  10. OPTIMAL BETA-RAY SHIELDING THICKNESSES FOR DIFFERENT THERAPEUTIC RADIONUCLIDES AND SHIELDING MATERIALS.

    Science.gov (United States)

    Cho, Yong In; Kim, Ja Mee; Kim, Jung Hoon

    2016-04-06

    To better understand the distribution of deposited energy of beta and gamma rays according to changes in shielding materials and thicknesses when radionuclides are used for therapeutic nuclear medicine, a simulation was conducted. The results showed that due to the physical characteristics of each therapeutic radionuclide, the thicknesses of shielding materials at which beta-ray shielding takes place varied. Additional analysis of the shielding of gamma ray was conducted for radionuclides that emit both beta and gamma rays simultaneously with results showing shielding effects proportional to the atomic number and density of the shielding materials. Also, analysis of bremsstrahlung emission after beta-ray interactions in the simulation revealed that the occurrence of bremsstrahlung was relatively lower than theoretically calculated and varied depending on different radionuclides.

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

    OpenAIRE

    Zhang, Guanheng

    2015-01-01

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the 200 Displacements per Atom (DPA) radiation damage constraint of presently verified cladding materials. The S&B core is designed to have an elongated seed (or “driver”) to maximize the fraction of neutrons that radially leak into the su...

  12. Study of thorium-uranium based molten salt blanket in a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhao Jing, E-mail: zhao_jing@mail.tsinghua.edu.cn [INET, Tsinghua University, Beijing 100084 (China); Yang Yongwei; Zhou Zhiwei [INET, Tsinghua University, Beijing 100084 (China)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer A molten salt blanket has been designed for the fusion-fission hybrid reactor. Black-Right-Pointing-Pointer The use of Thorium in the molten salt fuels has been studied. Black-Right-Pointing-Pointer The molten salt was consisted of F-Li-Be and with the thickness of 40 cm. Black-Right-Pointing-Pointer The concentration of {sup 6}Li was chosen to be the natural enrichment ratio. Black-Right-Pointing-Pointer The result shows that TBR is greater than 1, M is about 15-16. - Abstract: Not only solid fuels, but also liquid fuels can be used for the fusion-fission symbiotic reactor blanket. The operational record of the molten salt reactor with F-Li-Be was very successful, so the F-Li-Be blanket was chosen for research. The molten salt has several features which are suited for the fusion-fission applications. The fuel material uranium and thorium were dissolved in the F-Li-Be molten salt. A combined program, COUPLE, was used for neutronics analysis of the molten salt blanket. Several cases have been calculated and compared. Not only the influence of the different fuels have been studied, but also the thickness of the molten salt, and the concentration of the {sup 6}Li in the molten salt. Preliminary studies indicate that when thorium-uranium-plutonium fuels were added into a F-Li-Be molten salt blanket and with a component of 71% LiF-2% BeF{sub 2}-13.5% ThF{sub 4}-8.5% UF{sub 4}-5% PuF{sub 3}, and also with the molten salt thickness of 40 cm and natural concentration of {sup 6}Li, the appropriate blanket energy multiplication factor and TBR can be obtained. The result shows that thorium-uranium molten salt can be used in the blanket of a fusion-fission symbiotic reactor. The research on the molten salt blanket must be valuable for the design of fusion-fission symbiotic reactor.

  13. 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.

  14. 76 FR 2093 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2011-01-12

    ... Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY... November 30, 2010, by Eni USA Gas Marketing LLC (Eni USA), requesting blanket authorization to export... and Gas Global Security and Supply, Office of Fossil Energy, Forrestal Building, Room 3E-042,...

  15. 76 FR 33746 - Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Liquefied Natural...

    Science.gov (United States)

    2011-06-09

    ... Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Liquefied Natural Gas..., 2011, by Freeport LNG Development, L.P. (Freeport LNG), requesting blanket authorization to export liquefied natural gas (LNG) that previously had been imported into the United States from foreign sources...

  16. 75 FR 13755 - Freeport LNG Development, L.P.; Application To Amend Blanket Authorization To Export Liquefied...

    Science.gov (United States)

    2010-03-23

    ... Freeport LNG Development, L.P.; Application To Amend Blanket Authorization To Export Liquefied Natural Gas... application filed on March 4, 2010, by Freeport LNG Development, L.P. (Freeport LNG), requesting an amendment to its blanket authorization to export liquefied natural gas (LNG) granted by DOE/FE on May 28,...

  17. 75 FR 62510 - Chevron U.S.A. Inc.; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-10-12

    ... U.S.A. Inc.; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY: Office of..., 2010, by Chevron U.S.A. Inc. (Chevron), requesting blanket authorization to export liquefied natural... up to the equivalent of 72 billion cubic feet (Bcf) of natural gas on a short-term or spot...

  18. 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 magnets...

  19. Boron-10 loaded inorganic shielding material

    Science.gov (United States)

    Baker, S. I.; Ryskiewicz, R. S.

    1972-01-01

    Shielding material containing Boron 10 and gadoliunium for neutron absorption has been developed to reduce interference from low energy neutrons in measurement of fission neutron spectrum using Li-6 fast neutron spectrometer.

  20. 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...

  1. Shielded ADR Magnets For Space Applications Project

    Data.gov (United States)

    National Aeronautics and Space Administration — An important consideration of the use of superconducting magnets in ADR applications is shielding of the other instruments in the vicinity of the superconducting...

  2. 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...

  3. 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...

  4. A Shielding Model for an Inflatable Vehicle, TransHab, and the Associated Astronaut Space Radiation Risk Assessment

    Science.gov (United States)

    Atwell, William; Badhwar, Gautam

    2000-01-01

    TransHab, a habitable inflatable structure, has been proposed as a possible module for the International Space Station that provides significant increase in the available volume compared with the US Hab module and fo r a human Mars mission . A study was undertaken to understand and provide design inputs for crew radiation exposures. The results show that the current design provides sufficient shielding to assure that the crew exposures are below the crew exposure limits currently adopted for the ISS. In addition, the shielding provides adequate protection from the largest solar particle events (SPEs) observed during the last 40 years.

  5. 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.

  6. Enhanced radiation shielding with galena concrete

    OpenAIRE

    Hadad Kamal; Majidi Hosein; Sarshough Samira

    2015-01-01

    A new concrete, containing galena mineral, with enhanced shielding properties for gamma sources is developed. To achieve optimized shielding properties, ten types of galena concrete containing different mixing ratios and a reference normal concrete of 2300 kg/m3 density are studied experimentally and numerically using Monte Carlo and XCOM codes. For building galena concrete, in addition to the main composition, micro-silica and water, galena mineral (contai...

  7. 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

  8. 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.

  9. Reliability Methods for Shield Design Process

    Science.gov (United States)

    Tripathi, R. K.; Wilson, J. W.

    2002-01-01

    Providing protection against the hazards of space radiation is a major challenge to the exploration and development of space. The great cost of added radiation shielding is a potential limiting factor in deep space operations. In this enabling technology, we have developed methods for optimized shield design over multi-segmented missions involving multiple work and living areas in the transport and duty phase of space missions. The total shield mass over all pieces of equipment and habitats is optimized subject to career dose and dose rate constraints. An important component of this technology is the estimation of two most commonly identified uncertainties in radiation shield design, the shielding properties of materials used and the understanding of the biological response of the astronaut to the radiation leaking through the materials into the living space. The largest uncertainty, of course, is in the biological response to especially high charge and energy (HZE) ions of the galactic cosmic rays. These uncertainties are blended with the optimization design procedure to formulate reliability-based methods for shield design processes. The details of the methods will be discussed.

  10. 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.

  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. 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.

  13. Micromagnetic modeling of the shielding properties of nanoscale ferromagnetic layers

    Science.gov (United States)

    Iskandarova, I. M.; Knizhnik, A. A.; Popkov, A. F.; Potapkin, B. V.; Stainer, Q.; Lombard, L.; Mackay, K.

    2016-09-01

    Ferromagnetic shields are widely used to concentrate magnetic fields in a target region of space. Such shields are also used in spintronic nanodevices such as magnetic random access memory and magnetic logic devices. However, the shielding properties of nanostructured shields can differ considerably from those of macroscopic samples. In this work, we investigate the shielding properties of nanostructured NiFe layers around a current line using a finite element micromagnetic model. We find that thin ferromagnetic layers demonstrate saturation of magnetization under an external magnetic field, which reduces the shielding efficiency. Moreover, we show that the shielding properties of nanoscale ferromagnetic layers strongly depend on the uniformity of the layer thickness. Magnetic anisotropy in ultrathin ferromagnetic layers can also influence their shielding efficiency. In addition, we show that domain walls in nanoscale ferromagnetic shields can induce large increases and decreases in the generated magnetic field. Therefore, ferromagnetic shields for spintronic nanodevices require careful design and precise fabrication.

  14. 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

  15. Computational thermo-fluid exploratory design analysis for complex ITER first wall/shield components

    Energy Technology Data Exchange (ETDEWEB)

    Youchison, Dennis L. [Sandia National Laboratories, Albuquerque, NM 87185 (United States)], E-mail: dlyouch@sandia.gov; Natoni, Greg [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Narula, Manmeet; Ying, Alice [University of California, Los Angeles, CA 90095 (United States)

    2008-12-15

    Engineers in the ITER US Party Team used several computational fluid dynamics codes to evaluate design concepts for the ITER first wall panels and the neutron shield modules. The CFdesign code enabled them to perform design studies of modules 7 and 13 very efficiently. CFdesign provides a direct interface to the CAD program, CATIA v5. The geometry input and meshing are greatly simplified. CFdesign is a finite element code, rather than a finite volume code. Flow experiments and finite volume calculations from SC-Tetra, Fluent and CFD2000 verified the CFdesign results. Several new enhancements allow CFdesign to export temperatures, pressures and convective heat transfer coefficients to other finite element models for further analysis. For example, these loads and boundary conditions directly feed into codes such as ABAQUS to perform stress analysis. In this article, we review the use of 2- and 4-mm flow driver gaps in the shield modules and the use of 1-mm gaps along the tee-vane in the front water header to obtain a good flow distribution in both the first wall and shield modules for 7 and 13. Plasma heat flux as well as neutron heating derived from MCNP calculations is included in the first wall and shield module analyses. We reveal the non-uniformity of the convective heat transfer coefficient inside complex 3D geometries exposed to a one-sided heat flux and non-uniform volumetric heating. Most models consisted of 3-4 million tetrahedron elements. We obtained temperature and velocity distributions, as well as pressure drop information, for models of nearly exact geometry compared to the CATIA fabrication models. We also describe the coupling to thermal stress analysis in ABAQUS. The results presented provide confidence that the preliminary design of these plasma facing components will meet ITER requirements.

  16. Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

    Science.gov (United States)

    Zhou, Z.; Yang, Y.; Xu, H.

    2011-10-01

    This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

  17. Non-LTE Line Blanketing in Stars With Extended Outflowing Atmospheres.

    Science.gov (United States)

    Hillier, D. J.; Miller, D. L.

    1995-05-01

    With continuing advances in radiative transfer techniques, increases in computing power, and the availability of at least some of the necessary atomic data, it is now possible to consider the computation of detailed non-LTE model atmospheres in which the full effects of non-LTE line blanketing are taken into account. We discuss our own implementation of non-LTE line blanketing in a spherical non-LTE code developed for the investigation of objects with extended outflows. A partial linearization technique is used to simultaneously solve the radiative transfer equation in conjunction with the equations of statistical equilibrium. Convergence properties are similar to that obtained with an ``Optimal'' Approximate-Lambda Operator. CNO line blanketing has been incorporated without major difficulty, while Fe blanketing is currently being installed. Comparisons of model spectra with recent HST observations of an LMC WC star will be presented. When completed we anticipate the code will be applicable to the study of a wide range of phenomena exhibiting outflows including Luminous-Blue variables, Supernovae, Wold-Rayet stars and Novae. Partial support for this work was provided by NASA through grant Nos GO-5460.01-93A and GO-4550.01-92A from the Space Science Institute which is operated under the Association of Universities for Research in Astronomy, Inc., under NASA contract NAS5-26555. Support from NASA award NAGW-3828 is also gratefully acknowledged.

  18. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Science.gov (United States)

    2010-04-01

    ... granted a blanket certificate of public convenience and necessity pursuant to section 7 of the Natural Gas... the sales customer to arrange for any pipeline-provided service necessary to deliver gas to the customer. (e) Small customer cost-based rate. A pipeline that provided bundled sales service to a...

  19. 78 FR 44558 - Stingray Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-07-24

    ... Energy Regulatory Commission Stingray Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization Take notice that on July 3, 2013, Stingray Pipeline Company, L.L.C. (Stingray), 1100 Louisiana... in the federal waters offshore Louisiana. Specifically, Stingray proposes to abandon, by sale, its...

  20. 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...

  1. 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...

  2. 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...

  3. Acoustic contributions of a sound absorbing blanket placed in a double panel structure: absorption versus transmission.

    Science.gov (United States)

    Doutres, Olivier; Atalla, Noureddine

    2010-08-01

    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 absorption behavior can affect the performance of the double panel structure.

  4. 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)

  5. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Science.gov (United States)

    2010-07-01

    .... 76-07. (h) Routine Use—Disclosure to the Office of Personnel Management. A record from a system of... Personnel Management (OPM) concerning information on pay and leave, benefits, retirement deduction, and any... Blanket Routine Uses (a) Routine Use—Law Enforcement. If a system of records maintained by a DoD...

  6. 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...-free, (866) 208-3676 or TTY, (202) 502-8659. Specifically, Sabine proposes to abandon, in place,...

  7. 77 FR 53874 - The Dow Chemical Company; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-09-04

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY The Dow... application (Application), filed on July 13, 2012, by The Dow Chemical Company (Dow), requesting blanket... on a short-term or spot market basis for a two-year period commencing on October 5, 2012.\\1\\...

  8. Salted lamb meat blanket of Petrolina-Pernambuco, Brazil: process and quality

    Directory of Open Access Journals (Sweden)

    Nely de Almeida Pedrosa

    2014-03-01

    Full Text Available Salted lamb meat blanket, originated from boning, salting, and drying of whole lamb carcass, was studied aiming at obtaining information that support the search for guarantees of origin for this typical regional product from the city of Petrolina-Pernambuco-Brazil. Data from three processing units were obtained, where it was observed the use of a traditional local technology that uses salting, an ancient preservation method; however, with a peculiar boning technique, resulting in a meat product with great potential for exploitation in the form of meat blanket. Based on the values of pH (6.22 ± 0.22, water activity (0.97 ± 0.02, and moisture (69.86 ± 2.26 lamb meat blanket is considered a perishable product, and consequently it requires the use of other preservation methods combined with salt, which along with the results of the microbiological analyses (absence of Salmonella sp, score <10 MPN/g of halophilic bacteria, total coliforms between 6.7 × 10³ and 5.2 × 10(6 FUC/g, and Staphylococcus from 8.1 × 10³ CFU/g at uncountable reinforce the need of hygienic practices to ensure product safety. These results, together with the product notoriety and the organization of the sector are important factors in achieving Geographical Indication of the Salted lamb Meat blanket of Petrolina.

  9. 78 FR 51182 - Sea Robin Pipeline Company, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-08-20

    ... Energy Regulatory Commission Sea Robin Pipeline Company, LLC; Notice of Request Under Blanket Authorization Take notice that on July 31, 2013, Sea Robin Pipeline Company, LLC (Sea Robin), P. O. Box 4967....205(b) and 157.216 of the Commission's Regulations under the Natural Gas Act (NGA), and Sea...

  10. 76 FR 31326 - Gulf LNG Pipeline, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2011-05-31

    ... Energy Regulatory Commission Gulf LNG Pipeline, LLC; Notice of Request Under Blanket Authorization Take notice that on May 18, 2011, Gulf LNG Pipeline, LLC (GLNG Pipeline), Colonial Brookwood Center, 569... to Margaret G. Coffman, Counsel, Gulf LNG Pipeline Company, LLC, Colonial Brookwood Center,...

  11. 75 FR 8327 - Golden Pass Pipeline LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-02-24

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Golden Pass Pipeline LLC; Notice of Request Under Blanket Authorization February 17, 2010. Take notice that on October 29, 2009, Golden Pass Pipeline, LLC (GPPL), filed in...

  12. 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,...

  13. 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,...

  14. 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,...

  15. 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,...

  16. 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,...

  17. 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...

  18. 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...

  19. 75 FR 17708 - Kinder Morgan Louisiana Pipeline LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-04-07

    ... Energy Regulatory Commission Kinder Morgan Louisiana Pipeline LLC; Notice of Request Under Blanket Authorization March 30, 2010. Take notice that on March 25, 2010, Kinder Morgan Louisiana Pipeline LLC (KMLP... directed to Norman Watson, Director, Business Development, Kinder Morgan Louisiana Pipeline LLC, 500...

  20. 75 FR 35019 - Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-06-21

    ... Energy Regulatory Commission Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization June 11, 2010. Take notice that on June 3, 2009, Kinder Morgan Interstate Gas..., Kinder Morgan Interstate Gas Transmission LLC, P.O. Box 281304, Lakewood, Colorado 80228-8304, or...

  1. 75 FR 53966 - Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-09-02

    ... Federal Energy Regulatory Commission Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization August 27, 2010. Take notice that on August 25, 2010, Kinder Morgan Interstate Gas Transmission, LLC (Kinder Morgan), 370 Van Gordon Street, Lakewood, Colorado 80228-8304...

  2. 75 FR 45111 - Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-08-02

    ... Energy Regulatory Commission Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization July 26, 2010. Take notice that on July 20, 2010, Kinder Morgan Interstate Gas..., Vice President, Regulatory, Kinder Morgan Interstate Gas Transmission LLC, 370 Van Gordon...

  3. 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

  4. 77 FR 25711 - Cheniere Marketing, LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-05-01

    ... Cheniere Marketing, LLC; Application for Blanket Authorization To Export Previously Imported Liquefied... to export * * *'' \\9\\ \\6\\ Cheniere Marketing, LLC, DOE/FE Order No 2795 at 11. \\7\\ See Dominion Cove... application (Application), filed on March 30, 2012, by Cheniere Marketing, LLC (CMI), requesting...

  5. Targeted and shielded adenovectors for cancer therapy.

    Science.gov (United States)

    Hedley, Susan J; Chen, Jian; Mountz, John D; Li, Jing; Curiel, David T; Korokhov, Nikolay; Kovesdi, Imre

    2006-11-01

    Conditionally replicative adenovirus (CRAd) vectors are novel vectors with utility as virotherapy agents for alternative cancer therapies. These vectors have already established a broad safety record in humans and overcome some of the limitations of non-replicative adenovirus (Ad) vectors. In addition, one potential problem with these vectors, attainment of tumor or tissue selectivity has widely been addressed. However, two confounding problems limiting efficacy of these drug candidates remains. The paucity of the native Ad receptor on tumor tissues, and host humoral response due to pre-existing titers of neutralizing antibodies against the vector itself in humans have been highlighted in the clinical context. The well-characterized CRAd, AdDelta24-RGD, is infectivity enhanced, thus overcoming the lack of coxsackievirus and adenovirus receptor (CAR), and this agent is already rapidly progressing towards clinical translation. However, the perceived host humoral response potentially will limit gains seen from the infectivity enhancement and therefore a strategy to blunt immunity against the vector is required. On the basis of this caveat a novel strategy, termed shielding, has been developed in which the genetic modification of a virion capsid protein would provide uniformly shielded Ad vectors. The identification of the pIX capsid protein as an ideal locale for genetic incorporation of shielding ligands to conceal the Ad vector from pre-existing neutralizing antibodies is a major progression in the development of shielded CRAds. Preliminary data utilizing an Ad vector with HSV-TK fused to the pIX protein indicates that a shield against neutralizing antibodies can be achieved. The utility of various proteins as shielding molecules is currently being addressed. The creation of AdDelta24S-RGD, an infectivity enhanced and shielded Ad vector will provide the next step in the development of clinically and commercially feasible CRAds that can be dosed multiple times for

  6. 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.

  7. Equalization characteristics of an upflow sludge blanket-aerated biofilter (USB-AF) system.

    Science.gov (United States)

    Jun, H B; Park, S M; Park, J K; Lee, S H

    2005-01-01

    Equalization characteristics of the upflow sludge blanket-aerated bio-filter (USB-AF) were investigated with the fluctuated raw domestic sewage. Recycle of nitrified effluent from AF to USB triggered the equalization characteristics of the sludge blanket on both soluble and particulate organic matter. Increment of EPS in sludge blanket by nitrate recycle was detected and removal of turbidity and particulates increased at higher recycle ratios by bio-flocculation. Increased TCOD removal in the USB was due to both denitrification of recycled nitrate and entrapment of the particulate organic matter in sludge blanket. Capture of both soluble and particulate organic matter increased sludge blanket layer in the USB, which improved the reactor performances and reduced the organic load on the subsequent AF. Overall TCOD and SS removal efficiencies were about 98% and 96%, respectively in the USB-AF system. Turbidity in the USB effluent was about 44, 20 and 5.5 NTU, at recycle ratios of 0, 100 and 200%, respectively. Particle counts in the range 2-4 microm in the USB effluent were higher than those in influent without nitrate recycle, while particle counts in the range of 0.5-15 microm in the USB effluent decreased 70% at recycle ratio of 200%. The major constituent of EPS extracted from anaerobic sludge was protein and total EPS increased from 109.1 to 165.7 mg/g-VSS with nitrate recycle of 100%. Removal efficiency and concentration of T-N in the UBS-AF effluent was over 70% and below 16 mg/L, respectively.

  8. Electromagnetic shielding mats: facts and fiction.

    Science.gov (United States)

    Leitgeb, N; Cech, R

    2007-01-01

    The use of electricity is accompanied by electric and magnetic fields which, intended or not, became a part of our environment. However, fear from environmental electromagnetic fields (EMFs) is widespread and so is business with fear. A number of more or less serious products including miracle products are placed on the market partly at excessive costs. By numerical simulation the efficiency of electromagnetic shielding mats was investigated and claims of manufacturers and their cited expert opinions checked. It could be shown that such products do not fulfil the justified expectations of customers, neither in the extremely low frequency (ELF) nor in the radiofrequency (RF) range. On the contrary, these mats usually make things even worse. The connection to ground, if available, might increase the belief on shielding efficiency, but in fact it even enhances fields instead of improving shielding. The electric conductivity of the mat material plays a minor role in the ELF range and enhances field increase in the RF range. It can not explain the enormous price differences. It could be shown that positive reports can be explained by result picking and exceptional arrangements of selected field sources. Overall, the investigation showed that manufacturer's claims about the shielding effectiveness are misleading and fool the customers about the real situation. Therefore, acquisition and use of electromagnetic shielding mats must be strongly discouraged.

  9. Cosmic Ray Interactions in Shielding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Ankney, Austin S.; Orrell, John L.; Berguson, Timothy J.; Troy, Meredith D.

    2011-09-08

    This document provides a detailed study of materials used to shield against the hadronic particles from cosmic ray showers at Earth’s surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during transport for the MAJORANA collaboration. The materials suitable for cosmic-ray shield design are materials such as lead and iron that will stop the primary protons, and materials like polyethylene, borated polyethylene, concrete and water that will stop the induced neutrons. The interaction of the different cosmic-ray components at ground level (protons, neutrons, muons) with their wide energy range (from kilo-electron volts to giga-electron volts) is a complex calculation. Monte Carlo calculations have proven to be a suitable tool for the simulation of nucleon transport, including hadron interactions and radioactive isotope production. The industry standard Monte Carlo simulation tool, Geant4, was used for this study. The result of this study is the assertion that activation at Earth’s surface is a result of the neutronic and protonic components of the cosmic-ray shower. The best material to shield against these cosmic-ray components is iron, which has the best combination of primary shielding and minimal secondary neutron production.

  10. Advances in space radiation shielding codes

    Science.gov (United States)

    Wilson, John W.; Tripathi, Ram K.; Qualls, Garry D.; Cucinotta, Francis A.; Prael, Richard E.; Norbury, John W.; Heinbockel, John H.; Tweed, John; De Angelis, Giovanni

    2002-01-01

    Early space radiation shield code development relied on Monte Carlo methods and made important contributions to the space program. Monte Carlo methods have resorted to restricted one-dimensional problems leading to imperfect representation of appropriate boundary conditions. Even so, intensive computational requirements resulted and shield evaluation was made near the end of the design process. Resolving shielding issues usually had a negative impact on the design. Improved spacecraft shield design requires early entry of radiation constraints into the design process to maximize performance and minimize costs. As a result, we have been investigating high-speed computational procedures to allow shield analysis from the preliminary concept to the final design. For the last few decades, we have pursued deterministic solutions of the Boltzmann equation allowing field mapping within the International Space Station (ISS) in tens of minutes using standard Finite Element Method (FEM) geometry common to engineering design methods. A single ray trace in such geometry requires 14 milliseconds and limits application of Monte Carlo methods to such engineering models. A potential means of improving the Monte Carlo efficiency in coupling to spacecraft geometry is given.

  11. Asymmetric Electrostatic Radiation Shielding for Spacecraft

    Science.gov (United States)

    Metzger, Philip T.; Youngquist, Robert C.; Lane, John E.

    2005-01-01

    A paper describes the types, sources, and adverse effects of energetic-particle radiation in interplanetary space, and explores a concept of using asymmetric electrostatic shielding to reduce the amount of such radiation impinging on spacecraft. Typically, such shielding would include a system of multiple inflatable, electrically conductive spheres deployed in clusters in the vicinity of a spacecraft on lightweight structures that would maintain the spheres in a predetermined multipole geometry. High-voltage generators would maintain the spheres at potential differences chosen in conjunction with the multipole geometry so that the resulting multipole field would gradually divert approaching energetic atomic nuclei from a central region occupied by the spacecraft. The spheres nearest the center would be the most positive, so as to repel the positively charged impinging nuclei from the center. At the same time, the monopole potential of the overall spacecraft-and-shielding system would be made negative so as to repel thermal electrons. The paper presents results of computational simulations of energetic-particle trajectories and shield efficiency for a trial system of 21 spheres arranged in three clusters in an overall linear quadrupole configuration. Further development would be necessary to make this shielding concept practical.

  12. Correlated Uncertainties in Radiation Shielding Effectiveness

    Science.gov (United States)

    Werneth, Charles M.; Maung, Khin Maung; Blattnig, Steve R.; Clowdsley, Martha S.; Townsend, Lawrence W.

    2013-01-01

    The space radiation environment is composed of energetic particles which can deliver harmful doses of radiation that may lead to acute radiation sickness, cancer, and even death for insufficiently shielded crew members. Spacecraft shielding must provide structural integrity and minimize the risk associated with radiation exposure. The risk of radiation exposure induced death (REID) is a measure of the risk of dying from cancer induced by radiation exposure. Uncertainties in the risk projection model, quality factor, and spectral fluence are folded into the calculation of the REID by sampling from probability distribution functions. Consequently, determining optimal shielding materials that reduce the REID in a statistically significant manner has been found to be difficult. In this work, the difference of the REID distributions for different materials is used to study the effect of composition on shielding effectiveness. It is shown that the use of correlated uncertainties allows for the determination of statistically significant differences between materials despite the large uncertainties in the quality factor. This is in contrast to previous methods where uncertainties have been generally treated as uncorrelated. It is concluded that the use of correlated quality factor uncertainties greatly reduces the uncertainty in the assessment of shielding effectiveness for the mitigation of radiation exposure.

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

    Science.gov (United States)

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

    2014-04-01

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

  14. 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...

  15. Carbohydrate based materials for gamma radiation shielding

    Science.gov (United States)

    Tabbakh, F.; Babaee, V.; Naghsh-Nezhad, Z.

    2015-05-01

    Due to the limitation in using lead as a shielding material for its toxic properties and limitation in abundance, price or non-flexibility of other commonly used materials, finding new shielding materials and compounds is strongly required. In this conceptual study carbohydrate based compounds were considered as new shielding materials. The simulation of radiation attenuation is performed using MCNP and Geant4 with a good agreement in the results. It is found that, the thickness of 2 mm of the proposed compound may reduce up to 5% and 50% of 1 MeV and 35 keV gamma-rays respectively in comparison with 15% and 100% for the same thickness of lead.

  16. Electronics Shielding and Reliability Design Tools

    Science.gov (United States)

    Wilson, John W.; ONeill, P. M.; Zang, Thomas A., Jr.; Pandolf, John E.; Koontz, Steven L.; Boeder, P.; Reddell, B.; Pankop, C.

    2006-01-01

    It is well known that electronics placement in large-scale human-rated systems provides opportunity to optimize electronics shielding through materials choice and geometric arrangement. For example, several hundred single event upsets (SEUs) occur within the Shuttle avionic computers during a typical mission. An order of magnitude larger SEU rate would occur without careful placement in the Shuttle design. These results used basic physics models (linear energy transfer (LET), track structure, Auger recombination) combined with limited SEU cross section measurements allowing accurate evaluation of target fragment contributions to Shuttle avionics memory upsets. Electronics shielding design on human-rated systems provides opportunity to minimize radiation impact on critical and non-critical electronic systems. Implementation of shielding design tools requires adequate methods for evaluation of design layouts, guiding qualification testing, and an adequate follow-up on final design evaluation including results from a systems/device testing program tailored to meet design requirements.

  17. 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.

  18. Self-Shielding Of Transmission Lines

    Energy Technology Data Exchange (ETDEWEB)

    Christodoulou, Christos [Univ. of New Mexico, Albuquerque, NM (United States)

    2017-03-01

    The use of shielding to contend with noise or harmful EMI/EMR energy is not a new concept. An inevitable trade that must be made for shielding is physical space and weight. Space was often not as much of a painful design trade in older larger systems as they are in today’s smaller systems. Today we are packing in an exponentially growing number of functionality within the same or smaller volumes. As systems become smaller and space within systems become more restricted, the implementation of shielding becomes more problematic. Often, space that was used to design a more mechanically robust component must be used for shielding. As the system gets smaller and space is at more of a premium, the trades starts to result in defects, designs with inadequate margin in other performance areas, and designs that are sensitive to manufacturing variability. With these challenges in mind, it would be ideal to maximize attenuation of harmful fields as they inevitably couple onto transmission lines without the use of traditional shielding. Dr. Tom Van Doren proposed a design concept for transmission lines to a class of engineers while visiting New Mexico. This design concept works by maximizing Electric field (E) and Magnetic Field (H) field containment between operating transmission lines to achieve what he called “Self-Shielding”. By making the geometric centroid of the outgoing current coincident with the return current, maximum field containment is achieved. The reciprocal should be true as well, resulting in greater attenuation of incident fields. Figure’s 1(a)-1(b) are examples of designs where the current centroids are coincident. Coax cables are good examples of transmission lines with co-located centroids but they demonstrate excellent field attenuation for other reasons and can’t be used to test this design concept. Figure 1(b) is a flex circuit design that demonstrate the implementation of self-shielding vs a standard conductor layout.

  19. 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.

  20. 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.

  1. Exploring climatic controls on blanket bog litter decomposition across an altitudinal gradient

    Science.gov (United States)

    Bell, Michael; Ritson, Jonathan P.; Clark, Joanna M.; Verhoef, Anne; Brazier, Richard E.

    2016-04-01

    The hydrological and ecological functioning of blanket bogs is strongly coupled, involving multiple ecohydrological feedbacks which can affect carbon cycling. Cool and wet conditions inhibit decomposition, and favour the growth of Sphagnum mosses which produce highly recalcitrant litter. A small but persistent imbalance between production and decomposition has led to blanket bogs in the UK accumulating large amounts of carbon. Additionally, healthy bogs provide a suite of other ecosystems services including water regulation and drinking water provision. However, there is concern that climate change could increase rates of litter decomposition and disrupt this carbon sink. Furthermore, it has been argued that the response of these ecosystems in the warmer south west and west of the UK may provide an early analogue for later changes in the more extensive northern peatlands. In order to investigate the effects of climate change on blanket bog litter decomposition, we set-up a litter bag experiment across an altitudinal gradient spanning 200 m of elevation (including a transition from moorland to healthy blanket bog) on Dartmoor, an area of hitherto unstudied, climatically marginal blanket bog in the south west of the UK. At seven sites, water table depth and soil and surface temperature were recorded continuously. Litter bags filled with the litter of three vegetation species dominant on Dartmoor were incubated just below the bog surface and retrieved over a period of 12 months. We found significant differences in the rate of decomposition between species. At all sites, decomposition progressed in the order Calluna vulgaris (dwarf shrub) > Molinia caerulea (graminoid) > Sphagnum (bryophyte). However, while soil temperature did decrease along the altitudinal gradient, being warmer in the lower altitudes, a hypothesised accompanying decrease in decomposition rates did not occur. This could be explained by greater N deposition at the higher elevation sites (estimated

  2. RadShield: semiautomated shielding design using a floor plan driven graphical user interface.

    Science.gov (United States)

    DeLorenzo, Matthew C; Wu, Dee H; Yang, Kai; Rutel, Isaac B

    2016-09-01

    The purpose of this study was to introduce and describe the development of RadShield, a Java-based graphical user interface (GUI), which provides a base design that uniquely performs thorough, spatially distributed calculations at many points and reports the maximum air-kerma rate and barrier thickness for each barrier pursuant to NCRP Report 147 methodology. Semiautomated shielding design calculations are validated by two approaches: a geometry-based approach and a manual approach. A series of geometry-based equations were derived giving the maximum air-kerma rate magnitude and location through a first derivative root finding approach. The second approach consisted of comparing RadShield results with those found by manual shielding design by an American Board of Radiology (ABR)-certified medical physicist for two clinical room situations: two adjacent catheterization labs, and a radiographic and fluoroscopic (R&F) exam room. RadShield's efficacy in finding the maximum air-kerma rate was compared against the geometry-based approach and the overall shielding recommendations by RadShield were compared against the medical physicist's shielding results. Percentage errors between the geometry-based approach and RadShield's approach in finding the magnitude and location of the maximum air-kerma rate was within 0.00124% and 14 mm. RadShield's barrier thickness calculations were found to be within 0.156 mm lead (Pb) and 0.150 mm lead (Pb) for the adjacent catheterization labs and R&F room examples, respectively. However, within the R&F room example, differences in locating the most sensitive calculation point on the floor plan for one of the barriers was not considered in the medical physicist's calculation and was revealed by the RadShield calculations. RadShield is shown to accurately find the maximum values of air-kerma rate and barrier thickness using NCRP Report 147 methodology. Visual inspection alone of the 2D X-ray exam distribution by a medical physicist may not

  3. Design considerations for a Space Station radiation shield for protection from both man-made and natural sources

    Science.gov (United States)

    Bolch, Wesley E.; Peddicord, K. Lee; Felsher, Harry; Smith, Simon

    1994-12-01

    This study was conducted to analyze scenarios involving the use of nuclear-power vehicles in the vicinity of a manned Space Station (SS) in low-earth-orbit (LEO) to quantify their radiological impact to the station crew. In limiting the radiant dose to crew members, mission planners may (1) shut the reactor down prior to reentry, (2) position the vehicle at a prescribed parking distance, and (3) deploy radiation shield about the shutdown reactor. The current report focuses on the third option in which point-kernel gamma-ray shielding calculations were performed for a variety of shield configurations for both nuclear electric propulsion (NEP) and nuclear thermal rocket (NTR) vehicles. For a returning NTR vehicle, calculations indicate that a 14.9 MT shield would be needed to limit the integrated crew exposure to no more than 0.05 Sv over a period of six months (25 percent of the allowable exposure to man-made radiation sources). During periods of low vehicular activity in LEO, the shield may be redeployed about the SS habitation module in order to decrease crew exposures to trapped proton radiations by approximately a factor of 10. The corresponding shield mass required for deployment at a returning NEP vehicle is 2.21 MT. Additional scenarios examined include the radioactivation of various metals as might be found in tools used in EVA activities.

  4. Upflow anaerobic sludge blanket reactor--a review.

    Science.gov (United States)

    Bal, A S; Dhagat, N N

    2001-04-01

    . Concentrated waste (usually sewage sludge) can be added continuously or periodically (semi-batch operation), where it is mixed with the contents of the reactor. Theoretically, the conventional digester is operated as a once-through, completely mixed, reactor. In this particular mode of operation the hydraulic retention time (HRT) is equal to the solids retention time (SRT). Basically, the required process efficiency is related to the sludge retention time (SRT), and hence longer SRT provided, results in satisfactory population (by reproduction) for further waste stabilization. By reducing the hydraulic retention time (HRT) in the conventional mode reactor, the quantity of biological solids within the reactor is also decreased as the solids escape with the effluent. The limiting HRT is reached when the bacteria are removed from the reactor faster than they can grow. Methanogenic bacteria are slow growers and are considered the rate-limiting component in the anaerobic digestion process. The first anaerobic process developed, which separated the SRT from the HRT was the anaerobic contact process. In 1963, Young and McCarty (1968) began work, which eventually led to the development of the anaerobic upflow filter (AF) process. The anaerobic filter represented a significant advance in anaerobic waste treatment, since the filter can trap and maintain a high concentration of biological solids. By trapping these solids, long SRT's could be obtained at large waste flows, necessary to anaerobically treat low strength wastes at nominal temperatures economically. Another anaerobic process which relies on the development of biomass on the surfaces of a media is an expanded bed upflow reactor. The primary concept of the process consists of passing wastewater up through a bed of inert sand sized particles at sufficient velocities to fluidize and partially expand the sand bed. One of the more interesting new processes is the upflow anaerobic sludge blanket process (UASB), which was developed

  5. Purchasing. They Got Out From Under Blanket Orders With a Stores-Based Buy-Out System

    Science.gov (United States)

    Deal, Ralph E.

    1974-01-01

    By having a university 'stores' set up with a blanket order with about 115 local vendors, a university has eliminated petty cash disbursements for small purchases and proliferating purchase orders. (Author/PG)

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

    Science.gov (United States)

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

    2016-10-01

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

  7. Summary of Prometheus Radiation Shielding Nuclear Design Analysis

    Energy Technology Data Exchange (ETDEWEB)

    J. Stephens

    2006-01-13

    This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL & Bettis) shielding nuclear design analyses done for the project.

  8. Neutron shielding material based on colemanite and epoxy resin.

    Science.gov (United States)

    Okuno, Koichi

    2005-01-01

    In recent years, there has been a need for compact shielding design such as self-shielding of a PET cyclotron or upgradation of radiation machinery in existing facilities. In these cases, high performance shielding materials are needed. Concrete or polyethylene have been used for a neutron shield. However, for compact shielding, they fall short in terms of performance or durability. Therefore, a new type of neutron shielding material based on epoxy resin and colemanite has been developed. Slab attenuation experiments up to 40 cm for the new shielding material were carried out using a 252Cf neutron source. Measurement was carried out using a REM-counter, and compared with calculation. The results show that the shielding performance is better than concrete and polyethylene mixed with 10 wt% boron oxide. From the result, we confirmed that the performance of the new material is suitable for practical use.

  9. MPACT Subgroup Self-Shielding Efficiency Improvements

    Energy Technology Data Exchange (ETDEWEB)

    Stimpson, Shane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-31

    Recent developments to improve the efficiency of the MOC solvers in MPACT have yielded effective kernels that loop over several energy groups at once, rather that looping over one group at a time. These kernels have produced roughly a 2x speedup on the MOC sweeping time during eigenvalue calculation. However, the self-shielding subgroup calculation had not been reevaluated to take advantage of these new kernels, which typically requires substantial solve time. The improvements covered in this report start by integrating the multigroup kernel concepts into the subgroup calculation, which are then used as the basis for further extensions. The next improvement that is covered is what is currently being termed as “Lumped Parameter MOC”. Because the subgroup calculation is a purely fixed source problem and multiple sweeps are performed only to update the boundary angular fluxes, the sweep procedure can be condensed to allow for the instantaneous propagation of the flux across a spatial domain, without the need to sweep along all segments in a ray. Once the boundary angular fluxes are considered to be converged, an additional sweep that will tally the scalar flux is completed. The last improvement that is investigated is the possible reduction of the number of azimuthal angles per octant in the shielding sweep. Typically 16 azimuthal angles per octant are used for self-shielding and eigenvalue calculations, but it is possible that the self-shielding sweeps are less sensitive to the number of angles than the full eigenvalue calculation.

  10. Oxygen Abundance Measurements of SHIELD Galaxies

    CERN Document Server

    Haurberg, Nathalie C; Cannon, John M; Marshall, Melissa V

    2015-01-01

    We have derived oxygen abundances for 8 galaxies from the Survey of HI in Extremely Low-mass Dwarfs (SHIELD). The SHIELD survey is an ongoing study of very low-mass galaxies, with M$_{\\rm HI}$ between 10$^{6.5}$ and 10$^{7.5}$ M$_{\\odot}$, that were detected by the Arecibo Legacy Fast ALFA (ALFALFA) survey. H$\\alpha$ images from the WIYN 3.5m telescope show that these 8 SHIELD galaxies each possess one or two active star-forming regions which were targeted with long-slit spectral observations using the Mayall 4m telescope at KPNO. We obtained a direct measurement of the electron temperature by detection of the weak [O III] $\\lambda$4363 line in 2 of the HII regions. Oxygen abundances for the other HII regions were estimated using a strong-line method. When the SHIELD galaxies are plotted on a B-band luminosity-metallicity diagram they appear to suggest a slightly shallower slope to the relationship than normally seen. However, that offset is systematically reduced when the near-infrared luminosity is used ins...

  11. The Tower Shielding Facility: Its glorious past

    Energy Technology Data Exchange (ETDEWEB)

    Muckenthaler, F.J.

    1997-05-07

    The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

  12. New shield for gamma-ray spectrometry

    Science.gov (United States)

    Brar, S. S.; Gustafson, P. F.; Nelson, D. M.

    1969-01-01

    Gamma-ray shield that can be evacuated, refilled with a clean gas, and pressurized for exclusion of airborne radioactive contaminants effectively lowers background noise. Under working conditions, repeated evacuation and filling procedures have not adversely affected the sensitivity and resolution of the crystal detector.

  13. Radiation Shielding for Manned Deep Space Missions

    Science.gov (United States)

    Adams, James H., Jr.

    2003-01-01

    The arrival of the Expedition 1 Crew at the International Space Station represents the beginning of the continuous presence of man in space. Already we are deploying astronauts and cosmonauts for missions of approx. 6 months onboard the ISS. In the future we can anticipate that more people will be in space and they will be there for longer periods. Even with 6-months deployments to the ISS, the radiation exposure that crew members receive is approaching the exposure limits imposed by the governments of the space- faring nations. In the future we can expect radiation protection to be a dominant consideration for long manned missions. Recognizing this, NASA has expanded their research program on radiation health. This program has three components, bioastronautics, fundamental biology and radiation shielding materials. Bioastronautics is concerned with the investigating the effects of radiation on humans. Fundamental biology investigates the basic mechanisms of radiation damage to tissue. Radiation shielding materials research focuses on developing accurate computational tools to predict the radiation shielding effectiveness of materials. It also investigates new materials that can be used for spacecraft. The radiation shielding materials program will be described and examples of results from the ongoing research will be shown.

  14. Lightweight concrete with enhanced neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Brindza, Paul Daniel; Metzger, Bert Clayton

    2016-09-13

    A lightweight concrete containing polyethylene terephthalate in an amount of 20% by total volume. The concrete is enriched with hydrogen and is therefore highly effective at thermalizing neutrons. The concrete can be used independently or as a component of an advanced neutron radiation shielding system.

  15. EFFECTS OF INTERFACES ON GAMMA SHIELDING

    Energy Technology Data Exchange (ETDEWEB)

    Clifford, C.E.

    1963-06-15

    A survey is presented of studies of interface effects in gamma shielding problems. These studies are grouped into three types of approaches, viz.: sources at the interface; radiation backscattered from the interface; and radiation transmitted through the interface. A bibliography of 54 references is included. Limitations on the applicability of the results are discussed. (T.F.H.)

  16. In-beam background suppression shield

    DEFF Research Database (Denmark)

    Santoro, V.; Cai, Xiao Xiao; DiJulio, D. D.

    2015-01-01

    , which do not use a bender to help mitigate the fast neutron background, are the most challenging. For these beam lines we propose the innovative shielding of placing blocks of material directly into the guide system, which allow a minimum attenuation of the cold and thermal fluxes relative...

  17. An electro-hydraulic servo control system research for CFETR blanket RH

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Changqi [Hefei University of Technology, Hefei 230009, Anhui (China); Tang, Hongjun, E-mail: taurustang@126.com [Hefei University of Technology, Hefei 230009, Anhui (China); Qi, Songsong [Hefei University of Technology, Hefei 230009, Anhui (China); Cheng, Yong; Feng, Hansheng; Peng, Xuebing; Song, Yuntao [Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2014-11-15

    Highlights: • We discussed the conceptual design of CFETR blanket RH maintenance system. • The mathematical model of electro-hydraulic servo system was calculated. • A fuzzy adaptive PD controller was designed based on control theory and experience. • The co-simulation models of the system were established with AMESim/Simulink. • The fuzzy adaptive PD algorithm was designed as the core strategy of the system. - Abstract: Based on the technical design requirements of China Fusion Engineering Test Reactor (CFETR) blanket remote handling (RH) maintenance, this paper focus on the control method of achieving high synchronization accuracy of electro-hydraulic servo system. Based on fuzzy control theory and practical experience, a fuzzy adaptive proportional-derivative (PD) controller was designed. Then a more precise co-simulation model was established with AMESim/Simulink. Through the analysis of simulation results, a fuzzy adaptive PD control algorithm was designed as the core strategy of electro-hydraulic servo control system.

  18. Integrated application of upflow anaerobic sludge blanket reactor for the treatment of wastewaters.

    Science.gov (United States)

    Latif, Muhammad Asif; Ghufran, Rumana; Wahid, Zularisam Abdul; Ahmad, Anwar

    2011-10-15

    The UASB process among other treatment methods has been recognized as a core method of an advanced technology for environmental protection. This paper highlights the treatment of seven types of wastewaters i.e. palm oil mill effluent (POME), distillery wastewater, slaughterhouse wastewater, piggery wastewater, dairy wastewater, fishery wastewater and municipal wastewater (black and gray) by UASB process. The purpose of this study is to explore the pollution load of these wastewaters and their treatment potential use in upflow anaerobic sludge blanket process. The general characterization of wastewater, treatment in UASB reactor with operational parameters and reactor performance in terms of COD removal and biogas production are thoroughly discussed in the paper. The concrete data illustrates the reactor configuration, thus giving maximum awareness about upflow anaerobic sludge blanket reactor for further research. The future aspects for research needs are also outlined.

  19. Preliminary lifetime predictions for 304 stainless steel as the LANL ABC blanket material

    Energy Technology Data Exchange (ETDEWEB)

    Park, J.J.; Buksa, J.J.; Houts, M.G.; Arthur, E.D.

    1997-11-01

    The prediction of materials lifetime in the preconceptual Los Alamos National Laboratory (LANL) Accelerator-Based Conversion of Plutonium (ABC) is of utmost interest. Because Hastelloy N showed good corrosion resistance to the Oak Ridge National Laboratory Molten Salt Reactor Experiment fuel salt that is similar to the LANL ABC fuel salt, Hastelloy N was originally proposed for the LANL ABC blanket material. In this paper, the possibility of using 304 stainless steel as a replacement for the Hastelloy N is investigated in terms of corrosion issues and fluence-limit considerations. An attempt is made, based on the previous Fast Flux Test Facility design data, to predict the preliminary lifetime estimate of the 304 stainless steel used in the blanket region of the LANL ABC.

  20. 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...... outcome measures: The sleep latency, number of awakenings and total length of sleep was measured, as was the possible influence on parent- and teacher-rated ADHD symptom load. Results: The results of this study will show that the time it takes for a child to fall asleep is shortened when using a Ball...

  1. Preconceptual engineering design for the APT {sup 3}He target/blanket concept

    Energy Technology Data Exchange (ETDEWEB)

    Mensink, D.L. [Babcock & Wilcox Co., Naval Nuclear Fuel Division, P.O. Box 785, Mt. Athos Rd., Lynchburg, Virginia 24505-0785 (United States); Rose, S.C. Jr. [Reactor Design and Analysis, Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States)

    1995-01-20

    A preconceptual engineering design has been developed for the {sup 3}He Target/Blanket (T/B) System for the Accelerator Production of Tritium Project. This concept uses an array of pressure tubes containing tungsten rods for the neutron spallation source and {sup 3}He gas contained in a metal tank and blanket tubes as the tritium production material. The engineering design is based on a physics model optimized for efficient tritium production. Principle engineering consideration were: provisions for cooling all materials including the {sup 3}He gas; containment of the gas and radionuclides; remote handling; material compatibility; minimization of {sup 3}He, D{sub 2}O, and activated waste; modularity; and manufacturability. The design provides a basis for estimating the cost to implement the system.

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

    Science.gov (United States)

    Furuya, K.; Wakai, E.; Ando, M.; Sawai, T.; Nakamura, K.; Takeuchi, H.; Iwabuchi, A.

    2002-12-01

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

  3. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  4. Multihelix rotating shield brachytherapy for cervical cancer

    Energy Technology Data Exchange (ETDEWEB)

    Dadkhah, Hossein [Department of Biomedical Engineering, University of Iowa, 1402 Seamans Center for the Engineering Arts and Sciences, Iowa City, Iowa 52242 (United States); Kim, Yusung; Flynn, Ryan T., E-mail: ryan-flynn@uiowa.edu [Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States); Wu, Xiaodong [Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 and Department of Electrical and Computer Engineering, University of Iowa, 4016 Seamans Center for the Engineering Arts and Sciences, Iowa City, Iowa 52242 (United States)

    2015-11-15

    Purpose: To present a novel brachytherapy technique, called multihelix rotating shield brachytherapy (H-RSBT), for the precise angular and linear positioning of a partial shield in a curved applicator. H-RSBT mechanically enables the dose delivery using only linear translational motion of the radiation source/shield combination. The previously proposed approach of serial rotating shield brachytherapy (S-RSBT), in which the partial shield is rotated to several angular positions at each source dwell position [W. Yang et al., “Rotating-shield brachytherapy for cervical cancer,” Phys. Med. Biol. 58, 3931–3941 (2013)], is mechanically challenging to implement in a curved applicator, and H-RSBT is proposed as a feasible solution. Methods: A Henschke-type applicator, designed for an electronic brachytherapy source (Xoft Axxent™) and a 0.5 mm thick tungsten partial shield with 180° or 45° azimuthal emission angles and 116° asymmetric zenith angle, is proposed. The interior wall of the applicator contains six evenly spaced helical keyways that rigidly define the emission direction of the partial radiation shield as a function of depth in the applicator. The shield contains three uniformly distributed protruding keys on its exterior wall and is attached to the source such that it rotates freely, thus longitudinal translational motion of the source is transferred to rotational motion of the shield. S-RSBT and H-RSBT treatment plans with 180° and 45° azimuthal emission angles were generated for five cervical cancer patients with a diverse range of high-risk target volume (HR-CTV) shapes and applicator positions. For each patient, the total number of emission angles was held nearly constant for S-RSBT and H-RSBT by using dwell positions separated by 5 and 1.7 mm, respectively, and emission directions separated by 22.5° and 60°, respectively. Treatment delivery time and tumor coverage (D{sub 90} of HR-CTV) were the two metrics used as the basis for evaluation and

  5. On New Limits of the Coefficient of Gravitation Shielding

    Indian Academy of Sciences (India)

    Michele Caputo

    2006-12-01

    New limits of the shielding coefficients in the supposed phenomenon of gravitation shielding have recently become available. The new values are briefly reviewed and discussed in order to update the state of art since some new limits for gravitation shielding are not necessarily the lowest ones which, instead, are those of interest when planning new experimental research or studying theoretically the possible effects of gravitation shielding.

  6. Effective shielding to measure beam current from an ion source

    Energy Technology Data Exchange (ETDEWEB)

    Bayle, H., E-mail: bayle@bergoz.com [Bergoz Instrumentation, Saint-Genis-Pouilly (France); Delferrière, O.; Gobin, R.; Harrault, F.; Marroncle, J.; Senée, F.; Simon, C.; Tuske, O. [CEA, Saclay (France)

    2014-02-15

    To avoid saturation, beam current transformers must be shielded from solenoid, quad, and RFQ high stray fields. Good understanding of field distribution, shielding materials, and techniques is required. Space availability imposes compact shields along the beam pipe. This paper describes compact effective concatenated magnetic shields for IFMIF-EVEDA LIPAc LEBT and MEBT and for FAIR Proton Linac injector. They protect the ACCT Current Transformers beyond 37 mT radial external fields. Measurements made at Saclay on the SILHI source are presented.

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

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

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

  8. Numerical analysis of heat transfer in the first wall of CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Deng, Weiping; Ge, Zhihao; Li, Yuanjie

    2016-04-15

    Highlights: • Detailed numerical analysis of heat transfer in a water-cooling first wall was carried out based on the conceptual design of CFETR WCSB blanket. • Investigation of the influences of buoyancy effect and surface roughness on heat transfer in the water-cooling first wall was presented. • Analysis of the effect of the front wall thickness on temperature was carried out for the water-cooling first wall design. • Simulation results of two 1D CFD methods were evaluated by the 3D CFD data. - Abstract: China Fusion Engineering Test Reactor (CFETR), the first fusion reactor experiment project planned in China, is now being investigated in detail. Recently, a conceptual structural design of the Water-Cooled-Solid-Breeder (WCSB) blanket was proposed as one of the breeding blanket candidates for CFETR. In this research, based on the present design of the CFETR WCSB blanket, the heat transfer performance in the first wall (FW) under the pressurized water cooling condition was analyzed. The 3D computational fluid dynamics (CFD) results show that the maximal temperature of the FW will not exceed the limited temperature under normal or even higher heat flux condition. In addition, the effect of buoyancy on heat transfer is negligible under both conditions. The influence of roughness becomes increasingly important when the roughness height lies in the fully turbulent regime. The maximal temperature increases approximately linearly as the thickness of the front wall increases. It is also found that the heat flux and the local heat transfer coefficient are extremely non-uniform in the circumferential direction. Two 1D CFD methods are also evaluated by 3D CFD data, with the conclusion that both 1D results have some differences with the 3D data. The improved 1D method is more accurate than the former one. However, we ascertain that 1D methods should be used with caution for the water-cooling FW design.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  10. Neutronics Optimization of Tritium Breeding Blan-ket for the FDS

    Institute of Scientific and Technical Information of China (English)

    郑善良; 吴宜灿; 黄群英

    2002-01-01

    Neutronics optimization calculations have been performed for thc tritium breeding blankets with solid ceramic breeder Li2O and liquid eutectie breeder Li17Pb83, respectively,based on a 2-D geometrical configuration using the Monte Carlo neutron-photon transport code MCNP/4B. The effects of beryllium, 6Li enrichment and various structural materials on Tritium Breeding Ratio have been systematically analyzed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji [ed.

    1998-03-01

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

  12. Radiation shielding evaluation based on five years of data from a busy CyberKnife center.

    Science.gov (United States)

    Yang, Jun; Feng, Jing

    2014-11-08

    We examined the adequacy of existing shielding guidelines using five-year clinical data from a busy CyberKnife center. From June 2006 through July 2011, 1,370 patients were treated with a total of 4,900 fractions and 680,691 radiation beams using a G4 CyberKnife. Prescription dose and total monitor units (MU) were analyzed to estimate the shielding workload and modulation factor. In addition, based on the beam's radiation source position, targeting position, MU, and beam collimator size, the MATLAB program was used to project each beam toward the shielding barrier. The summation of the projections evaluates the distribution of the shielding load. On average, each patient received 3.6 fractions, with an average 9.1 Gy per fraction prescribed at the 71.1% isodose line, using 133.7 beams and 6,200 MU. Intracranial patients received an average of 2.7 fractions, with 8.6 Gy per fraction prescribed at the 71.4% isodose line, using 133 beams and 5,083 MU. Extracranial patients received an average of 3.94 fractions, with 9.2 Gy per frac- tion prescribed at the 71% isodose line, using 134 beams and 6,514 MU. Most- used collimator sizes for intracranial patients were smaller (7.5 to 20 mm) than for extracranial patients (20 to 40 mm). Eighty-five percent of the beams exited through the floor, and about 40% of the surrounding wall area received no direct beam. For the rest of the wall, we found "hot" areas that received above-average MU. The locations of these areas were correlated with the projection of the nodes for extracranial treatments. In comparison, the beam projections on the wall were more spread for intracranial treatments. The maximum MU any area received from intracranial treatment was less than 0.25% of total MU used for intracranial treatments, and was less than 1.2% of total MU used for extracranial treatments. The combination of workload, modulation factor, and use factor in our practice are about tenfold less than recommendations in the existing Cyber

  13. Electromagnetic shielding. Citations from the NTIS data base

    Science.gov (United States)

    Reed, W. E.

    1980-06-01

    The bibliography presents research on electromagnetic shielding of electronic and electrical equipment personnel, and ordnance. The shielding effectiveness of materials and structures is covered. Nuclear electromagnetic pulse shielding is included. This updated bibliography contains 301 abstracts, 19 of which are new entries to the previous edition.

  14. 30 CFR 56.14213 - Ventilation and shielding for welding.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Ventilation and shielding for welding. 56.14213... Equipment Safety Practices and Operational Procedures § 56.14213 Ventilation and shielding for welding. (a) Welding operations shall be shielded when performed at locations where arc flash could be hazardous...

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

    Science.gov (United States)

    Kerwat, Klaus; Piechowiak, Karolin; Wulf, Hinnerk

    2013-01-01

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

  16. Preflow stresses in Martian rampart ejecta blankets - A means of estimating the water content

    Science.gov (United States)

    Woronow, A.

    1981-01-01

    Measurements of extents of rampart ejecta deposits as a function of the size of the parent craters support models which, for craters larger than about 6 km diameter, constrain ejecta blankets to all have a similar maximum thickness regardless of the crater size. These volatile-rich ejecta blankets may have failed under their own weights, then flowed radially outward. Assuming this to be so, some of the physicomechanical properties of the ejecta deposits at the time of their emplacement can then be determined. Finite-element studies of the stress magnitudes, distributions, and directions in hypothetical Martian rampart ejecta blankets reveal that the material most likely failed when the shear stresses were less than 500 kPa and the angle of internal friction was between 26 and 36 deg. These figures imply that the ejecta has a water content between 16 and 72%. Whether the upper limit or the lower limit is more appropriate depends on the mode of failure which one presumes: namely, viscous flow of plastic deformation.

  17. Parametric Weight Comparison of Advanced Metallic, Ceramic Tile, and Ceramic Blanket Thermal Protection Systems

    Science.gov (United States)

    Myers, David E.; Martin, Carl J.; Blosser, Max L.

    2000-01-01

    A parametric weight assessment of advanced metallic panel, ceramic blanket, and ceramic tile thermal protection systems (TPS) was conducted using an implicit, one-dimensional (I-D) finite element sizing code. This sizing code contained models to account for coatings fasteners, adhesives, and strain isolation pads. Atmospheric entry heating profiles for two vehicles, the Access to Space (ATS) vehicle and a proposed Reusable Launch Vehicle (RLV), were used to ensure that the trends were not unique to a certain trajectory. Ten TPS concepts were compared for a range of applied heat loads and substructural heat capacities to identify general trends. This study found the blanket TPS concepts have the lightest weights over the majority of their applicable ranges, and current technology ceramic tiles and metallic TPS concepts have similar weights. A proposed, state-of-the-art metallic system which uses a higher temperature alloy and efficient multilayer insulation was predicted to be significantly lighter than the ceramic tile stems and approaches blanket TPS weights for higher integrated heat loads.

  18. AB Blanket for Cities (for continual pleasant weather and protection from chemical, biological and radioactive weapons)

    CERN Document Server

    Bolonkin, Alexander

    2009-01-01

    In a series of previous articles (see references) the author offered to cover a city or other important large installations or subregions by a transparent thin film supported by a small additional air overpressure under the form of an AB Dome. The building of a gigantic inflatable AB Dome over an empty flat surface is not difficult. However, if we want to cover a city, garden, forest or other obstacle course we cannot easily deploy the thin film over building or trees. In this article is suggested a new method which solves this problem. The idea is to design a double film blanket filled by light gas (for example, methane, hydrogen, or helium). Sections of this AB Blanket are lighter then air and fly in atmosphere. They can be made on a flat area (serving as an assembly area) and delivered by dirigible or helicopter to station at altitude over the city. Here they connect to the already assembled AB Blanket subassemblies, cover the city in an AB Dome and protect it from bad weather, chemical, biological and rad...

  19. Blanketing effect of expansion foam on liquefied natural gas (LNG) spillage pool.

    Science.gov (United States)

    Zhang, Bin; Liu, Yi; Olewski, Tomasz; Vechot, Luc; Mannan, M Sam

    2014-09-15

    With increasing consumption of natural gas, the safety of liquefied natural gas (LNG) utilization has become an issue that requires a comprehensive study on the risk of LNG spillage in facilities with mitigation measures. The immediate hazard associated with an LNG spill is the vapor hazard, i.e., a flammable vapor cloud at the ground level, due to rapid vaporization and dense gas behavior. It was believed that high expansion foam mitigated LNG vapor hazard through warming effect (raising vapor buoyancy), but the boil-off effect increased vaporization rate due to the heat from water drainage of foam. This work reveals the existence of blocking effect (blocking convection and radiation to the pool) to reduce vaporization rate. The blanketing effect on source term (vaporization rate) is a combination of boil-off and blocking effect, which was quantitatively studied through seven tests conducted in a wind tunnel with liquid nitrogen. Since the blocking effect reduces more heat to the pool than the boil-off effect adds, the blanketing effect contributes to the net reduction of heat convection and radiation to the pool by 70%. Water drainage rate of high expansion foam is essential to determine the effectiveness of blanketing effect, since water provides the boil-off effect.

  20. Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonat Sen; Gilles Youinou

    2013-02-01

    It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this case the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)

  1. Upflow Sludge Blanket Filtration (USBF: An Innovative Technology in Activated Sludge Process

    Directory of Open Access Journals (Sweden)

    R Saeedi

    2010-06-01

    Full Text Available Background: A new biological domestic wastewater treatment process, which has been presented these days in activated sludge modification, is Upflow Sludge Blanket Filtration (USBF. This process is aerobic and acts by using a sludge blanket in the separator of sedimentation tank. All biological flocs and suspended solids, which are presented in the aeration basin, pas through this blanket. The performance of a single stage USBF process for treatment of domestic wastewater was studied in laboratory scale.Methods: The pilot of USBF has been made from fiberglass and the main electromechanical equipments consisted of an air com­pressor, a mixing device and two pumps for sludge return and wastewater injection. The wastewater samples used for the experiments were prepared synthetically to have qualitative characteristics similar to a typical domestic wastewater (COD= 277 mg/l, BOD5= 250 mg/l and TSS= 1 mg/l.Results: On the average, the treatment system was capable to remove 82.2% of the BOD5 and 85.7% of COD in 6 h hydraulic re­tention time (HRT. At 2 h HRT BOD and COD removal efficiencies dramatically reduced to 50% and 46.5%, respectively.Conclusion: Even by increasing the concentrations of pollutants to as high as 50%, the removal rates of all pollutants were re­mained similar to the HRT of 6 h.

  2. First wall fabrication of 1/3 scale china dual functional lithium lead blanket

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Bo, E-mail: bo.huang@fds.org.cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Zhai, Yutao [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Zhang, Junyu [University of Science and Technology of China, Hefei, Anhui 230027 (China); Li, Chunjing; Wu, Qingsheng [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Huang, Qunying [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • RAFM rectangular tubes were fabricated by cold drawing, and the dimensional accuracy and mechanical properties of rectangular tubes were tested. • Rectangular tubes were bent by rotary bending, and milled plates were curved by molding. Its accuracy meets the requirement for TBM assembly. • FW were pre-sealed by electron beam welding, and assembled by hot isostatic pressing–diffusion bonding. • The as-HIPed FW mock-up was tested by optical observation and X-ray detection, it revealed obviously that the tubes and plates were bonded well. - Abstract: The dual functional lithium lead blanket is chosen as one of the candidate blankets for China fusion reactor, for its advantages of tritium breeding and good heat exchange performance. As one of the most important components of the blanket, the first wall (FW) is assembled with China low activation martensitic (CLAM) rectangular tubes and plates by hot isostatic pressing (HIP)–diffusion bonding (DB). In this work, the rectangular tube fabrication and FW assembly were carried out in order to verify the feasibility of the FW fabrication scheme. The mechanical property and dimensional accuracy of CLAM rectangular tubes were tested, the microstructure observation and non-destructive detection revealed the sound of the FW mock-up, and the reliability of the FW mock-ups is under evaluation.

  3. Fetal shielding combined with state of the art CT dose reduction strategies during maternal chest CT

    Energy Technology Data Exchange (ETDEWEB)

    Chatterson, Leslie C., E-mail: lch088@mail.usask.ca [Department of Diagnostic Imaging, University of Saskatchewan (Canada); Leswick, David A.; Fladeland, Derek A. [Department of Diagnostic Imaging, University of Saskatchewan (Canada); Hunt, Megan M.; Webster, Stephen [Saskatchewan Ministry of Labour Relations and Workplace Safety (Canada); Lim, Hyun [Department of Community Health and Epidemiology, College of Medicine, University of Saskatchewan (Canada)

    2014-07-15

    Purpose: Custom bismuth-antimony shields were previously shown to reduce fetal dose by 53% on an 8DR (detector row) CT scanner without dynamic adaptive section collimation (DASC), automatic tube current modulation (ATCM) or adaptive statistical iterative reconstruction (ASiR). The purpose of this study is to compare the effective maternal and average fetal organ dose reduction both with and without bismuth-antimony shields on a 64DR CT scanner using DASC, ATCM and ASiR during maternal CTPA. Materials and methods: A phantom with gravid prosthesis and a bismuth-antimony shield were used. Thermoluminescent dosimeters (TLDs) measured fetal radiation dose. The average fetal organ dose and effective maternal dose were determined using 100 kVp, scanning from the lung apices to the diaphragm utilizing DASC, ATCM and ASiR on a 64DR CT scanner with and without shielding in the first and third trimester. Isolated assessment of DASC was done via comparing a new 8DR scan without DASC to a similar scan on the 64DR with DASC. Results: Average third trimester unshielded fetal dose was reduced from 0.22 mGy ± 0.02 on the 8DR to 0.13 mGy ± 0.03 with the conservative 64DR protocol that included 30% ASiR, DASC and ATCM (42% reduction, P < 0.01). Use of a shield further reduced average third trimester fetal dose to 0.04 mGy ± 0.01 (69% reduction, P < 0.01). The average fetal organ dose reduction attributable to DASC alone was modest (6% reduction from 0.17 mGy ± 0.02 to 0.16 mGy ± 0.02, P = 0.014). First trimester fetal organ dose on the 8DR protocol was 0.07 mGy ± 0.03. This was reduced to 0.05 mGy ± 0.03 on the 64DR protocol without shielding (30% reduction, P = 0.009). Shields further reduced this dose to below accurately detectable levels. Effective maternal dose was reduced from 4.0 mSv on the 8DR to 2.5 mSv on the 64DR scanner using the conservative protocol (38% dose reduction). Conclusion: ASiR, ATCM and DASC combined significantly reduce effective maternal and fetal

  4. How stable are the 'stable ancient shields'?

    Science.gov (United States)

    Viola, Giulio; Mattila, Jussi

    2014-05-01

    "Archean cratons are relatively flat, stable regions of the crust that have remained undeformed since the Precambrian, forming the ancient cores of the continents" (King, EPSL, 2005). While this type of statement is supported by a wealth of constraints in the case of episodes of thoroughgoing ductile deformation affecting shield regions of Archean and also Peleoproterozoic age, a growing amount of research indicates that shields are not nearly as structurally stable within the broad field of environmental conditions leading to brittle deformation. In fact, old crystalline basements usually present compelling evidence of long brittle deformation histories, often very complex and challenging to unfold. Recent structural and geochronological studies point to a significant mechanical instability of the shield areas, wherein large volumes of 'stable' rocks actually can become saturated with fractures and brittle faults soon after regional cooling exhumes them to below c. 300-350° C. How cold, rigid and therefore strong shields respond to applied stresses remains, however, still poorly investigated and understood. This in turn precludes a better definition of the shallow rheological properties of large, old crystalline blocks. In particular, we do not yet have good constraints on the mechanisms of mechanical reactivation that control the partial (if not total) accommodation of new deformational episodes by preexisting structures, which remains a key to untangle brittle histories lasting several hundred Myr. In our analysis, we use the Svecofennian Shield (SS) as an example of a supposedly 'stable' region with Archean nucleii and Paleoproterozoic cratonic areas to show how it is possible to unravel the details of brittle histories spanning more than 1.5 Gyr. New structural and geochronological results from Finland are integrated with a review of existing data from Sweden to explore how the effects of far-field stresses are partitioned within a shield, which was growing

  5. Shielding of Electronic Systems against Transient Electromagnetic Interferences

    Directory of Open Access Journals (Sweden)

    H. Herlemann

    2005-01-01

    Full Text Available In order to protect electronic systems against the effects of transient electromagnetic interferences, shields made of electrically conductive material can be used. The subject of this paper is an electrically conductive textile. When applying the shield, a reliable measure is needed in order to determine the effectiveness of the shield to protect against electromagnetic pulses. For this purpose, a time domain measurement technique is presented using double exponential pulses. With these pulses, the susceptibility of an operating electronic device with and without the shield is determined. As a criterion of quality of a shield, the breakdown failure rate found in both cases is compared.

  6. Measurement of shielding characteristics in the prototype FBR Monju

    Energy Technology Data Exchange (ETDEWEB)

    Usami, Shin; Sasaki, Kenji; Deshimaru, Takehide; Nakashima, Fumiaki [Japan Nuclear Cycle Development Institute, Tsuruga, Fukui (Japan)

    2000-03-01

    In the prototype fast breeder reactor Monju, shielding measurements were made around the reactor core, the primary heat transport system (PHTS), and the fuel handling and storage system during the system start-up tests at different power levels between 0% and 45%. The objectives of the tests were to evaluate the margins by which the shielding performance exceeds the original design requirements, to demonstrate the validity of the shielding analysis method, and to acquire basic data for use in future FBR design. This paper summarizes the important features of the Monju shielding structures and the shielding measurement. (author)

  7. Shield Insertion to Minimize Noise Amplitude in Global Interconnects

    Directory of Open Access Journals (Sweden)

    Kalpana.A.B

    2012-09-01

    Full Text Available Shield insertion is an effective technique for minimise crosstalk noise and signal delay uncertainty .To reduce the effects of coupling uniform or simultaneous shielding may be used on either or both sides of a signal line. Shields are ground or power lines placed between two signal wires to prevent direct coupling between them as the shield width increases, the noise amplitude decreases, in this paper inserting a shield line between two coupled interconnects is shown to be more effective in reducing crosstalk noise for different technology nodes .

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-04

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and

  9. Detailed technical plan for Test Program Element-III (TPE-III) of the first wall/blanket shield engineering test program

    Energy Technology Data Exchange (ETDEWEB)

    Turner, L.R.; Praeg, W.F.

    1982-03-01

    The experimental requirements, test-bed design, and computational requirements are reviewed and updated. Next, in Sections 3, 4 and 5, the experimental plan, instrumentation, and computer plan, respectively, are described. Finally, Section 6 treats other considerations, such as personnel, outside participation, and distribution of results.

  10. Evaluation of Personal Shields Used in Selected Radiology Departments

    Directory of Open Access Journals (Sweden)

    Mohsen Salmanvandi

    2015-05-01

    Full Text Available Introduction The purpose of this study was to evaluate personal shields in radiation departments of hospitals affiliated to Mashhad University of Medical Sciences. Materials and Methods First, the information related to 109 personal shields was recorded and evaluated by imaging equipment. Afterwards, the equivalent lead thickness (ELT of 62 personal shields was assessed, using dosimeter and standard lead layers at 100 kVp. Results In this study, 109 personal shields were assessed in terms of tears, holes and cracks. The results showed that 18 shields were damaged. Moreover, ELT was evaluated in 62 shields. As the results indicated, ELT was unacceptable in 8 personal shields and lower than expected in 9 shields. Conclusion According to the results, 16.5% of personal shields had defects (tears, holes and cracks and 13% of them were unacceptable in terms of ELT and needed to be replaced. Therefore, regular quality control of personal shields and evaluation of new shields are necessary at any radiation department.

  11. PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology

    Directory of Open Access Journals (Sweden)

    Mario Matijević

    2016-01-01

    Full Text Available The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS were demonstrated when applied to a realistic deep penetration Monte Carlo (MC shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR parameters is based on deterministic transport theory (SN method providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between SN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

  12. Electrodynamic Dust Shield for Space Applications

    Science.gov (United States)

    Mackey, Paul J.; Johansen, Michael R.; Olsen, Robert C.; Raines, Matthew G.; Phillips, James R., III; Cox, Rachel E.; Hogue, Michael D.; Pollard, Jacob R. S.; Calle, Carlos I.

    2016-01-01

    Dust mitigation technology has been highlighted by NASA and the International Space Exploration Coordination Group (ISECG) as a Global Exploration Roadmap (GER) critical technology need in order to reduce life cycle cost and risk, and increase the probability of mission success. The Electrostatics and Surface Physics Lab in Swamp Works at the Kennedy Space Center has developed an Electrodynamic Dust Shield (EDS) to remove dust from multiple surfaces, including glass shields and thermal radiators. Further development is underway to improve the operation and reliability of the EDS as well as to perform material and component testing outside of the International Space Station (ISS) on the Materials on International Space Station Experiment (MISSE). This experiment is designed to verify that the EDS can withstand the harsh environment of space and will look to closely replicate the solar environment experienced on the Moon.

  13. Thermoforming plastic in lead shield construction

    Energy Technology Data Exchange (ETDEWEB)

    Abrahams, M.E.; Chow, C.H.; Loyd, M.D. (Univ. of Texas Medical Branch, Galveston (USA))

    1989-09-01

    Radiation treatments using low energy X-rays or electrons frequently require a final field defining shield to be placed on the patient's skin. A custom made lead cut-out is used to provide a close fit to a particular patient's surface contours. We have developed a procedure which utilizes POLYFORM thermoplastic to obtain a negative mold of the patient instead of the traditional plaster bandage or dental impression gel. The Polyform is softened in warm water, molded carefully over the patient's surface, and is removed when set or hardened, usually within five minutes. Then lead sheet cut-outs can be formed within this negative. For shielding cut-outs requiring thicker lead sheet, a positive is made from dental stone using this Polyform negative. We have found this procedure to be neat, fast and comfortable for both patient and the dosimetrist.

  14. Thermoforming plastic in lead shield construction.

    Science.gov (United States)

    Abrahams, M E; Chow, C H; Loyd, M D

    1989-09-01

    Radiation treatments using low energy X-rays or electrons frequently require a final field defining shield to be placed on the patient's skin. A custom made lead cut-out is used to provide a close fit to a particular patient's surface contours. We have developed a procedure which utilizes POLYFORM thermoplastic to obtain a negative mold of the patient instead of the traditional plaster bandage or dental impression gel. The Polyform is softened in warm water, molded carefully over the patient's surface, and is removed when "set" or hardened, usually within five minutes. Then lead sheet cut-outs can be formed within this negative. For shielding cut-outs requiring thicker lead sheet, a positive is made from dental stone using this Polyform negative. We have found this procedure to be neat, fast and comfortable for both patient and the dosimetrist.

  15. In-Beam Background Suppression Shield

    CERN Document Server

    Santoro, V; DiJulio, D D; Ansell, S; Bentley, P M

    2015-01-01

    The long (3ms) proton pulse of the European Spallation Source (ESS) gives rise to unique and potentially high backgrounds for the instrument suite. In such a source an instrument capabilities will be limited by it's Signal to Noise (S/N) ratio. The instruments with a direct view of the moderator, which do not use a bender to help mitigate the fast neutron background, are the most challenging. For these beam lines we propose the innovative shielding of placing blocks of material directly into the guide system, which allow a minimum attenuation of the cold and thermal fluxes relative to the background suppression. This shielding configuration has been worked into a beam line model using Geant4. We study particularly the advantages of single crystal sapphire and silicon blocks .

  16. Grounding and shielding circuits and interference

    CERN Document Server

    Morrison, Ralph

    2016-01-01

    Applies basic field behavior in circuit design and demonstrates how it relates to grounding and shielding requirements and techniques in circuit design This book connects the fundamentals of electromagnetic theory to the problems of interference in all types of electronic design. The text covers power distribution in facilities, mixing of analog and digital circuitry, circuit board layout at high clock rates, and meeting radiation and susceptibility standards. The author examines the grounding and shielding requirements and techniques in circuit design and applies basic physics to circuit behavior. The sixth edition of this book has been updated with new material added throughout the chapters where appropriate. The presentation of the book has also been rearranged in order to reflect the current trends in the field.

  17. EMC Test Report Electrodynamic Dust Shield

    Science.gov (United States)

    Carmody, Lynne M.; Boyette, Carl B.

    2014-01-01

    This report documents the Electromagnetic Interference E M I evaluation performed on the Electrodynamic Dust Shield (EDS) which is part of the MISSE-X System under the Electrostatics and Surface Physics Laboratory at Kennedy Space Center. Measurements are performed to document the emissions environment associated with the EDS units. The purpose of this report is to collect all information needed to reproduce the testing performed on the Electrodynamic Dust Shield units, document data gathered during testing, and present the results. This document presents information unique to the measurements performed on the Bioculture Express Rack payload; using test methods prepared to meet SSP 30238 requirements. It includes the information necessary to satisfy the needs of the customer per work order number 1037104. The information presented herein should only be used to meet the requirements for which it was prepared.

  18. SHIELD II: WSRT HI Spectral Line Observations

    Science.gov (United States)

    Gordon, Alex Jonah Robert; Cannon, John M.; Adams, Elizabeth A.; SHIELD II Team

    2016-01-01

    The "Survey of HI in Extremely Low-mass Dwarfs II" ("SHIELD II") is a multiwavelength, legacy-class observational campaign that is facilitating the study of both internal and global evolutionary processes in low-mass dwarf galaxies discovered by the Arecibo Legacy Fast ALFA (ALFALFA) survey. We present new results from WSRT HI spectral line observations of 22 galaxies in the SHIELD II sample. We explore the morphology and kinematics by comparing images of the HI surface densities and the intensity weighted velocity fields with optical images from HST, SDSS, and WIYN. In most cases the HI and stellar populations are cospatial; projected rotation velocities range from less than 10 km/s to roughly 30 km/s.Support for this work was provided by NSF grant AST-1211683 to JMC at Macalester College, and by NASA through grant GO-13750 from the Space Telescope Science Institute, which is operated by AURA, Inc., under NASA contract NAS5-26555.

  19. SQUID holder with high magnetic shielding

    Science.gov (United States)

    Rigby, K. W.; Marek, D.; Chui, T. C. P.

    1990-01-01

    A SQUID holder designed for high magnetic shielding is discussed. It is shown how to estimate the attenuation of the magnetic field from the normal magnetic modes for an approximate geometry. The estimate agrees satisfactorily with the attenuation measured with a commercial RF SQUID installed in the holder. The holder attenuates external magnetic fields by more than 10 to the 9th at the SQUID input. With the SQUID input shorted, the response to external fields is 0.00001 Phi(0)/G.

  20. Homogeneous Dielectric Equivalents of Composite Material Shields

    Directory of Open Access Journals (Sweden)

    P. Tobola

    2009-04-01

    Full Text Available The paper deals with the methodology of replacing complicated parts of an airplane skin by simple homogeneous equivalents, which can exhibit similar shielding efficiency. On one hand, the airplane built from the virtual homogeneous equivalents can be analyzed with significantly reduced CPU-time demands and memory requirements. On the other hand, the equivalent model can estimate the internal fields satisfactory enough to evaluate the electromagnetic immunity of the airplane.

  1. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  2. Heavy Metal Pad Shielding during Fluoroscopic Interventions

    OpenAIRE

    Dromi, Sergio; Wood, Bradford J.; Oberoi, Jay; Neeman, Ziv

    2006-01-01

    Significant direct and scatter radiation doses to patient and physician may result from routine interventional radiology practice. A lead-free disposable tungsten antimony shielding pad was tested in phantom patients during simulated diagnostic angiography procedures. Although the exact risk of low doses of ionizing radiation is unknown, dramatic dose reductions can be seen with routine use of this simple, sterile pad made from lightweighttungsten antimony material.

  3. Synthesis and fabrication of lithium-titanate pebbles for ITER breeding blanket by solid state reaction and spherodization

    Energy Technology Data Exchange (ETDEWEB)

    Mandal, D., E-mail: dmandal@barc.gov.i [Chemical Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Shenoi, M.R.K.; Ghosh, S.K. [Chemical Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2010-08-15

    {sup 6}Li produces tritium by (n, {alpha}) nuclear reaction, {sup 6}Li + {sup 1}n {yields} {sup 4}He + {sup 3}H. Lithium titanate (Li{sub 2}TiO{sub 3}) enriched with {sup 6}Li, is the most promising candidate for solid test blanket module (TBM) material for fusion reactors. Various processes are reported in literature for the fabrication of Li{sub 2}TiO{sub 3} pebbles for its use as TBM material. A process has been developed based on the solid state reaction of lithium-carbonate and titanium-dioxide for the synthesis of lithium titanate and pebble fabrication by extrusion, spherodization and sintering. This paper discusses the sequence of steps followed in this process and the properties obtained. Analytical grade titanium-dioxide and lithium-carbonate were taken in stoichiometric ratio and were milled to ensure thorough intimate mixing and obtain fine particles less than 45 {mu}m before its calcination at 900 {sup o}C. Following calcination, the agglomerated product was again milled to fine particles of size less than 45 {mu}m. Aqueous solution of ploy-vinyl-alcohol was added as binder prior to its feeding in the extruder. The extruded strips were spherodized and spherical pebbles were dried and sintered at 900 {sup o}C for different duration. Pebbles of desired density and porosity were obtained by suitable combination of sintering temperature and duration of sintering. Properties of the prepared pebbles were also characterized for sphericity, pore size distribution, grain size, crushing load strength, etc. The values were found to be conforming to the desired properties for use as solid breeder. The attractive feature of this process is almost no waste generation.

  4. An attenuation Layer for Electromagnetic Shielding in X- Band Frequency

    Directory of Open Access Journals (Sweden)

    vida Zaroushani

    2015-06-01

    Full Text Available Uncontrolled exposure to X-band frequency leads to health damage. One of the principles of radiation protection is shielding. But, conventional shielding materials have disadvantages. Therefore, studies of novel materials, as an alternative to conventional shielding materials, are required to obtain new electromagnetic shielding material. Therefore, this study investigated the electromagnetic shielding of two component epoxy thermosetting resin for the X - band frequency with workplace approach. Two components of epoxy resin mixed according to manufacturing instruction with the weight ratio that was 100:10 .Epoxy plates fabricated in three different thicknesses (2, 4 and 6mm and shielding effectiveness measured by Vector Network Analyzer. Then, shielding effectiveness measured by the scattering parameters.The results showed that 6mm thickness of epoxy had the highest and 2mm had the lowest average of shielding effectiveness in X-band frequency that is 4.48 and 1.9 dB, respectively. Also, shielding effectiveness increased by increasing the thickness. But this increasing is useful up to 4mm. Percentage shielding effectiveness of attenuation for 6, 4 and 2mm thicknesses is 64.35%, 63.31% and 35.40%. Also, attenuation values for 4mm and 6mm thicknesses at 8.53 GHz and 8.52 GHz frequency are 77.15% and 82.95%, respectively, and can be used as favourite shields for the above frequency. 4mm-Epoxy is a suitable candidate for shielding application in X-band frequency range but, in the lower section, 6mm thickness is recommended. Finely, the shielding matrix can be used for selecting the proper thickness for electromagnetic shielding in X- Band frequency.

  5. Beta Bremsstrahlung dose in concrete shielding

    Science.gov (United States)

    Manjunatha, H. C.; Chandrika, B. M.; Rudraswamy, B.; Sankarshan, B. M.

    2012-05-01

    In a nuclear reactor, beta nuclides are released during nuclear reactions. These betas interact with shielding concrete and produces external Bremsstrahlung (EB) radiation. To estimate Bremsstrahlung dose and shield efficiency in concrete, it is essential to know Bremsstrahlung distribution or spectra. The present work formulated a new method to evaluate the EB spectrum and hence Bremsstrahlung dose of beta nuclides (32P, 89Sr, 90Sr-90Y, 90Y, 91Y, 208Tl, 210Bi, 234Pa and 40K) in concrete. The Bremsstrahlung yield of these beta nuclides in concrete is also estimated. The Bremsstrahlung yield in concrete due to 90Sr-90Y is higher than those of other given nuclides. This estimated spectrum is accurate because it is based on more accurate modified atomic number (Zmod) and Seltzer's data, where an electron-electron interaction is also included. Presented data in concrete provide a quick and convenient reference for radiation protection. The present methodology can be used to calculate the Bremsstrahlung dose in nuclear shielding materials. It can be quickly employed to give a first pass dose estimate prior to a more detailed experimental study.

  6. Mars14 Monte Carlo simulation for the shielding studies of the J-PARC 3 GeV ring.

    Science.gov (United States)

    Nakao, Noriaki; Mokhov, Nikolai; Yamamoto, Kazami; Irie, Yoshiro; Drozhdin, Alexander

    2005-01-01

    MARS14 Monte Carlo simulations were performed for collimation and shielding studies of the J-PARC 3 GeV synchrotron ring. The beam line module locations in the 348.3 m ring and the curved tunnel sections were described by the 'MAD-MARS beam line builder' tool. A 400 MeV proton beam loss distribution, calculated with the STRUCT code, was used as a 4 kW source term in the collimator region, with 1 kW source terms in the injection and extraction regions at 400 MeV and 3 GeV, respectively. Deep penetration calculations were carried out with good statistics using a newly developed three-dimensional multi-layer technique. Prompt dose-rate distributions were calculated inside and outside the concrete and soil shield up to the ground level. Using the calculation results obtained thus, an effective shielding design was made.

  7. Beta radiation shielding with lead and plastic: effect on bremsstrahlung radiation when switching the shielding order.

    Science.gov (United States)

    Van Pelt, Wesley R; Drzyzga, Michael

    2007-02-01

    Lead and plastic are commonly used to shield beta radiation. Radiation protection literature is ubiquitous in advising the placement of plastic first to absorb all the beta particles before any lead shielding is used. This advice is based on the well established theory that radiative losses (bremsstrahlung production) are more prevalent in higher atomic number (Z) materials than in low Z materials. Using 32P beta radiation, we measured bremsstrahlung photons transmitted through lead and plastic (Lucite) shielding in different test configurations to determine the relative efficacy of lead alone, plastic alone, and the positional order of lead and plastic. With the source (32P) and detector held at a constant separation distance, we inserted lead and/or plastic absorbers and measured the reduction in bremsstrahlung radiation level measured by the detector. With these test conditions, analysis of measured bremsstrahlung radiation in various thicknesses and configurations of lead and plastic shielding shows the following: placing plastic first vs. lead first reduces the transmitted radiation level only marginally (10% to 40%); 2 mm of additional lead is sufficient to correct the "mistake" of placing the lead first; and for equal thicknesses or weights of lead and plastic, lead is a more efficient radiation shield than plastic.

  8. Analysis and improvement of cyclotron thallium target room shield.

    Science.gov (United States)

    Hajiloo, N; Raisali, G; Aslani, G

    2008-01-01

    Because of high neutron and gamma-ray intensities generated during bombardment of a thallium-203 target, a thallium target-room shield and different ways of improving it have been investigated. Leakage of neutron and gamma ray dose rates at various points behind the shield are calculated by simulating the transport of neutrons and photons using the Monte Carlo N Particle transport computer code. By considering target-room geometry, its associated shield and neutron and gamma ray source strengths and spectra, three designs for enhancing shield performance have been analysed: a shielding door at the maze entrance, covering maze walls with layers of some effective materials and adding a shadow-shield in the target room in front of the radiation source. Dose calculations were carried out separately for different materials and dimensions for all the shielding scenarios considered. The shadow-shield has been demonstrated to be one suitable for neutron and gamma dose equivalent reduction. A 7.5-cm thick polyethylene shadow-shield reduces both dose equivalent rate at maze entrance door and leakage from the shield by a factor of 3.

  9. Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

    Directory of Open Access Journals (Sweden)

    I. Kodeli

    2008-01-01

    Full Text Available An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL concept will be performed in 2008 in the Frascati Neutron Generator (FNG in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR, their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package, the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n and (n,3n reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

  10. Fast Breeder Blanket Facility (FBBF). Quarterly progress report, January 1, 1976--August 30, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Ott, K.O. (ed.)

    1976-08-01

    The work performed was primarily concerned with the preparation of the experiments to be performed on the Fast Breeder Blanket Facility (FBBF) and the corresponding analysis. The work on the experimental program has been started. Since experiments are subject to safety constraints, a safety investigation program (for a hypothetically flooded facility) is reported. The neutronics part of the preanalysis is also reported. The testing of the first configuration has largely been prepared. The identification of the experiment need has been worked on extensively, largely through unsponsored research which had been started before the contract became effective. The work done in this area by other groups is being reviewed.

  11. Molecule-surface interaction processes of relevance to gas blanket type fusion device divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Snowdon, K.J. [Newcastle Univ. (United Kingdom). Dept. of Physics; Tawara, H.

    1997-01-01

    The mechanisms which may lead to the departure of molecular species from surfaces exposed to low energy (0.1-100 eV) particle or photon and electron irradiation are reviewed. Where possible, the charge and electronic state, angular, translational and internal energy distributions of the departing molecules are described and the physical origin of the nature of those distributions identified. The consequences, for the departing molecules, of certain material choices become apparent from such an analysis. Such information may help guide the choice of appropriate materials for plasma facing components of gas-blanket type divertors such as that recently proposed for the International Thermonuclear Experimental Reactor (ITER). (author). 71 refs.

  12. Experimental facility for studying MHD effects in liquid metal cooled blankets

    Science.gov (United States)

    Reed, C. B.; Picologlou, B. F.; Dauzvardis, P. V.

    The capabilities of a facility, brought into service to collect data on magnetohydrodynamic (MHD) effects, pertinent to liquid metal cooled fusion reactor blankets, are presented. The facility, design to extend significantly the existing data base on liquid metal MHD, employs eutectic NaK as the working fluid in a room temperature closed loop. The instrumentation system is capable of collecting detailed data on pressure, voltage, and velocity distributions at any axial position within the base of a 2 Tesla conventional magnet. The axial magnetic field distribution can be uniform or varying with either rapid or slow spatial variations.

  13. Extracellular Polymers in Granular Sludge from Different Upflow Anaerobic Sludge Blanket (UASB) Reactors

    DEFF Research Database (Denmark)

    Schmidt, Jens Ejbye; Ahring, Birgitte Kiær

    1994-01-01

    Thermal extraction was used to quantify extracellular polymers (ECP) in granules from anaerobic upflow reactors. The optimal time for extraction was determined as the time needed before the intracellular material gives a significant contribution to the extracted extracellular material due to cell...... of an upflow anaerobic sludge blanket reactor from a sugar-containing waste-water to a synthetic waste-water containing acetate, propionate and butyrate resulted in a decrease in both the protein and polysaccharide content and an increase in the lipid content of the extracellular material. Furthermore...

  14. 聚变-裂变混合堆高功率密度包层的设计研究%High Power Density Blanket Design Study for Fusion-fission Hybrid Reactors

    Institute of Scientific and Technical Information of China (English)

    黄锦华; 邓培智

    2001-01-01

    A conceptual design study of a high power density blanket was carried out. The blanket is cooled by high-pressure helium in tubes in the form of cooling panels. A great number of cooling panels is arranged inside the blanket yet maintaining a fairly simple configuration. The module is robust and fabricable. The concept of LiPb eutectic/transuranium oxide suspension is adopted. The neutronics design is performed giving a flattened power density distribution with the peak value of 70 W/cm3. Thermal analysis shows the design can satisfy technical requirements. Preliminary structural analysis has also been done.%进行了高功率密度包层的概念设计研究。包层冷却采用管道承压的氦气。虽然引入了众多的氦冷却管道,包层结构仍然比较简单、坚固并便于制造。采用了超铀氧化物颗粒悬浮在锂铅共熔体的方案,中子学计算给出峰值功率密度为70 MW*m-3,功率密度分布比较平坦。热工分析计算表明设计能满足技术要求。此外,进行了初步的结构分析计算。

  15. Structural Monitoring of Metro Infrastructure during Shield Tunneling Construction

    Directory of Open Access Journals (Sweden)

    L. Ran

    2014-01-01

    Full Text Available Shield tunneling construction of metro infrastructure will continuously disturb the soils. The ground surface will be subjected to uplift or subsidence due to the deep excavation and the extrusion and consolidation of the soils. Implementation of the simultaneous monitoring with the shield tunnel construction will provide an effective reference in controlling the shield driving, while how to design and implement a safe, economic, and effective structural monitoring system for metro infrastructure is of great importance and necessity. This paper presents the general architecture of the shield construction of metro tunnels as well as the procedure of the artificial ground freezing construction of the metro-tunnel cross-passages. The design principles for metro infrastructure monitoring of the shield tunnel intervals in the Hangzhou Metro Line 1 are introduced. The detailed monitoring items and the specified alarming indices for construction monitoring of the shield tunneling are addressed, and the measured settlement variations at different monitoring locations are also presented.

  16. Neutron shielding performance of water-extended polyester

    Energy Technology Data Exchange (ETDEWEB)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)

    2006-07-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  17. Long-term impacts of peatland restoration on the net ecosystem exchange (NEE) of blanket bogs in Northern Scotland.

    Science.gov (United States)

    Hambley, Graham; Hill, Timothy; Saunders, Matthew; Arn Teh, Yit

    2016-04-01

    Unmanaged peatlands represent an important long-term C sink and thus play an important part of the global C cycle. Despite covering only 12 % of the UK land area, peatlands are estimated to store approximately 20 times more carbon than the UK's forests, which cover 13% of the land area. The Flow Country of Northern Scotland is the largest area of contiguous blanket bog in the UK, and one of the biggest in Europe, covering an area in excess of 4000 km2 and plays a key role in mediating regional atmospheric exchanges of greenhouse gases (GHGs) such as carbon dioxide (CO2), and water vapour (H2O). However, these peatlands underwent significant afforestation in the 1980s, when over 670 km2 of blanket bog were drained and planted with Sitka spruce (Picea sitchensis) and Lodgepole pine (Pinus contorta). This resulted in modifications to hydrology, micro-topography, vegetation and soil properties all of which are known to influence the production, emission and sequestration of key GHGs. Since the late 1990s restoration work has been carried out to remove forest plantations and raise water tables, by drain blocking, to encourage the recolonisation of Sphagnum species and restore ecosystem functioning. Here, we report findings of NEE and its constituent fluxes, GPP and Reco, from a study investigating the impacts of restoration on C dynamics over a chronosequence of restored peatlands. The research explored the role of environmental variables and microtopography in modulating land-atmosphere exchanges, using a multi-scale sampling approach that incorporated eddy covariance measurements with dynamic flux chambers. Key age classes sampled included an undrained peatland; an older restored peatland (17 years old); and a more recently restored site (12 years old). The oldest restored site showed the strongest uptake of C, with an annual assimilation rate of 858 g C m-2 yr-1 compared to assimilation rates of 501g C m-2 yr-1 and 575g C m-2 yr-1 from the younger restored site and

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

    Science.gov (United States)

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

    2015-09-01

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

  19. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  20. The heterogeneous anti-radiation shield for spacecraft*

    Science.gov (United States)

    Telegin, S. V.; Draganyuk, O. N.

    2016-04-01

    The paper deals with modeling of elemental composition and properties of heterogeneous layers in multilayered shields to protect spacecraft onboard equipment from radiation emitted by the natural Earth’s radiation belt. This radiation causes malfunctioning of semiconductor elements in electronic equipment and may result in a failure of the spacecraft as a whole. We consider four different shield designs and compare them to the most conventional radiation-protective material for spacecraft - aluminum. Out of light and heavy chemical elements we chose the materials with high reaction cross sections and low density. The mass attenuation coefficient of boron- containing compounds is 20% higher than that of aluminum. Heterogeneous shields consist of three layers: a glass cloth, borated material, and nickel. With a protective shield containing heavy metal the output bremsstrahlung can be reduced. The amount of gamma rays that succeed to penetrate the shield is 4 times less compared to aluminum. The shields under study have the thicknesses of 5.95 and 6.2 mm. A comparative analysis of homogeneous and multilayered protective coatings of the same chemical composition has been performed. A heterogeneous protective shield has been found to be advantageous in weight and shielding properties over its homogeneous counterparts and aluminum. The dose characteristics and transmittance were calculated by the Monte Carlo method. The results of our study lead us to conclude that a three-layer boron carbide shield provides the most effective protection from radiation. This shield ensures twice as low absorbed dose and 4 times less the number of penetrated gamma-ray photons compared to its aluminum analogue. Moreover, a heterogeneous shield will have a weight 10% lighter than aluminum, with the same attenuation coefficient of the electron flux. Such heterogeneous shields can be used to protect spacecraft launched to geostationary orbit. Furthermore, a protective boron-containing and

  1. Optimal Shielding for Minimum Materials Cost of Mass

    Energy Technology Data Exchange (ETDEWEB)

    Woolley, Robert D. [PPPL

    2014-08-01

    Material costs dominate some shielding design problems. This is certainly the case for manned nuclear power space applications for which shielding is essential and the cost of launching by rocket from earth is high. In such situations or in those where shielding volume or mass is constrained, it is important to optimize the design. Although trial and error synthesis methods may succeed a more systematic approach is warranted. Design automation may also potentially reduce engineering costs.

  2. Radiation Shielding at High-Energy Electron and Proton Accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Rokni, Sayed H.; /SLAC; Cossairt, J.Donald; /Fermilab; Liu, James C.; /SLAC

    2007-12-10

    The goal of accelerator shielding design is to protect the workers, general public, and the environment against unnecessary prompt radiation from accelerator operations. Additionally, shielding at accelerators may also be used to reduce the unwanted background in experimental detectors, to protect equipment against radiation damage, and to protect workers from potential exposure to the induced radioactivity in the machine components. The shielding design for prompt radiation hazards is the main subject of this chapter.

  3. Nutrient Shielding in Clusters of Cells

    CERN Document Server

    Lavrentovich, Maxim O; Nelson, David R

    2013-01-01

    Cellular nutrient consumption is influenced by both the nutrient uptake kinetics of an individual cell and the cells' spatial arrangement. Large cell clusters or colonies have inhibited growth at the cluster's center due to the shielding of nutrients by the cells closer to the surface. We develop an effective medium theory that predicts a thickness $\\ell$ of the outer shell of cells in the cluster that receives enough nutrient to grow. The cells are treated as partially absorbing identical spherical nutrient sinks, and we identify a dimensionless parameter $\

  4. Interstitial rotating shield brachytherapy for prostate cancer

    Energy Technology Data Exchange (ETDEWEB)

    Adams, Quentin E., E-mail: quentin-adams@uiowa.edu; Xu, Jinghzu; Breitbach, Elizabeth K.; Li, Xing; Rockey, William R.; Kim, Yusung; Wu, Xiaodong; Flynn, Ryan T. [Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States); Enger, Shirin A. [Medical Physics Unit, McGill University, 1650 Cedar Ave, Montreal, Quebec H3G 1A4 (Canada)

    2014-05-15

    Purpose: To present a novel needle, catheter, and radiation source system for interstitial rotating shield brachytherapy (I-RSBT) of the prostate. I-RSBT is a promising technique for reducing urethra, rectum, and bladder dose relative to conventional interstitial high-dose-rate brachytherapy (HDR-BT). Methods: A wire-mounted 62 GBq{sup 153}Gd source is proposed with an encapsulated diameter of 0.59 mm, active diameter of 0.44 mm, and active length of 10 mm. A concept model I-RSBT needle/catheter pair was constructed using concentric 50 and 75 μm thick nickel-titanium alloy (nitinol) tubes. The needle is 16-gauge (1.651 mm) in outer diameter and the catheter contains a 535 μm thick platinum shield. I-RSBT and conventional HDR-BT treatment plans for a prostate cancer patient were generated based on Monte Carlo dose calculations. In order to minimize urethral dose, urethral dose gradient volumes within 0–5 mm of the urethra surface were allowed to receive doses less than the prescribed dose of 100%. Results: The platinum shield reduced the dose rate on the shielded side of the source at 1 cm off-axis to 6.4% of the dose rate on the unshielded side. For the case considered, for the same minimum dose to the hottest 98% of the clinical target volume (D{sub 98%}), I-RSBT reduced urethral D{sub 0.1cc} below that of conventional HDR-BT by 29%, 33%, 38%, and 44% for urethral dose gradient volumes within 0, 1, 3, and 5 mm of the urethra surface, respectively. Percentages are expressed relative to the prescription dose of 100%. For the case considered, for the same urethral dose gradient volumes, rectum D{sub 1cc} was reduced by 7%, 6%, 6%, and 6%, respectively, and bladder D{sub 1cc} was reduced by 4%, 5%, 5%, and 6%, respectively. Treatment time to deliver 20 Gy with I-RSBT was 154 min with ten 62 GBq {sup 153}Gd sources. Conclusions: For the case considered, the proposed{sup 153}Gd-based I-RSBT system has the potential to lower the urethral dose relative to HDR-BT by 29

  5. Space Shuttle Orbiter AFT heat shield seal

    Science.gov (United States)

    Walkover, L. J.

    1979-01-01

    The evolution of the orbiter aft heat shield seal (AHSS) design, which involved advancing mechanical seal technology in severe thermal environment is discussed. The baseline design, various improvements for engine access, and technical problem solution are presented. It is a structure and mechanism at the three main propulsion system (MPS) engine interfaces to the aft compartment structure. Access to each MPS engine requires disassembly and removal of the AHSS. Each AHSS accommodates the engine movement, is exposed to an extremely high temperature environment, and is part of the venting control of the aft compartment.

  6. Self-shielding clumps in starburst clusters

    CERN Document Server

    Palouš, Jan; Ehlerová, Soňa; Tenorio-Tagle, Guillermo

    2016-01-01

    Young and massive star clusters above a critical mass form thermally unstable clumps reducing locally the temperature and pressure of the hot 10$^{7}$~K cluster wind. The matter reinserted by stars, and mass loaded in interactions with pristine gas and from evaporating circumstellar disks, accumulate on clumps that are ionized with photons produced by massive stars. We discuss if they may become self-shielded when they reach the central part of the cluster, or even before it, during their free fall to the cluster center. Here we explore the importance of heating efficiency of stellar winds.

  7. Nanoscale microwave microscopy using shielded cantilever probes

    KAUST Repository

    Lai, Keji

    2011-04-21

    Quantitative dielectric and conductivity mapping in the nanoscale is highly desirable for many research disciplines, but difficult to achieve through conventional transport or established microscopy techniques. Taking advantage of the micro-fabrication technology, we have developed cantilever-based near-field microwave probes with shielded structures. Sensitive microwave electronics and finite-element analysis modeling are also utilized for quantitative electrical imaging. The system is fully compatible with atomic force microscope platforms for convenient operation and easy integration of other modes and functions. The microscope is ideal for interdisciplinary research, with demonstrated examples in nano electronics, physics, material science, and biology.

  8. Radiation-Shielding Polymer/Soil Composites

    Science.gov (United States)

    Sen, Subhayu

    2007-01-01

    It has been proposed to fabricate polymer/ soil composites primarily from extraterrestrial resources, using relatively low-energy processes, with the original intended application being that habitat structures constructed from such composites would have sufficient structural integrity and also provide adequate radiation shielding for humans and sensitive electronic equipment against the radiation environment on the Moon and Mars. The proposal is a response to the fact that it would be much less expensive to fabricate such structures in situ as opposed to transporting them from Earth.

  9. Cerrobend shielding stents for buccal carcinoma patients

    Directory of Open Access Journals (Sweden)

    Karma Yangchen

    2016-01-01

    Full Text Available Buccal carcinoma is one of the most common oral malignant neoplasms, especially in the South Asian region. Radiotherapy, which plays a significant role in the treatment of this carcinoma, has severe adverse effects. Different types of prosthesis may be constructed to protect healthy tissues from the adverse effects of treatment and concentrate radiation in the region of the tumor mass. However, the technique for fabrication of shielding stent with Lipowitz's alloy (cerrobend/Wood's alloy has not been well documented. This article describes detailed technique for fabrication of such a stent for unilateral buccal carcinoma patients to spare the unaffected oral cavity from potential harmful effects associated with radiotherapy.

  10. Gamma shielding properties of Tamoxifen drug

    Science.gov (United States)

    Kanberoglu, Gulsah Saydan; Oto, Berna; Gulebaglan, Sinem Erden

    2017-02-01

    Tamoxifen (MW=371 g/mol) is an endocrine therapeutic drug widely prescribed as chemopreventive in women to prevent and to treat all stages of breast cancer. It is also being studied for other types of cancer. In this study, we have calculated some gamma shielding parameters such as mass attenuation coefficient (μρ), effective atomic number (Zeff) and electron density (Nel) for Tamoxifen drug. The values of μρ were calculated using WinXCom computer program and then the values of Zeff and Nel were derived using μρ values in the wide energy range (1 keV - 100 GeV).

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

    Science.gov (United States)

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

    2015-01-01

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

  12. InfuShield: a shielded enclosure for administering therapeutic radioisotope treatments using standard syringe pumps.

    Science.gov (United States)

    Rushforth, Dominic P; Pratt, Brenda E; Chittenden, Sarah J; Murray, Iain S; Causer, Louise; Grey, Matthew J; Gear, Jonathan I; Du, Yong; Flux, Glenn D

    2017-03-01

    The administration of radionuclide therapies presents significant radiation protection challenges. The aim of this work was to develop a delivery system for intravenous radioisotope therapies to substantially moderate radiation exposures to staff and operators. A novel device (InfuShield) was designed and tested before being used clinically. The device consists of a shielded enclosure which contains the therapeutic activity and, through the hydraulic action of back-to-back syringes, allows the activity to be administered using a syringe pump external to the enclosure. This enables full access to the pump controls while simultaneously reducing dose to the operator. The system is suitable for use with all commercially available syringe pumps and does not require specific consumables, maximising both the flexibility and economy of the system. Dose rate measurements showed that at key stages in an I mIBG treatment procedure, InfuShield can reduce dose to operators by several orders of magnitude. Tests using typical syringes and infusion speeds show no significant alteration in administered flow rates (maximum of 1.2%). The InfuShield system provides a simple, safe and low cost method of radioisotope administration.

  13. Gravity Scaling of a Power Reactor Water Shield

    Science.gov (United States)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  14. Neutron shielding for a {sup 252} Cf source

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M. [Unidades Academicas de Estudios Nucleares e Ingenieria Electrica, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Eduardo Gallego, Alfredo Lorente [Depto. de Ingenieria Nuclear, ETS Ingenieros Industriales, Universidad Politecnica de Madrid, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. e-mail: fermineutron@yahoo.com

    2006-07-01

    To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source. During calculations a detailed model for the {sup 252}Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare {sup 252}Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)

  15. Polyolefin-Nanocrystal Composites for Radiation Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — EIC Laboratories Inc. is proposing a lightweight multifunctional polymer/nanoparticle composite for radiation shielding during long-duration lunar missions. Isolated...

  16. Movable radiation shields for the CLEO II silicon vertex detector

    Energy Technology Data Exchange (ETDEWEB)

    Dumas, D.J.; Ward, C.W.; Alexander, J.; Cherwinka, J.; Henderson, S. [Cornell Univ., Ithaca, NY (United States); Cinabro, D. [Harvard University, Cambridge, MA 02138 (United States); Fast, J. [Purdue University, Lafayette, IN 47907 (United States); Morrison, R. [University of California at Santa Barbara, Santa Barbara, CA 93106 (United States); O`Neill, M. [CRPP, Carleton University, Ottawa, Ont. (Canada)

    1998-02-11

    Two movable tungsten radiation shields were installed on the beam pipe during the upgrade of the CLEO II detector, operating at the Cornell electron storage ring (CESR). This upgrade included the installation of a silicon vertex detector (SVX) and the purpose of the shields is to protect the SVX readout electronics from synchrotron radiation produced during injection and non-high-energy physics operation of CESR. Shield motion is controlled remotely by cables, keeping the associated motors and controls outside the detection volume. We discuss the design and performance of the radiation shields and the associated control system. (orig.). 8 refs.

  17. Graphene shield enhanced photocathodes and methods for making the same

    Science.gov (United States)

    Moody, Nathan Andrew

    2014-09-02

    Disclosed are graphene shield enhanced photocathodes, such as high QE photocathodes. In certain embodiments, a monolayer graphene shield membrane ruggedizes a high quantum efficiency photoemission electron source by protecting a photosensitive film of the photocathode, extending operational lifetime and simplifying its integration in practical electron sources. In certain embodiments of the disclosed graphene shield enhanced photocathodes, the graphene serves as a transparent shield that does not inhibit photon or electron transmission but isolates the photosensitive film of the photocathode from reactive gas species, preventing contamination and yielding longer lifetime.

  18. Mercuric Iodide Anticoincidence Shield for Gamma-Ray Spectrometer Project

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to utilize a new detector material, polycrystalline mercuric iodide, for background suppression by active anticoincidence shielding in gamma-ray...

  19. Mercuric Iodide Anticoincidence Shield for Gamma-Ray Spectrometer Project

    Data.gov (United States)

    National Aeronautics and Space Administration — We utilize a new detector material, polycrystalline mercuric iodide, for background suppression by active anticoincidence shielding in gamma-ray spectrometers. Two...

  20. Crack tip shielding and anti-shielding effects of parallel cracks for a superconductor slab under an electromagnetic force

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Zhi Wen; Zhou, You He [Ministry of Education, Singapore (China); Lee, Kang Yong [Yonsei University, Seoul (Korea, Republic of)

    2012-02-15

    In this letter, the shielding or anti-shielding effect is firstly applied to obtain the behavior of two parallel cracks in a two-dimensional type-II superconducting under electromagnetic force. Fracture analysis is performed by the finite element method and the magnetic behavior of superconductor is described by the critical state Bean model. The stress intensity factors at the crack tips can be obtained and discussed for decreasing field after zero-field cooling. The shielding or anti-shielding effect at the crack tips depend on the distance between two parallel cracks and the crack length. The results indicate that the shielding effects of the two parallel cracks increase when the distance between the two parallel cracks decreases. It can be also obtained that the superconductors with shorter cracks has more remarkable shielding effect than those with longer cracks.

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

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-01-01

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

  3. Octalithium plumbate as breeding blanket ceramic: Neutronic performances, synthesis and partial characterization

    Energy Technology Data Exchange (ETDEWEB)

    Colominas, S., E-mail: sergi.colominas@iqs.es [Universitat Ramon Llull, ETS Institut Quimic de Sarria, Electrochemical Methods Laboratory - Analytical Chemistry Department, Via Augusta, 390, 08017 Barcelona (Spain); Palermo, I., E-mail: iole.palermo@ciemat.es [CIEMAT, Av. Complutense 22, E-28040 Madrid (Spain); Abella, J., E-mail: jordi.abella@iqs.es [Universitat Ramon Llull, ETS Institut Quimic de Sarria, Electrochemical Methods Laboratory - Analytical Chemistry Department, Via Augusta, 390, 08017 Barcelona (Spain); Gomez-Ros, J.M., E-mail: jm.gomezros@ciemat.es [CIEMAT, Av. Complutense 22, E-28040 Madrid (Spain); Sanz, J., E-mail: jsanz@ind.uned.es [UNED, Department of Nuclear Energy, c./Juan del Rosal 12, E-28040 Madrid (Spain); Sedano, L., E-mail: luis.sedano@ciemat.es [CIEMAT, Av. Complutense 22, E-28040 Madrid (Spain)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Definition of a suitable configuration for the Li{sub 8}PbO{sub 6} breeding blanket design. Black-Right-Pointing-Pointer Demonstration of the feasibility of Li{sub 8}PbO{sub 6} as a breeding material. Black-Right-Pointing-Pointer Synthesis optimization in the Li{sub 8}PbO{sub 6} production. Black-Right-Pointing-Pointer Characterization of Li{sub 8}PbO{sub 6} by X-ray phase analysis is discussed. - Abstract: A neutronic assessment of the performances of a helium-cooled Li{sub 8}PbO{sub 6} breeding blanket (BB) for the conceptual design of a DEMO fusion reactor is given. Different BB configurations have been considered in order to minimize the amount of beryllium required for neutron multiplication, including the use of graphite as reflector material. The calculated neutronic responses: tritium breeding ratio (TBR), power deposition in TF coils and power amplification factor, indicate the feasibility of Li{sub 8}PbO{sub 6} as breeding material. Furthermore, the synthesis and characterization of Li{sub 8}PbO{sub 6} by X-ray phase analysis are also discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-20

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

  5. Stellar model atmospheres with magnetic line blanketing. II. Introduction of polarized radiative transfer

    CERN Document Server

    Khan, S A

    2006-01-01

    The technique of model atmosphere calculation for magnetic Ap and Bp stars with polarized radiative transfer and magnetic line blanketing is presented. A grid of model atmospheres of A and B stars are computed. These calculations are based on direct treatment of the opacities due to the bound-bound transitions that ensures an accurate and detailed description of the line absorption and anomalous Zeeman splitting. The set of model atmospheres was calculated for the field strengths between 1 and 40 kG. The high-resolution energy distribution, photometric colors and the hydrogen Balmer line profiles are computed for magnetic stars with different metallicities and are compared to those of non-magnetic reference models and to the previous paper of this series. The results of modelling confirmed the main outcomes of the previous study: energy redistribution from UV to the visual region and flux depression at 5200A. However, we found that effects of enhanced line blanketing when transfer for polarized radiation take...

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

    Energy Technology Data Exchange (ETDEWEB)

    Reed, C.B.; Haglund, R.C.; Miller, M.E. [and others

    1995-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

  8. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 2: with Beam Shutdown Only

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report. This report documents the results of simulations of a Loss-of-Flow Accident (LOFA) where power is lost to all of the pumps that circulate water in the blanket region, the accelerator beam is shut off and neither the residual heat removal nor cavity flood systems operate.

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

    Directory of Open Access Journals (Sweden)

    A. Kirichek

    2013-12-01

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

  10. Geomorphological control of water tables in a blanket peat landscape: implications for carbon cycling

    Science.gov (United States)

    Allott, Tim; Evans, Martin; Lindsay, John; Agnew, Clive; Freer, Jim

    2010-05-01

    Water tables are an important control on carbon cycling and rates of carbon sequestration in peatland systems, and water table depth is therefore a key parameter in carbon models for blanket peat systems. Although there is a wide literature on blanket peat hydrology, including studies which specifically evaluate water table conditions, detailed data on water table behaviour and variability at the landscape scale are sparse. In particular, many British blanket peats are affected by gully erosion and this has been generally assumed to influence water table conditions. However, there has been limited evaluation of this geomomorphological control on peatland water tables. This paper presents results from a project which evaluated water table conditions in the blanket peatlands of the Peak District National Park, UK. A key aim was to quantify the impact of gully erosion on peatland water tables. A detailed programme of water table monitoring was undertaken during 2008/09, involving regular measurements of water table depth in over 530 dipwells at 19 sites across the 47 km2 peatland landscape of the Kinder Scout / Bleaklow area. This included a campaign of regular, simultaneous water table measurements from clusters of dipwells at the main sites, supplemented by continuous (hourly) water table monitoring in selected dipwells. It also included studies to evaluate within-site variation in water table conditions and local water table drawdown effects associated with gully erosion. Results indicate that gully erosion causes water table drawdown through two distinct processes. The first is local water table drawdown immediately adjacent to erosion gullies. This effect is restricted to a zone within 2 m of gully edges, and water tables within the gully edge drawdown zone are approximately 200 mm lower than in the adjacent peatland. The second effect is a more general water table lowering at eroded sites, with median water table depths at heavily eroded sites up to 300 mm lower

  11. SHIELD: Observations of Three Candidate Interacting Systems

    Science.gov (United States)

    Ruvolo, Elizabeth; Miazzo, Masao; Cannon, John M.; McNichols, Andrew; Teich, Yaron; Adams, Elizabeth A.; Giovanelli, Riccardo; Haynes, Martha P.; McQuinn, Kristen B.; Salzer, John Joseph; Skillman, Evan D.; Dolphin, Andrew E.; Elson, Edward C.; Haurberg, Nathalie C.; Huang, Shan; Janowiecki, Steven; Jozsa, Gyula; Leisman, Luke; Ott, Juergen; Papastergis, Emmanouil; Rhode, Katherine L.; Saintonge, Amelie; Van Sistine, Angela; Warren, Steven R.

    2017-01-01

    Abstract:The “Survey of HI in Extremely Low-mass Dwarfs” (SHIELD) is a multiwavelength study of local volume low-mass galaxies. Using the now-complete Arecibo Legacy Fast ALFA (ALFALFA) source catalog, 82 systems are identified that meet distance, line width, and HI flux criteria for being gas-rich, low-mass galaxies. These systems harbor neutral gas reservoirs smaller than 3x10^7 M_sun, thus populating the faint end of the HI mass function with statistical confidence for the first time. In a companion poster, we present new Karl G. Jansky Very Large Array D-configuration HI spectral line observations of 32 previously unobserved galaxies. Three galaxies in that study have been discovered to lie in close angular proximity to more massive galaxies. Here we present VLA HI imaging of these candidate interacting systems. We compare the neutral gas morphology and kinematics with optical images from SDSS. We discuss the frequency of low-mass galaxies undergoing tidal interaction in the complete SHIELD sample.Support for this work was provided by NSF grant 1211683 to JMC at Macalester College.

  12. Technique and results of cartilage shield tympanoplasty

    Directory of Open Access Journals (Sweden)

    Sohil I Vadiya

    2014-01-01

    Full Text Available Aim: Use of cartilage for repair of tympanic membrane is recommended by many otologists. The current study aims at evaluating results of cartilage shield tympanoplasty in terms of graft take up and hearing outcomes. Material and Methods: In the current study, cartilage shield tympanoplasty(CST is used in ears with high risk perforations of the tympanic membrane. A total of 40 ears were selected where type I CST was done in 30 ears and type III CST was done in 10 ears. Results: An average of 37.08 dB air bone gap(ABG was present in pre operative time and an average of 19.15 dB of ABG was observed at 6 months after the surgery with hearing gain of 17.28 dB on average was observed. Graft take up rate of 97.5% was observed. The technique is modified to make it easier and to minimize chances of lateralization of graft. Conclusion: The hearing results of this technique are comparable to other methods of tympanic membrane repair.

  13. Discussion on variance reduction technique for shielding

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    As the task of the engineering design activity of the international thermonuclear fusion experimental reactor (ITER), on 316 type stainless steel (SS316) and the compound system of SS316 and water, the shielding experiment using the D-T neutron source of FNS in Japan Atomic Energy Research Institute has been carried out. However, in these analyses, enormous working time and computing time were required for determining the Weight Window parameter. Limitation or complication was felt when the variance reduction by Weight Window method of MCNP code was carried out. For the purpose of avoiding this difficulty, investigation was performed on the effectiveness of the variance reduction by cell importance method. The conditions of calculation in all cases are shown. As the results, the distribution of fractional standard deviation (FSD) related to neutrons and gamma-ray flux in the direction of shield depth is reported. There is the optimal importance change, and when importance was increased at the same rate as that of the attenuation of neutron or gamma-ray flux, the optimal variance reduction can be done. (K.I.)

  14. The AA disappearing under concrete shielding

    CERN Multimedia

    1982-01-01

    When the AA started up in July 1980, the machine stood freely in its hall, providing visitors with a view through the large window in the AA Control Room. The target area, in which the high-intensity 26 GeV/c proton beam from the PS hit the production target, was heavily shielded, not only towards the outside but also towards the AA-Hall. However, electrons and pions emanating from the target with the same momentum as the antiprotons, but much more numerous, accompanied these through the injection line into the AA ring. The pions decayed with a half-time corresponding to approximately a revolution period (540 ns), whereas the electrons lost energy through synchrotron radiation and ended up on the vacuum chamber wall. Electrons and pions produced the dominant component of the radiation level in the hall and the control room. With operation times far exceeding original expectations, the AA had to be buried under concrete shielding in order to reduce the radiation level by an order of magnitude.

  15. MicroShield/ISOCS gamma modeling comparison.

    Energy Technology Data Exchange (ETDEWEB)

    Sansone, Kenneth R

    2013-08-01

    Quantitative radiological analysis attempts to determine the quantity of activity or concentration of specific radionuclide(s) in a sample. Based upon the certified standards that are used to calibrate gamma spectral detectors, geometric similarities between sample shape and the calibration standards determine if the analysis results developed are qualitative or quantitative. A sample analyzed that does not mimic a calibrated sample geometry must be reported as a non-standard geometry and thus the results are considered qualitative and not quantitative. MicroShieldR or ISOCSR calibration software can be used to model non-standard geometric sample shapes in an effort to obtain a quantitative analytical result. MicroShieldR and Canberras ISOCSR software contain several geometry templates that can provide accurate quantitative modeling for a variety of sample configurations. Included in the software are computational algorithms that are used to develop and calculate energy efficiency values for the modeled sample geometry which can then be used with conventional analysis methodology to calculate the result. The response of the analytical method and the sensitivity of the mechanical and electronic equipment to the radionuclide of interest must be calibrated, or standardized, using a calibrated radiological source that contains a known and certified amount of activity.

  16. A superconducting shield to protect astronauts

    CERN Multimedia

    Antonella Del Rosso

    2015-01-01

    The CERN Superconductors team in the Technology department is involved in the European Space Radiation Superconducting Shield (SR2S) project, which aims to demonstrate the feasibility of using superconducting magnetic shielding technology to protect astronauts from cosmic radiation in the space environment. The material that will be used in the superconductor coils on which the project is working is magnesium diboride (MgB2), the same type of conductor developed in the form of wire for CERN for the LHC High Luminosity Cold Powering project.   Image: K. Anthony/CERN. Back in April 2014, the CERN Superconductors team announced a world-record current in an electrical transmission line using cables made of the MgB2 superconductor. This result proved that the technology could be used in the form of wire and could be a viable solution for both electrical transmission for accelerator technology and long-distance power transportation. Now, the MgB2 superconductor has found another application: it wi...

  17. Concrete enclosure to shield a neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Villagrana M, L. E.; Rivera P, E.; De Leon M, H. A.; Soto B, T. G.; Hernandez D, V. M.; Vega C, H. R., E-mail: emmanuelvillagrana@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Apdo. Postal 336, 98000 Zacatecas (Mexico)

    2012-10-15

    In the aim to design a shielding for a {sup 239}PuBe isotopic neutron source several Monte Carlo calculations were carried out using MCNP5 code. First, a point-like source was modeled in vacuum and the neutron spectrum and the ambient dose equivalent were calculated at several distances ranging from 5 up to 150 cm, these calculations were repeated including air, and a 1 x 1 x 1 m{sup 3} enclosure that was shielded with 5, 15, 20, 25, 30, 50 and 80 cm-thick Portland type concrete walls. At all the points located inside the enclosure neutron spectra from 10{sup -8} up 0.5 MeV were the same regardless the distance from the source showing the room-return effect, for energies larger than 0.5 MeV neutron spectra are diminished as the distance increases. Outside the enclosure it was noticed that neutron spectra becomes -softer- as the concrete thickness increases due to reduction of mean neutron energy. With the ambient dose values the attenuation curve in terms of concrete thickness was calculated. (Author)

  18. Optimization of Shielded Scintillator for Neutron Detection

    Science.gov (United States)

    Belancourt, Patrick; Morrison, John; Akli, Kramer; Freeman, Richard; High Energy Density Physics Team

    2011-10-01

    The High Energy Density Physics group is interested in the basic science of creating a neutron and gamma ray source. The neutrons and gamma rays are produced by accelerating ions via a laser into a target and creating fusion neutrons and gamma rays. A scintillator and photomultiplier tube will be used to detect these neutrons. Neutrons and photons produce ionizing radiation in the scintillator which then activates metastable states. These metastable states have both short and long decay rates. The initial photon count is orders of magnitude higher than the neutron count and poses problems for accurately detecting the neutrons due to the long decay state that is activated by the photons. The effects of adding lead shielding on the temporal response and signal level of the neutron detector will be studied in an effort to minimize the photon count without significant reduction to the temporal resolution of the detector. MCNP5 will be used to find the temporal response and energy deposition into the scintillator by adding lead shielding. Results from the simulations will be shown. Optimization of our scintillator neutron detection system is needed to resolve the neutron energies and neutron count of a novel neutron and gamma ray source.

  19. Enhanced radiation shielding with galena concrete

    Directory of Open Access Journals (Sweden)

    Hadad Kamal

    2015-01-01

    Full Text Available A new concrete, containing galena mineral, with enhanced shielding properties for gamma sources is developed. To achieve optimized shielding properties, ten types of galena concrete containing different mixing ratios and a reference normal concrete of 2300 kg/m3 density are studied experimentally and numerically using Monte Carlo and XCOM codes. For building galena concrete, in addition to the main composition, micro-silica and water, galena mineral (containing lead were used. The built samples have high density of 4470 kg/m3 to 5623 kg/m3 and compressive strength of 628 kg/m2 to 685 kg/m2. The half and tenth value layers (half value layer and tenth value layers for the galena concrete, when irradiated with 137Cs gamma source, were found to be 1.45 cm and 4.94 cm, respectively. When irradiated with 60Co gamma source, half value layer was measured to be 2.42 cm. The computation modeling by FLUKA and XCOM shows a good agreement between experimental and computational results.

  20. Large scale mechanical metamaterials as seismic shields

    Science.gov (United States)

    Miniaci, Marco; Krushynska, Anastasiia; Bosia, Federico; Pugno, Nicola M.

    2016-08-01

    Earthquakes represent one of the most catastrophic natural events affecting mankind. At present, a universally accepted risk mitigation strategy for seismic events remains to be proposed. Most approaches are based on vibration isolation of structures rather than on the remote shielding of incoming waves. In this work, we propose a novel approach to the problem and discuss the feasibility of a passive isolation strategy for seismic waves based on large-scale mechanical metamaterials, including for the first time numerical analysis of both surface and guided waves, soil dissipation effects, and adopting a full 3D simulations. The study focuses on realistic structures that can be effective in frequency ranges of interest for seismic waves, and optimal design criteria are provided, exploring different metamaterial configurations, combining phononic crystals and locally resonant structures and different ranges of mechanical properties. Dispersion analysis and full-scale 3D transient wave transmission simulations are carried out on finite size systems to assess the seismic wave amplitude attenuation in realistic conditions. Results reveal that both surface and bulk seismic waves can be considerably attenuated, making this strategy viable for the protection of civil structures against seismic risk. The proposed remote shielding approach could open up new perspectives in the field of seismology and in related areas of low-frequency vibration damping or blast protection.