WorldWideScience

Sample records for internal blanket variations

  1. First wall fusion blanket temperature variation - slab geometry

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1978-01-01

    The first wall of a fusion blanket is approximated by a slab, with the surface facing the plasma subjected to an applied heat flux, while the rear surface is convectively cooled. The relevant parameters affecting the heat transfer during the early phases of heating as well as for large times are established. Analytical solutions for the temperature variation with time and space are derived. Numerical calculations for an aluminum and stainless steel slab are performed for a wall loading of 1 MW(th)/m 2 . Both helium and water cooling are considered. (Auth.)

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

    International Nuclear Information System (INIS)

    Tanigawa, Hisashi; Enoeda, Mikio

    2010-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hisashi; Enoeda, Mikio [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki (Japan)

    2010-03-15

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

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

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2004-07-01

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

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

    International Nuclear Information System (INIS)

    Noda, Kenji

    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 2 TiO 3 , tritium release behavior of Li 2 TiO 3 and Li 2 ZrO 3 including tritium diffusion, modeling of tritium release from Li 2 ZrO 3 in ITER condition, helium release behavior from Li 2 O, results of tritium release irradiation tests of Li 4 SiO 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.)

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

    International Nuclear Information System (INIS)

    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

  7. 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.)

  8. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

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

  9. An experimental study of the heterogeneous LMFBR core using FCA assemblies with axial internal blanket

    International Nuclear Information System (INIS)

    Nakano, M.; Iijima, S.; Shirakata, K.; Hirota, J.

    1980-01-01

    To investigate physics properties of the heterogeneous LMFBR core and to examine the reliability of the current data and method for heterogeneous core configuration, an experimental study has been made on FCA VII-3 assemblies which have an internal blanket (IB) at midplane of the cylindrical core. Systematic experiments were carried out on the heterogeneous cores whose IBs were different in composition and thickness. A homogeneous core was also built to compare the results with those obtained on the heterogeneous cores. The sodium-void worth is not sensitive to the composition of IB. The positive void worth in the core of the 40 cm IB is lowered by about 40% compared with that in the homogeneous core. The analysis was made using the JAERI-Fast Set Version II and the diffusion code CITATION. Directional diffusion coefficients were used to take account of the axial streaming. To evaluate transport effects, the S 4 calculation was made. Comparison between the calculated and experimental results reveals the following: ksub(eff) and Pu worth in the core are not well predicted for the heterogeneous core, although they are represented satisfactorily for the homogeneous core. Reaction rates sensitive to the low-energy neutron are underestimated in the IB when they are normalized in the core. Sodium-void worths are fairly well predicted. However, the positive void worth in the heterogeneous core is underestimated, while that in the homogeneous core is overestimated. (author)

  10. Fusion fuel blanket technology

    International Nuclear Information System (INIS)

    Hastings, I.J.; Gierszewski, P.

    1987-05-01

    The fusion blanket surrounds the burning hydrogen core of a fusion reactor. It is in this blanket that most of the energy released by the nuclear fusion of deuterium-tritium is converted into useful product, and where tritium fuel is produced to enable further operation of the reactor. As fusion research turns from present short-pulse physics experiments to long-burn engineering tests in the 1990's, energy removal and tritium production capabilities become important. This technology will involve new materials, conditions and processes with applications both to fusion and beyond. In this paper, we introduce features of proposed blanket designs and update and status of international research. In focusing on the Canadian blanket technology program, we discuss the aqueous lithium salt blanket concept, and the in-reactor tritium recovery test program

  11. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  12. Practice Variation in Public Sector Internal Auditing

    DEFF Research Database (Denmark)

    Arena, Marika; Jeppesen, Kim Klarskov

    2016-01-01

    This article aims to analyze the rise of practice variations in public sector internal auditing (IA), giving special attention to the role of agents’ embeddedness in multiple institutional arrangements. IA's trends of development and the characteristics of the public sector context, in fact, make...... types of IA developed in three case settings, shaped by the agents’ embeddedness in different institutional fields. This article provides a more comprehensive approach to the study of IA adoption and development in public sector organizations than previous literature, and it highlights the relevance...... of the interplay between actors’ contemporary embeddedness in professional systems and the focal social system as a relevant source of practice variation. In this respect, the case of IA can contribute to previous studies of practice variation in the field of management accounting, shedding some light on the types...

  13. Disruption problematics in segmented blanket concepts

    International Nuclear Information System (INIS)

    Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

    1994-01-01

    In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

  14. Magnetoconvection in HCLL blankets

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Buehler, L.

    2014-01-01

    In the present work we consider magneto-convective flows in one of the proposed European liquid metal blankets that will be tested in the experimental fusion reactor ITER. Here the PbLi alloy is used as breeder material and helium as coolant. In order to finalize the design of the helium cooled lead lithium (HCLL) blanket, studies are still required to fully understand the behavior of the electrically conducting breeder under the influence of the intense magnetic field that confines the fusion plasma and in case of non-uniform thermal conditions. Liquid metal HCLL blanket flows are expected to be mainly driven by buoyancy forces caused by non-isothermal operating conditions due to neutron volumetric heating and cooling of walls, since only a weak forced ow is foreseen for tritium extraction in external ancillary systems. Buoyancy can therefore become very important and modify the velocity distribution and related heat transfer performance of the blanket. The present numerical study aims at clarifying the influence of electromagnetic and thermal coupling of neighboring fluid domains on magneto-convective flows in geometries relevant for the HCLL blanket concept. According to the last design review two internal cooling plates subdivide the fluid domain into three slender flow regions, which are thermally and electrically coupled through common walls. First a uniform volumetric heat source is considered to identify the basic convective patterns that establish in the liquid metal. Results are then compared with those obtained by applying a realistic radial distribution of the power density as obtained from a neutronic analysis. Velocity and temperature distributions are discussed for various volumetric heat sources and magnetic field strengths.

  15. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

    General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

  16. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    Proust, E.; Giancarli, L.

    1992-01-01

    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 investigated within the framework of the European Test-Blanket Development Programme

  17. LMFBR blanket physics project progress report No. 4

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Lanning, D.D.; Kaplan, I.; Supple, A.T.

    1973-01-01

    During the period covered by the report, July 1, 1972, through June 30, 1973, work was devoted to completion of experimental measurements and data analysis on Blanket Mockup No. 3, a graphite-reflected blanket, and to initiation of experimental work on Blanket Mockup No. 4, a steel-reflected assembly designed to mock up a demonstration plant blanket. Work was also carried out on the analysis of a number of advanced blanket concepts, including the use of high-albedo reflectors, the use of thorium in place of uranium in the blanket region, and the ''parfait'' or completely internal blanket concept. Finally, methods development work was initiated to develop the capability for making gamma heating measurements in the blanket mockups. (U.S.)

  18. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1995-09-01

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

  19. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1996-01-01

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

  20. Cassette blanket and vacuum building: key elements in fusion reactor maintenance

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

    The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

  1. International Variation in Asthma and Bronchiolitis Guidelines.

    Science.gov (United States)

    Bakel, Leigh Anne; Hamid, Jemila; Ewusie, Joycelyne; Liu, Kai; Mussa, Joseph; Straus, Sharon; Parkin, Patricia; Cohen, Eyal

    2017-11-01

    Guideline recommendations for the same clinical condition may vary. The purpose of this study was to determine the degree of agreement among comparable asthma and bronchiolitis treatment recommendations from guidelines. National and international guidelines were searched by using guideline databases (eg, National Guidelines Clearinghouse: December 16-17, 2014, and January 9, 2015). Guideline recommendations were categorized as (1) recommend, (2) optionally recommend, (3) abstain from recommending, (4) recommend against a treatment, and (5) not addressed by the guideline. The degree of agreement between recommendations was evaluated by using an unweighted and weighted κ score. Pairwise comparisons of the guidelines were evaluated similarly. There were 7 guidelines for asthma and 4 guidelines for bronchiolitis. For asthma, there were 166 recommendation topics, with 69 recommendation topics given in ≥2 guidelines. For bronchiolitis, there were 46 recommendation topics, with 21 recommendation topics provided in ≥2 guidelines. The overall κ for asthma was 0.03, both unweighted (95% confidence interval [CI]: -0.01 to 0.07) and weighted (95% CI: -0.01 to 0.10); for bronchiolitis, it was 0.32 unweighted (95% CI: 0.16 to 0.52) and 0.15 weighted (95% CI: -0.01 to 0.5). Less agreement was found in national and international guidelines for asthma than for bronchiolitis. Additional studies are needed to determine if differences are based on patient preferences and values and economic considerations or if other recommendation-level, guideline-level, and condition-level factors are driving these differences. Copyright © 2017 by the American Academy of Pediatrics.

  2. 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)

  3. Mirror reactor blankets

    International Nuclear Information System (INIS)

    Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

    1976-01-01

    The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

  4. Blanket testing in NET

    International Nuclear Information System (INIS)

    Chazalon, M.; Daenner, W.; Libin, B.

    1989-01-01

    The testing stages in NET for the performance assessment of the various breeding blanket concepts developed at the present time in Europe for DEMO (LiPb and ceramic blankets) and the requirements upon NET to perform these tests are reviewed. Typical locations available in NET for blanket testing are the central outboard segments and the horizontal ports of in-vessel sectors. These test positions will be connectable with external test loops. The number of test loops (helium, water, liquid metal) will be such that each major class of blankets can be tested in NET. The test positions, the boundary conditions and the external test loops are identified and the requirements for test blankets are summarized (author). 6

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

    International Nuclear Information System (INIS)

    Jackson, D.P.; Selander, W.N.; Townes, B.M.

    1985-01-01

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  6. Limitations on blanket performance

    International Nuclear Information System (INIS)

    Malang, S.

    1999-01-01

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

  7. INTERNAL-CYCLE VARIATION OF SOLAR DIFFERENTIAL ROTATION

    International Nuclear Information System (INIS)

    Li, K. J.; Xie, J. L.; Shi, X. J.

    2013-01-01

    The latitudinal distributions of the yearly mean rotation rates measured by Suzuki in 1998 and 2012 and Pulkkinen and Tuominen in 1998 are utilized to investigate internal-cycle variation of solar differential rotation. The rotation rate at the solar equator seems to have decreased since cycle 10 onward. The coefficient B of solar differential rotation, which represents the latitudinal gradient of rotation, is found to be smaller in the several years after the minimum of a solar cycle than in the several years after the maximum time of the cycle, and it peaks several years after the maximum time of the solar cycle. The internal-cycle variation of the solar rotation rates looks similar in profile to that of the coefficient B. A new explanation is proposed to address such a solar-cycle-related variation of the solar rotation rates. Weak magnetic fields may more effectively reflect differentiation at low latitudes with high rotation rates than at high latitudes with low rotation rates, and strong magnetic fields may more effectively repress differentiation at relatively low latitudes than at high latitudes. The internal-cycle variation is inferred as the result of both the latitudinal migration of the surface torsional pattern and the repression of strong magnetic activity in differentiation.

  8. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  9. Tritium breeding blanket

    International Nuclear Information System (INIS)

    Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

    1991-01-01

    The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

  10. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  11. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

    Purpose: To produce tritium more than consumed, through thermonuclear reaction. Constitution: The energy spectrum of neutron generated by neutron multiplying reaction in a neutron multiplying blanket and moderated neutrons has a large ratio in a low energy section. In the low-energy absorption region of stainless steel which is a material of cooling pipes constituting a neutron multiplying blanket cooling channel, the neutrons are absorbed, lessening the neutron multiplying effect. To prevent this, the neutron multiplying blanket cooling channel is covered with tritium breeding blankets, thereby enabling the production of a substantially great amount of tritium more than the amount of tritium to be consumed by the thermonuclear reaction by preventing neutron absorption by the component materials of the cooling channel, improving the tritium breeding ratio by 20 to 25 %, and increasing the efficiency of use of neutrons for tritium generation. (Horiuchi, T.)

  12. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  13. Novel blanket design for ICTR's

    International Nuclear Information System (INIS)

    Abdel-Khalik, S.I.; Conn, R.W.; Wolfer, W.G.; Larsen, E.N.; Sviatoslavsky, I.N.

    1978-01-01

    A novel blanket design for ICTRs is described. This blanket is used in SOLASE, the conceptual laser fusion reactor of the University of Wisconsin. The blanket to be described offers numerous advantages, including low cost, low weight, low induced radioactivity levels, the potential for hands-on maintenance, modular construction, low pressure, ability to decouple first wall and blanket coolant temperatures, adequate breeding, low tritium inventory and leakage, and sufficiently long life

  14. 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.

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

    International Nuclear Information System (INIS)

    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 2 O) and lithium zirconate (Li 2 ZrO 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

  16. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

    1989-01-01

    The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

  17. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

    1993-01-01

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

  18. Beyond blanket terms: Challenges for the explanatory value of variational (neuro-)ethology. Comment on "Answering Schrödinger's question: A free-energy formulation" by Maxwell James Désormeau Ramstead et al.

    Science.gov (United States)

    Bruineberg, Jelle; Hesp, Casper

    2018-03-01

    Ramstead et al. [9] integrate the free-energy principle (FEP) [5] and evolutionary systems theory (EST) [1] in order to develop a "meta-theoretical ontology of life", called 'variational neuro-ethology' (VNE). In drawing upon such abstract notions and integrating them even further, they prove themselves to be the ultimate "hedgehogs" [2]: aiming for the ultimate integration of the life sciences and social sciences under one unifying principle. We endorse this pursuit of theoretical integration, especially when derived from first principles. The fundamental nature of their work is exemplified by the book the authors take as their starting point: Schrödinger's What is Life?[10]. Given the variety and levels of complexity involved in defining "life", providing an answer to this question is challenging. We first briefly comment on VNE as a label and then highlight some possible problems for the kinds of explanations that would follow from VNE. We address the interrelated charges of (1) merely providing Bayesian and evolutionary "just-so" stories [4,6], and (2) limited interpretative clarity when casting "life" as a series of nested Markov blankets. As a pre-emptive response to these critical remarks, we sketch a few ways forward that we find promising.

  19. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    Ozawa, Yoshihiro; Uda, Tatsuhiko; Maki, Koichi.

    1993-01-01

    The present invention provides a blanket of a thermonuclear device which produces tritium fuels consumed in plasmas while converting neutrons generated in the plasmas into heat energy. That is, zirconium is coated to at least one of neutron breeder pebbles and breeder pebbles, to suppress reaction between them by being in direct contact with each other at a high temperature. Further, fins are attached to a cooling pipe at a pitch smaller than the diameter of both of the pebbles, to prevent direct contact at whole surface of the pebbles and the cooling pipe, which would lower a temperature excessively. The length of the fin is controlled to control the thickness of a helium gas gap. With such constitution, direct contact of neutron breeder pebbles and the breeder pebble which are to be filled and mixed, and tend to react at a high temperature, can be prevented. The temperature of the breeding blanket is reliably prevented from lowering below a tritium emitting temperature. The structure is simplified and the production is facilitated. (I.S.)

  20. Sulfate and dissolved sulfide variation under low COD/Sulfate ratio in Up-flow Anaerobic Sludge Blanket (UASB treating domestic wastewater

    Directory of Open Access Journals (Sweden)

    Sérvio Túlio Alves Cassini

    2012-04-01

    Full Text Available In this study, the dynamics of sulfate reduction and dissolved sulfide generation (S2-, HS-, H2Saq in liquid phase was evaluated in an UASB reactor treating domestic wastewater with low COD/Sulfate content. The evaluation in the UASB reactor was performed at three sludge heights (0.25, 1.25, 2.25 taps and effluent of the reactor. Sulfate reduction was verified in the reactor, with an average reduction of 24 % throughout the experiment period. However, the dissolved sulfide concentration in the reactor was not higher than 5.0 mg Sdiss/L. The kinetic model of first order showed good fit to describe the sulfate reduction under different COD/sulfate ratio, with K1app between 2.94x10-5 s-1 and 1.17x10-5 s-1 with correlation coefficients for data over 91%. The maximum rate to sulfate reduction was 18.0 mg SO42-/L.h-1 and small variation in COD/sulfate ratio promotes a significant change both in sulfate and sulfide concentrations.

  1. Application of vanadium alloys to a fusion reactor blanket

    Energy Technology Data Exchange (ETDEWEB)

    Bethin, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center)

    1984-05-01

    Vanadium and vanadium alloys are of interest in fusion reactor blanket applications due to their low induced radioactivity and outstanding elevated temperature mechanical properties during neutron irradiation. The major limitation to the use of vanadium is its sensitivity to oxygen impurities in the blanket environment, leading to oxygen embrittlement. A quantitative analysis was performed of the interaction of gaseous impurities in a helium coolant with vanadium and the V-15Cr-5Ti alloy under conditions expected in a fusion reactor blanket. It was shown that the use of unalloyed V would impose severe restrictions on the helium gas cleanup system due to excessive oxygen buildup and embrittlement of the metal. However, internal oxidation effects and the possibly lower terminal oxygen solubility in the alloy would impose much less severe cleanup constraints. It is suggested that V-15Cr-5Ti is a promising candidate for certain blanket applications and deserves further consideration.

  2. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

    Le Marois, G.; Federzoni, L.; Bucci, P.; Revirand, P.

    2000-01-01

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

  3. Structural design study of tritium breeding blanket with a lead layer as a neutron multiplier

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kitamura, Kazunori; Minato, Akio; Sakamoto, Hiroki; Yamamoto, Takashi

    1980-12-01

    Thermal and structural design study of a tritium breeding blanket with a lead layer for a International Tokamak Reactor (INTOR) is carried out. Tube in shell type blanket with a lead layer is found to be promising. The volume fraction of structural material in the lead layer can be small enough to keep the neutron multiplication effect of lead. Reasonable value of shell effect is attainable due to lead layer in the front part of the blanket. (author)

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

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Yican

    2000-01-01

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

  5. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    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

  6. Design, Manufacture, and Experimental Serviceability Validation of ITER Blanket Components

    Science.gov (United States)

    Leshukov, A. Yu.; Strebkov, Yu. S.; Sviridenko, M. N.; Safronov, V. M.; Putrik, A. B.

    2017-12-01

    In 2014, the Russian Federation and the ITER International Organization signed two Procurement Arrangements (PAs) for ITER blanket components: 1.6.P1ARF.01 "Blanket First Wall" of February 14, 2014, and 1.6.P3.RF.01 "Blanket Module Connections" of December 19, 2014. The first PA stipulates development, manufacture, testing, and delivery to the ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels intended for withstanding the heat flux from the plasma up to 4.7MW/m2. Two Russian institutions, NIIEFA (Efremov Institute) and NIKIET, are responsible for the implementation of this PA. NIIEFA manufactures plasma-facing components (PFCs) of the EHF FW panels and performs the final assembly and testing of the panels, and NIKIET manufactures FW beam structures, load-bearing structures of PFCs, and all elements of the panel attachment system. As for the second PA, NIKIET is the sole official supplier of flexible blanket supports, electrical insulation key pads (EIKPs), and blanket module/vacuum vessel electrical connectors. Joint activities of NIKIET and NIIEFA for implementing PA 1.6.P1ARF.01 are briefly described, and information on implementation of PA 1.6.P3.RF.01 is given. Results of the engineering design and research efforts in the scope of the above PAs in 2015-2016 are reported, and results of developing the technology for manufacturing ITER blanket components are presented.

  7. Stress analysis of blanket vessel for JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

    A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

  8. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  9. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  10. Reducing beryllium content in mixed bed solid-type breeder blankets

    Energy Technology Data Exchange (ETDEWEB)

    Shimwell, J., E-mail: mail@jshimwell.com [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Lilley, S.; Morgan, L.; Packer, L.; Kovari, M.; Zheng, S. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); McMillan, J. [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom)

    2016-11-01

    Highlights: • The ratio of breeder ceramic to neutron multiplier of breeder blankets was varied linearly with depth. • Blankets with varying composition were found to perform better than uniform composition breeder blankets. • It was also possible to reduce the amount of beryllium required by the blanket. - Abstract: Beryllium (Be) is a precious resource with many high value uses, the low energy threshold (n,2n) reaction makes Be an excellent neutron multiplier for use in fusion breeder blankets. Estimates of Be requirements and available resources suggest that this could represent a major supply difficulty for solid-type blanket concepts. Reducing the quantity of Be required by breeder blankets would help to alleviate the problem to some extent. In addition, it is important that the reduction in the Be quantity does not diminish the blanket's performance in key aspects such as the tritium breeding ratio (TBR), energy multiplication and peak nuclear heating. Mixed pebble bed designs allow for the multiplier fraction to be varied throughout the blanket. This neutronics study used MCNP 6 to investigate linear variations of the multiplier fraction in relation to blanket depth, in order to better utilise the important multiplying Be(n,2n) and breeding reactions. Blankets with a uniform multiplier fraction showed little scope for reduction in Be mass. Blankets with varying multiplier fractions were able to simultaneously use 10% less Be, increase the energy amplification by 1%, reduce the peak heating by 7% and maintaining a sufficient TBR when compared to the performance achievable using a uniform composition.

  11. Internal Occipital Crest Misalignment with Internal Occipital Protuberance: A Case Report of Posterior Cranial Fossa Anatomic Variations

    Science.gov (United States)

    Kim, Jae Ha

    2016-01-01

    During gross anatomy head and neck laboratory session, one dissection group observed an abnormal anatomic variation in the posterior cranial fossa of a 94-year-old male cadaver. The internal occipital crest was not aligned with internal occipital protuberance and groove for superior sagittal sinus. It seemed that the internal occipital protuberance was shifted significantly to the right side. As a result the skull was overly stretched in order to connect with the internal occipital ridge. These internal skull variations of occipital bone landmarks can influence the location of adjacent dural venous sinuses and possibly influence cerebrospinal fluid flow. Similar anatomical anomalies have been attributed to presence of hydrocephalus and abnormalities in cisterna magna. PMID:27648322

  12. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

    In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

  13. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  14. Distillate Fuel Trends: International Supply Variations and Alternate Fuel Properties

    Science.gov (United States)

    2013-01-31

    fuel in NATO countries will have some amount of FAME present. There is some work being done on hydrocarbon alternatives but the regulatory structure ... synthesis or hydrotreatment – Requirements and test methods.” According to the specification, paraffinic diesel fuel does not meet the current requirements...or international specification for triglyceride based fuel oils (straight vegetable oil / raw vegetable oil). The same holds true for alcohol-based

  15. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    Smith, D.L.; Sze, D.K.

    1986-02-01

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

  16. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

    Sze, D.K.; Jung, J.; Cheng, E.T.

    1986-01-01

    Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

  17. The blanket interface to TSTA

    International Nuclear Information System (INIS)

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

    1988-01-01

    The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. 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. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

  18. MIT LMFBR blanket physics project progress report No. 7, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1976-01-01

    Work during the period was devoted primarily to a range of analytical/numerical investigations, including evaluation of means to improve external blanket designs, beneficial attributes of the use of internal blankets, improved methods for the calculation of heterogeneous self-shielding and parametric studies of calculated spectral indices. Experimental work included measurements of the ratio of U-238 captures to U-235 fissions in a standard blanket mockup, and completion of development work on the radiophotoluminescent readout of LiF thermoluminescent detectors. The most significant findings were that there is very little prospect for substantial improvement in the breeding performance of external blankets, but internal blankets continue to show promise, particularly if they are used in such a way as to increase the volume fraction of fuel inside the core envelope. An improved equivalence theorem was developed which may allow use of fast reactor methods to calculate heterogeneously self-shielded cross sections in both fast and thermal reactors

  19. Blanket maintenance by remote means using the cassette blanket approach

    International Nuclear Information System (INIS)

    Werner, R.W.

    1978-01-01

    Induced radioactivity in the blanket and other parts of a fusion reactor close to the plasma zone will dictate remote assembly, disassembly, and maintenance procedures. Time will be of the essence in these procedures. They must be practicable and certain. This paper discusses the reduction of a complicated Tokamak reactor to a simpler assembly via the use of a vacuum building in which to house the reactor and the introduction in this new model of cassette blanket modules. The cassettes significantly simplify remote handling

  20. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou

    1998-01-01

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

  1. Development of blanket remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

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

  2. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-06-01

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

  3. Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

    International Nuclear Information System (INIS)

    Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

    2006-01-01

    In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

  4. Comparative analysis of a fusion reactor blanket in cylindrical and toroidal geometry using Monte Carlo

    International Nuclear Information System (INIS)

    Chapin, D.L.

    1976-03-01

    Differences in neutron fluxes and nuclear reaction rates in a noncircular fusion reactor blanket when analyzed in cylindrical and toroidal geometry are studied using Monte Carlo. The investigation consists of three phases--a one-dimensional calculation using a circular approximation to a hexagonal shaped blanket; a two-dimensional calculation of a hexagonal blanket in an infinite cylinder; and a three-dimensional calculation of the blanket in tori of aspect ratios 3 and 5. The total blanket reaction rate in the two-dimensional model is found to be in good agreement with the circular model. The toroidal calculations reveal large variations in reaction rates at different blanket locations as compared to the hexagonal cylinder model, although the total reaction rate is nearly the same for both models. It is shown that the local perturbations in the toroidal blanket are due mainly to volumetric effects, and can be predicted by modifying the results of the infinite cylinder calculation by simple volume factors dependent on the blanket location and the torus major radius

  5. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

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

  6. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

    Strebkov, Yu.; Sidorov, A.; Danilov, I.

    1994-01-01

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

  7. Seasonal variation of semidiurnal internal tides in the East/Japan Sea

    Science.gov (United States)

    Jeon, Chanhyung; Park, Jae-Hun; Varlamov, Sergey M.; Yoon, Jong-Hwan; Kim, Young Ho; Seo, Seongbong; Park, Young-Gyu; Min, Hong Sik; Lee, Jae Hak; Kim, Cheol-Ho

    2014-05-01

    The seasonal variation of semidiurnal internal tides in the East/Japan Sea was investigated using 25 month long output from a real-time ocean forecasting system. The z coordinate eddy-resolving high-resolution numerical model, called the RIAM ocean model, incorporates data assimilation that nudges temperature and salinity fields together with volume transport through the Korea Strait to produce realistic oceanic currents and stratification. In addition to atmospheric forcing, it includes tidal forcing of 16 major components along open boundaries. The model generates energetic semidiurnal internal tides around the northern entrance of the Korea Strait. Energy conversion from barotropic to baroclinic (internal) tides varies seasonally with maxima in September (ranging 0.48-0.52 GW) and minima in March (ranging 0.11-0.16 GW). This seasonal variation is induced by the seasonality in stratification near the southwestern East/Japan Sea. The propagation distance of the internal tides is associated with generation intensity and wavelength. From late summer to early winter, the semidiurnal internal tides travel relatively far from the generation region due to stratification changes; its energy dissipates less as a result of longer wavelengths. Our results suggest that spatiotemporal variation of internal-tide-induced mixing due to the seasonality in the generation, propagation, and dissipation of internal tides should be considered for a more realistic simulation of water masses and circulation in models of the East/Japan Sea.

  8. TIME VARIATION AND ASYMMETRY IN THE WORLD PRICE OF COVARIANCE RISK: THE IMPLICATIONS FOR INTERNATIONAL DIVERSIFICATION

    OpenAIRE

    Olan T. Henry; Nilss Olekalns; Kalvinder Shields

    2004-01-01

    The International Capital Asset Pricing Model measures country risk in terms of the conditional covariance of national returns with the world return. Using impulse responses from a multivariate nonlinear model we provide evidence of time variation and asymmetry in the measure of country risk. and the implied benefit to international diversification. The evidence implies that the price of risk and the benefits from diversification may differ in a statistically and economically meaningful fashi...

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

    International Nuclear Information System (INIS)

    1979-11-01

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  10. Literacy Skills Gaps: A Cross-Level Analysis on International and Intergenerational Variations

    Science.gov (United States)

    Kim, Suehye

    2018-01-01

    The global agenda for sustainable development has centred lifelong learning on UNESCO's Education 2030 Framework for Action. The study described in this article aimed to examine international and intergenerational variations in literacy skills gaps within the context of the United Nations Sustainable Development Goals (SDGs). For this purpose, the…

  11. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    Energy Technology Data Exchange (ETDEWEB)

    Pereslavtsev, Pavel, E-mail: pavel.pereslavtsev@kit.edu [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Boltzmannstrasse 2, 85748 Garching (Germany); Fischer, Ulrich [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, {sup 6}Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  12. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    International Nuclear Information System (INIS)

    Pereslavtsev, Pavel; Bachmann, Christian; Fischer, Ulrich

    2016-01-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, "6Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  13. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

    The objectives of the Blanket Comparison and Selection Study (BCSS) can be stated as follows: (1) Define a small number (approx. 3) of blanket design concepts that should be the focus of the blanket R and D program. A design concept is defined by the selection of all materials (e.g., breeder, coolant, structure and multiplier) and other major characteristics that significantly influence the R and D requirements. (2) Identify and prioritize the critical issues for the leading blanket concepts. (3) Provide the technical input necessary to develop a blanket R and D program plan. Guidelines for prioritizing the R and D requirements include: (a) critical feasibility issues for the leading blanket concepts will receive the highest priority, and (b) for equally important feasibility issues, higher R and D priority will be given to those that require minimum cost and short time

  14. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

    2000-07-01

    This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

  15. Effects of buffer thickness on ATW blanket performance

    International Nuclear Information System (INIS)

    Yang, W. S.; Mercatali, L.; Taiwo, T. A.; Hill, R. N.

    2001-01-01

    This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy ( and lt; 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level

  16. Effects of Buffer Thickness on ATW Blanket Performance

    International Nuclear Information System (INIS)

    Yang, W.S.; Mercatali, L.; Taiwo, T.A.; Hill, R.N.

    2002-01-01

    This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level. (authors)

  17. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

    Malang, S.; Casal, V.; Arheidt, K.; Fischer, U.; Link, W.; Rust, K.

    1986-01-01

    A blanket concept for NET using liquid lithium-lead both as breeder material and as coolant is described. The need for inboard breeding is avoided by using beryllium as neutron multiplier in the outboard blanket. Novel flow channel inserts are employed in all poloidal ducts to reduce the MHD pressure drop. The concept offers a simple mechanical design and a higher tritium breeding ratio compared to water- and gas-cooled blankets. (author)

  18. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

    Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Usher, J.L.

    1980-04-01

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

  19. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    Gohar, Y.; Baker, C.C.; Smith, D.L.

    1987-10-01

    The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of ∼2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs

  20. 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.

  1. 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.

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

    International Nuclear Information System (INIS)

    Lee C. Cadwallader

    2007-01-01

    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

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

    OpenAIRE

    Doutres , Olivier; Atalla , Noureddine

    2010-01-01

    International audience; 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...

  4. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

  5. International variation in neighborhood walkability, transit, and recreation environments using geographic information systems

    DEFF Research Database (Denmark)

    Adams, Marc A; Frank, Lawrence D; Schipperijn, Jasper

    2014-01-01

    associations with physical activity comes from single countries or regions with limited variation in urban form. This paper documents variation in comparable built environment features across countries from diverse regions. METHODS: The International Physical Activity and the Environment Network (IPEN) study...... recreation facilities around each participant's residential address using 1-km and 500-m street network buffers. RESULTS: Eleven of 12 countries and 15 cities had objective GIS data on built environment features. We observed a 38-fold difference in median residential densities, a 5-fold difference in median......, and the United States followed a common research protocol to develop internationally comparable measures. Using detailed instructions, GIS-based measures included features such as walkability (i.e., residential density, street connectivity, mix of land uses), and access to public transit, parks, and private...

  6. Radon variations in active volcanoes and in regions with high seismicity: internal and external factors

    International Nuclear Information System (INIS)

    Segovia, N.; Cruz-Reyna, S. De la; Mena, M.

    1986-01-01

    The results of 4 years of observations of radon concentrations in soils of active volcanoes of Costa Rica and a highly seismic region in Mexico are discussed. A distinction is made between the influences of external (mostly meteorological) and internal (magmatic or tectonic) factors on the variation in radon levels. The geological meaning of the radon data can be thus enhanced if the external factors are excluded. (author)

  7. (D,T) Driven thorium hybrid blankets

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.; Khan, S.; Sahin, S.

    1983-01-01

    Recently, a project has started, with the aim to establish the neutronic performance and the basic design of an experimental fusionfission (hybrid) reactor facility, called AYMAN, in cylinderical geometry. The fusion reactor will have to be simulated by a (D,T) neutron generator. Fissile and fertile fuel will have to surround the neutron generator as a cylinderical blanket to simulate the boundary conditions of the hybrid blanket in a proper way. This geometry is consistent with Tandem Mirror Hybrid Blanket design and with most of the ICF blanket designs. A similar experimental installation will become operational around 1984 at the Swiss Federal Institute of Technology in Lausanne, Switzerland known under the project LOTUS. Due to the limited dimensions of the experimental cavity of the LOTUS-hybrid reactor, the LOTUS blankets have to be designed in plane geometry. Also, the bulky form of the Haefely neutron generator of the LOTUS facility obliges one to design a blanket in the plane geometry. This results in a vacuum left boundary conditions for the LOTUS blanket. The importance of a reflecting left boundary condition on the overall neutronic performance of a hybrid blanket has been analyzed in previous work in detail

  8. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-12-01

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

  9. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

  10. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

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

  11. Japanese contributions to ITER testing program of solid breeder blankets for DEMO

    International Nuclear Information System (INIS)

    Kuroda, Toshimasa; Yoshida, Hiroshi; Takatsu, Hideyuki; Maki, Koichi; Mori, Seiji; Kobayashi, Takeshi; Suzuki, Tatsushi; Hirata, Shingo; Miura, Hidenori.

    1991-04-01

    ITER Conceptual Design Activity (CDA), which has been conducted by four parties (Japan, EC, USA and USSR) since May 1988, has been finished on December 1990 with a great achievement of international design work of the integrated fusion experimental reactor. Numerous issues of physics and technology have been clarified for providing a framework of the next phase of ITER (Engineering Design Activity; EDA). Establishment of an ITER testing program, which includes technical test issues of neutronics, solid breeder blankets, liquid breeder blankets, plasma facing components, and materials, has been one of the goals of the CDA. This report describes Japanese proposal for the testing program of DEMO/power reactor blanket development. For two concepts of solid breeder blanket (helium-cooled and water-cooled), identification of technical issues, scheduling of test program, and conceptual design of test modules including required test facility such as cooling and tritium recovery systems have been carried out as the Japanese contribution to the CDA. (author)

  12. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

    The NET machine has three horizontal ports reserved for testing tritium breeding blanket designs during the physics phase and possibly five during the technology phase. The design of the ports and test blankets are modular to accept a range of blanket options, provide radiation shielding and allow routine replacement. Radiation levels during replacement or maintenance require that all operations must be carried out remotely. The paper describes the problems overcome in providing a port design which includes attachment to the vacuum vessel with double vacuum seals, an integrated cooled first wall and support guides for the test blanket module. The method selected to remotely replace the test module whilst controlling the spread of contamination is also adressed. The paper concludes that the provisions of a test blanket facility based on the NET machine design is feasible. (orig.)

  13. 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...... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....

  14. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

    This paper presents an overview of the critical materials issues that must be considered in the development of a tritium breeding blanket for a tokamak fusion reactor that operates on the D-T-Li fuel cycle. The primary requirements of the blanket system are identified and the important criteria that must be considered in the development of blanket technology are summarized. The candidate materials are listed for the different blanket components, e.g., breeder, coolant, structure and neutron multiplier. Three blanket concepts that appear to offer the most potential are: (1) liquid-metal breeder/coolant, (2) liquid-metal breeder/separate coolant, and (3) solid breeder/separate coolant. The major uncertainties associated with each of the design concepts are discussed and the key materials R and D requirements for each concept are identified

  15. Fusion blankets for catalyzed D--D and D--He3 reactors

    International Nuclear Information System (INIS)

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

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

  16. Fusion blankets for catalyzed D--D and D--3He reactors

    International Nuclear Information System (INIS)

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

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

  17. Improved modules for the blanket of RTO/RC ITER

    International Nuclear Information System (INIS)

    Elio, F.; Ioki, K.; Cardella, A.

    2000-01-01

    This paper describes innovative design aspects that are considered to optimise the blanket modules for the reduced technical objective/reduced cost international thermonuclear experimental reactor. The blanket modules have a vertical straight profile facing the plasma, and the first wall is built in small and flat panels. Copper may be applied only in front of the first row of cooling passages. The radial cooling of the shield block avoids a complex by-pass at the back and opens up the possibility to use cast instead of forged steel. Slits in the shield block and in the first wall reduce the electromagnetic forces enough to allow the support of the modules on the vessel and the mechanical attachment of the first wall panels

  18. Blanket handling concepts for future fusion power plants

    International Nuclear Information System (INIS)

    Bogusch, E.; Gottfried, R.; Maisonnier, D.

    2003-01-01

    In the frame of the power plant conceptual studies (PPCS) launched by the European Commission, two main blanket handling concepts have been investigated with respect to engineering feasibility and the impact on the plant availability and on cost: the large module handling concept (LMHC) and the large sector handling concept (LSHC). The LMHC has been considered as the reference handling concept while the LSHC has been considered as an attractive alternative to the LMHC due to its potential of smaller replacement times and hence increasing the plant availability. Although no principle feasibility issue has been identified, a number of engineering issues have been highlighted for the LSHC that would require considerable efforts for their resolution. Since its availability of about 77% based on a replacement time for all the internals of about 4.2 months is slightly lower than for the LMHC, the LMHC remains the reference blanket replacement concept for a conceptual reactor

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

    International Nuclear Information System (INIS)

    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.

    2006-07-01

    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

  20. Variation in Emergency Department vs Internal Medicine Excess Charges in the United States.

    Science.gov (United States)

    Xu, Tim; Park, Angela; Bai, Ge; Joo, Sarah; Hutfless, Susan M; Mehta, Ambar; Anderson, Gerard F; Makary, Martin A

    2017-08-01

    Uninsured and insured but out-of-network emergency department (ED) patients are often billed hospital chargemaster prices, which exceed amounts typically paid by insurers. To examine the variation in excess charges for services provided by emergency medicine and internal medicine physicians. Retrospective analysis was conducted of professional fee payment claims made by the Centers for Medicare & Medicaid Services for all services provided to Medicare Part B fee-for-service beneficiaries in calendar year 2013. Data analysis was conducted from January 1 to July 31, 2016. Markup ratios for ED and internal medicine professional services, defined as the charges submitted by the hospital divided by the Medicare allowable amount. Our analysis included 12 337 emergency medicine physicians from 2707 hospitals and 57 607 internal medicine physicians from 3669 hospitals in all 50 states. Services provided by emergency medicine physicians had an overall markup ratio of 4.4 (340% excess charges), which was greater than the markup ratio of 2.1 (110% excess charges) for all services performed by internal medicine physicians. Markup ratios for all ED services ranged by hospital from 1.0 to 12.6 (median, 4.2; interquartile range [IQR], 3.3-5.8); markup ratios for all internal medicine services ranged by hospital from 1.0 to 14.1 (median, 2.0; IQR, 1.7-2.5). The median markup ratio by hospital for ED evaluation and management procedure codes varied between 4.0 and 5.0. Among the most common ED services, laceration repair had the highest median markup ratio (7.0); emergency medicine physician review of a head computed tomographic scan had the greatest interhospital variation (range, 1.6-27.7). Across hospitals, markups in the ED were often substantially higher than those in the internal medicine department for the same services. Higher ED markup ratios were associated with hospital for-profit ownership (median, 5.7; IQR, 4.0-7.1), a greater percentage of uninsured patients seen

  1. Diurnal and Seasonal Variations in Mid-Latitude Geomagnetic Field During International Quiet Days: BOH Magnetometer

    Directory of Open Access Journals (Sweden)

    Junga Hwang

    2012-12-01

    Full Text Available Korea Astronomy and Space Science Institute researchers have installed and operated magnetometers at Bohyunsan Observatory to measure the Earth's magnetic field variations in South Korea. In 2007, we installed a fluxgate magnetometer (RFP-523C to measure H, D, and Z components of the geomagnetic field. In addition, in 2009, we installed a Overhauser proton sensor to measure the absolute total magnetic field F and a three-axis magneto-impedance sensor for spectrum analysis. Currently three types of magnetometer data have been accumulated. In this paper, we use the H, D, Z components of fluxgate magnetometer data to investigate the characteristics of mid-latitude geomagnetic field variation. To remove the temporary changes in Earth’s geomagnetic filed by space weather, we use the international quiet days’ data only. In other words, we performed a superposed epoch analysis using five days per each month during 2008-2011. We find that daily variations of H, D, and Z shows similar tendency compared to previous results using all days. That is, H, D, Z all three components’ quiet intervals terminate near the sunrise and shows maximum 2-3 hours after the culmination and the quiet interval start from near the sunset. Seasonal variations show similar dependences to the Sun. As it becomes hot season, the geomagnetic field variation’s amplitude becomes large and the quiet interval becomes shortened. It is well-known that these variations are effects of Sq current system in the Earth’s atmosphere. We confirm that the typical mid-latitude geomagnetic field variations due to the Sq current system by excluding all possible association with the space weather.

  2. Moving ring field-reversed mirror blanket design considerations

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, L.; Kessel, C.; Norman, J.; Schultz, K.R.

    1981-01-01

    A blanket design for the Moving Ring Field-Reversed Mirror Reactor (MRFRM) is presented in this paper. The design emphasis is placed on minimizing the induced radioactivities in the first-wall, blanket and shield. To this end, aluminum-alloy was selected as the reference structural material, giving dose rates two weeks after shutdown that are 3 to 4 orders of magnitude lower than comparable steel structures. The aluminum first-wall is water-cooled and thermally insulated from the high temperature SiC-clad Li 2 O tritium breeding zone. A local tritium breeding ratio of 1.05 was obtained for the design. The tritium is extracted from the Li 2 O by the use of a small dry helium purge stream through the SiC tubes. About 1 ppM hydrogen is added to the helium purge stream to enhance the tritium recovery rate. Helium at 28 atmospheres pressure is circulated through the blanket and shield, with an outlet temperature of 850 0 C, which is coupled with an existing small size closed-cycle gas turbine (CCGT) power conversion system. The spatial and temporal variations of the first-wall temperature caused by the translational movement of the plasma rings along the axis of the cylindrical reactor were evaluated. The after-heat cooling problems of the first-wall were also considered

  3. Corrosion characteristics of an aqueous self-cooled fusion blanket

    International Nuclear Information System (INIS)

    Bogaerts, W.F.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Jackson, D.

    1986-01-01

    A novel aqueous self-cooled blanket concept (ASCB) has recently been proposed. This blanket concept, as applied to a MARS-like tandem mirror reactor, consists of disks of spiraling tubes of Zircaloy-4 housed in a structural container of vanadium alloy (V-15 Ti-5 Cr). The Zircaloy tubes are cooled by a mixture of light and heavy water with 9 g of LiOH per 100 cm 3 of water dissolved in the coolant. A major issue for the feasibility of the integrated blanket coil concept is the chemical compatibility of the coolant and Zircaloy. Initial corrosion tests have been undertaken in order to resolve this question. Results clearly show that successful alloy heats can be prepared, for which corrosion problems will probably not be the limiting factor of the ASCB design concept. As is quite well known from fission engineering studies, small variations in the alloy compositions or in the metallurgical structure may, however, be able to cause significant alterations in the oxidation or corrosion rates. Further tests will be necessary to resolve the remaining uncertainties and to determine the behavior of successful alloy heats in the presence of trace impurities in order to address the sensitivity to localized corrosion phenomena such as pitting, stress corrosion cracking, and intergranular attack

  4. Accelerator-driven molten-salt blankets: Physics issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

    1994-01-01

    A number of nuclear physics issues concerning the Los Alamos molten-salt, accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external, moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

  5. Accelerator-driven molten-salt blankets: Physics issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

    1994-01-01

    A number of nuclear physics issues concerning the Los Alamos molten-salt accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

  6. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-05-01

    Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

  7. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

    1993-11-01

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

  8. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

    1999-07-01

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

  9. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

    2017-01-15

    Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

  10. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    International Nuclear Information System (INIS)

    WONG, CPC; MALANG, S; NISHIO, S; RAFFRAY, R; SAGARA, S

    2002-01-01

    OAK A271 ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS. First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

  11. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

    Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-10-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

  12. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

    1998-01-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

  13. Beryllium R&D for blanket application

    Science.gov (United States)

    Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

    1998-10-01

    The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

  14. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1978-01-01

    The thermal and mechanical aspects of hybrid reactor blanket design considerations are discussed. This paper is intended as a companion to that of J. D. Lee of Lawrence Livermore Laboratory on the nuclear aspects of hybrid reactor blanket design. The major design characteristics of hybrid reactor blankets are discussed with emphasis on the areas of difference between hybrid reactors and standard fusion or fission reactors. Specific examples are used to illustrate the design tradeoffs and choices that must be made in hybrid reactor design. These examples are drawn from the work on the Mirror Hybrid Reactor

  15. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Sawan, M.E.; Cheng, E.T.

    1985-01-01

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

  16. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Sawan, M.L.; Cheng, E.T.

    1986-01-01

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li 17 Pb 83 blankets. (author)

  17. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Young, J.R.

    1975-01-01

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

  18. New concepts for controlled fusion reactor blanket design

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Avci, H.; El-Maghrabi, M.

    1975-01-01

    Several new concepts for fusion reactor blanket design based on the idea of shifting, or tailoring, the neutron spectrum incident on the first structural wall are presented. The spectral shifter is a nonstructural element which can be made of graphite, silicon carbide, or three dimensionally woven carbon fibers (and containing other materials as appropriate) placed between the neutron source and the first structural wall. The softened neutron spectrum incident on the structural components leads to lower gas production and atom displacement rates than in more standard fusion blanket designs. In turn, this results in longer anticipated lifetimes for the structural materials and can significantly reduce radioactivity and afterheat levels. In addition, the neutron spectrum in the first structural wall can be made to approach the flux shape in fast breeder reactors. Such spectral softening means that existing radiation facilities may be more profitably used to provide relevant materials radiation damage data for the structural materials in these fusion blanket designs. This general class of blanket concepts are referred to as internal spectral shifter and energy converter, or ISSEC concepts. These specific design concepts fall into three main categories: ISSEC/EB concepts based on utilizing existing designs which breed tritium behind the first structural wall; ISSEC/IB concepts based on breeding tritium inside the first vacuum wall; and ISSEC/Bu concepts based on using boron, carbon, and perhaps, beryllium to obtain an energy multiplier and converter design that does not attempt to breed tritium or utilize lithium. The detailed analyses relate specifically to the nuclear performance of ISSEC systems and to a discussion of materials radiation damage problems in the structural material.(U.S.)

  19. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

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

    1976-01-01

    The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. 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, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)

  20. Blanket design for imploding liner systems

    International Nuclear Information System (INIS)

    Schaffer, M. J.

    1980-01-01

    The blanket design comprises hot, molten, rotating liquid vortex systems suitable for rapidly compressing confined plasmas, in which stratified immiscible liquid layers having successively greater mass densities outwardly of the axis of rotation are provided

  1. APT target-blanket fabrication development

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, D.L.

    1997-06-13

    Concepts for producing tritium in an accelerator were translated into hardware for engineering studies of tritium generation, heat transfer, and effects of proton-neutron flux on materials. Small-scale target- blanket assemblies were fabricated and material samples prepared for these performance tests. Blanket assemblies utilize composite aluminum-lead modules, the two primary materials of the blanket. Several approaches are being investigated to produce large-scale assemblies, developing fabrication and assembly methods for their commercial manufacture. Small-scale target-blanket assemblies, designed and fabricated at the Savannah River Site, were place in Los Alamos Neutron Science Center (LANSCE) for irradiation. They were subjected to neutron flux for nine months during 1996-97. Coincident with this test was the development of production methods for large- scale modules. Increasing module size presented challenges that required new methods to be developed for fabrication and assembly. After development, these methods were demonstrated by fabricating and assembling two production-scale modules.

  2. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

    The design features of the hybrid reactor blanket and main parameters are presented. The stress analysis is performed by using computer codes SAP5p and SAP6 with the three kinds of blanket module loadings, i.e, the pressure of coolant, the blanket weight and the thermal loading. Numerical calculation results indicate that the stresses of the blanket are smaller than the allowable ones of the material, the blanket design is therefore reasonable

  3. Nongenetic causes of childhood cancers: evidence from international variation, time trends, and risk factor studies

    International Nuclear Information System (INIS)

    Bunin, Greta R.

    2004-01-01

    Ionizing radiation and a variety of genetic conditions are thought to explain 5-10% of childhood cancers. Infection with Epstein-Barr virus (EBV) in parts of Africa and human immunodeficiency virus (HIV) increase the risk of Burkitt's lymphoma and Kaposi's sarcoma, respectively. Other risk factors have not been conclusively identified. A review of the data on international variation in incidence, recent changes in incidence, and risk factors suggests that many childhood cancers are likely to have nongenetic causes. The pattern of international variation and associations with surrogates of infection suggest an infectious etiology for acute lymphoblastic leukemia, although no agent has been identified. The biologic plausibility is strong that maternal consumption of food containing DNA topoisomerase II inhibitors may increase the risk of acute myeloid leukemia, although the data are limited now. For brain tumors, cured meats, polyomaviruses, and farm exposures may have etiologic roles. Changes in the incidence and characteristics of children with hepatoblastoma as well as risk factor studies suggest a role for an exposure of very low birth weight babies. High birth weight, tea or coffee consumption, and certain paternal occupations have shown some consistency in their association with Wilms' tumor. For most of the other cancers, very few epidemiologic studies have been conducted, so it is not surprising that nongenetic risk factors have not been detected. The most important difference between the cancers for which there are good etiologic clues and those for which there are not may be the number of relevant studies

  4. TAXATION AND INTERNAL MIGRATION - EVIDENCE FROM THE SWISS CENSUS USING COMMUNITY-LEVEL VARIATION IN INCOME TAX RATES

    OpenAIRE

    Liebig, Thomas; Puhani, Patrick A.; Sousa-Poza, Alfonso

    2006-01-01

    We investigate the relationship between income tax rate variation and internal migration for the unique case of Switzerland, whose system of determining tax rates primarily at the community level results in enough variation to permit analysis of their influence on migration. Specifically, using Swiss census data, we analyze migratory responses to tax rate variations for various groups defined by age, education, and nationality/residence permit. The results suggest that young Swiss college gra...

  5. The fusion blanket program at Chalk River

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-03-01

    Work on the Fusion Blanket Program commenced at Chalk River in 1984 June. Co-funded by Canadian Fusion Fuels Technology Project and Atomic Energy of Canada Limited, the Program utilizes Chalk River expertise in instrumented irradiation testing, ceramics, tritium technology, materials testing and compound chemistry. This paper gives highlights of studies to date on lithium-based ceramics, leading contenders for the fusion blanket

  6. National Trends and Geographic Variation in Bilateral Internal Mammary Artery Use in the United States.

    Science.gov (United States)

    Iribarne, Alexander; Goodney, Philip P; Flores, Alyssa M; DeSimone, Joseph; DiScipio, Anthony W; Austin, Andrea; McCullough, Jock N

    2017-12-01

    The goal of this study was to characterize the adoption rate and regional variation in bilateral internal mammary artery (BIMA) use during coronary artery bypass grafting (CABG) in the United States. Observational study of 100% sample of fee-for-service Medicare beneficiaries aged 65 years or older, continuously enrolled in Parts A and B from 2009 to 2014 (n = 162,860,439). Rates of beneficiaries receiving a BIMA versus single internal mammary artery (SIMA) during CABG are expressed per 1,000 beneficiaries and aggregated by Hospital Referral Region (HRR). An HRR is a validated unit for quantifying regional variation in health care. The absolute national rate of BIMA use declined during the study period from 0.21 claims per 1,000 beneficiaries in 2009 to 0.13 in 2014 (p < 0.001). When indexed to overall CABG volume, no change was seen in the frequency of BIMA use over time (p = 0.883). SIMA use ranged from 1.3 to 8.5 claims per 1,000 Medicare beneficiaries, whereas BIMA use ranged from 0 to 1.5 (p < 0.001). A significant correlation was found between regional volume of SIMA use and likelihood of BIMA use (correlation coefficient 0.673, p < 0.001). Although both SIMA and BIMA use correlated with regional volume of diagnostic cardiac catheterization, the correlation was stronger for SIMA use (correlation coefficient 0.962 versus 0.682, p < 0.001). Over the past 5 years, no growth was seen in BIMA use among Medicare beneficiaries, and the frequency of BIMA use during CABG remained low. There was significant regional variation in BIMA use, however, which demonstrates opportunity for continued growth of BIMA grafting. Copyright © 2017 The Society of Thoracic Surgeons. Published by Elsevier Inc. All rights reserved.

  7. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Malang, S.; Nishio, S.; Raffray, R.; Sagara, A.

    2002-01-01

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

  8. Workshop on cold-blanket research

    International Nuclear Information System (INIS)

    1977-05-01

    The objective of the workshop was to identify and discuss cold-plasma blanket systems. In order to minimize the bombardment of the walls by hot neutrals the plasma should be impermeable. This requires a density edge-thickness product of nΔ > 10 15 cm -2 . An impermeable cold plasma-gas blanket surrounding a hot plasma core reduces the plasma wall/limiter interaction. Accumulation of impurities in this blanket can be expected. Fuelling from a blanket may be possible as shown by experimental results, though not fully explained by classical transport of neutrals. Refuelling of a reacting plasma had to be ensured by inward diffusion. Experimental studies of a cold impermeable plasma have been done on the tokamak-like Ringboog device. Simulation calculations for the next generation of large tokamaks using a particular transport model, indicate that the plasma edge profile can be controlled to reduce the production of sputtered impurities to an acceptable level. Impurity control requires a small fraction of the radial space to accomodate the cold-plasma layer. The problem of exhaust is, however, more complicated. If the cold-blanket scheme works as predicted in the model calculations, then α-particles generated by fusion will be transported to the cold outside layer. The Communities' experimental programme of research has been discussed in terms of the tokamaks which are available and planned. Two options present themselves for the continuation of cold-blanket research

  9. Liquid lithium blanket processing studies

    International Nuclear Information System (INIS)

    Talbot, J.B.; Clinton, S.D.

    1979-01-01

    The sorption of tritium on yttrium from flowing molten lithium and the subsequent release of tritium from yttrium for regeneration of the metal sorbent were investigated to evaluate the feasibility of such a tritium-recovery process for a fusion reactor blanket of liquid lithium. In initial experiments with the forced convection loop, yttrium samples were contacted with lithium at 300 0 C. A mass transfer coefficient of 2.5 x 10 - cm/sec, which is more than an order of magnitude less than the value measured in earlier static experiments, was determined for the flowing lithium system. Rates of tritium release from yttrium samples were measured to evaluate possible thermal regeneration of the sorbent. Values for diffusion coefficients at 505, 800, and 900 0 C were estimated to be 1.1 x 10 -13 , 4.9 x 10 -12 , and 9.3 x 10 -10 cm 2 /sec, respectively. Tritium release from yttrium was investigated at higher temperatures and with hydrogen added to the argon sweep gas to provide a reducing atmosphere

  10. Variation in esophageal physiology testing in clinical practice: Results from an international survey.

    Science.gov (United States)

    Sweis, R; Heinrich, H; Fox, M

    2018-03-01

    Advances in clinical measurement of esophageal motility and function have improved the assessment of swallowing disorders and reflux symptoms. Variation in data acquisition, analysis, and reporting exists and impacts on diagnosis and management. This study examined variation in esophageal manometry methodology between institutions to establish the status in current practice. A structured survey was distributed through international NGM societies using an Internet-based platform. Questions explored infrastructure, technology, analysis, and reporting. Responses were received from 91 centers from 29 countries. Eighteen (20%) centers used "conventional" manometry, 75 (82%) high-resolution manometry, and 53 (58%) HR impedance manometry. All centers documented motility for single water swallows. The Chicago Classification was applied by 65 (71.4%) centers. In contrast, analysis of EGJ morphology varied widely. Adjunctive testing was often applied: multiple rapid swallows (77%), rapid drink challenge (77%), single solid swallows (63%), and a standard test meal (18%). Of 86 (94.5%) units that offered pH impedance (pH-Z) studies, approximately half (53.5%) performed tests on acid-suppressant medication in patients with a high pretest probability (eg, erosive esophagitis). Most (75.6%) centers manually reviewed every reflux event. Others examined pH-Z data only prior to symptoms. To assess symptom association with reflux events, 73.6% centers analyzed each symptom separately, whereas 29.7% centers pooled symptoms. There is marked variation in the data acquisition, analysis, and reporting of esophageal manometry studies. Further efforts to improve quality and uniformity in testing and reporting are required. This survey provides information upon which best-practice guidelines can be developed. © 2017 John Wiley & Sons Ltd.

  11. Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

    Science.gov (United States)

    Urgorri, F. R.; Moreno, C.; Carella, E.; Rapisarda, D.; Fernández-Berceruelo, I.; Palermo, I.; Ibarra, A.

    2017-11-01

    The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg h-1 while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

  12. Conceptual design and testing strategy of a dual functional lithium-lead test blanket module in ITER and EAST

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

    A dual functional lithium-lead (DFLL) test blanket module (TBM) concept has been proposed for testing in the International Thermonuclear Experimental Reactor (ITER) and the Experimental Advanced Superconducting Tokamak (EAST) in China to demonstrate the technologies of the liquid lithium-lead breeder blankets with emphasis on the balance between the risks and the potential attractiveness of blanket technology development. The design of DFLL-TBM concept has the flexibility of testing both the helium-cooled quasi-static lithium-lead (SLL) blanket concept and the He/PbLi dual-cooled lithium-lead (DLL) blanket concept. This paper presents an effective testing strategy proposed to achieve the testing target of SLL and DLL DEMO blankets relevant conditions, which includes three parts: materials R and D and small-scale out-of-pile mockups testing in loops, middle-scale TBMs pre-testing in EAST and full-scale consecutive TBMs testing corresponding to different operation phases of ITER during the first 10 years. The design of the DFLL-TBM concept and the testing strategy ability to test TBMs for both blanket concepts in sequence and or in parallel for both ITER and EAST are discussed

  13. Literacy skills gaps: A cross-level analysis on international and intergenerational variations

    Science.gov (United States)

    Kim, Suehye

    2018-02-01

    The global agenda for sustainable development has centred lifelong learning on UNESCO's Education 2030 Framework for Action. The study described in this article aimed to examine international and intergenerational variations in literacy skills gaps within the context of the United Nations Sustainable Development Goals (SDGs). For this purpose, the author examined the trend of literacy gaps in different countries using multilevel and multisource data from the OECD's Programme for the International Assessment of Adult Competencies (PIAAC) and UNESCO Institute for Lifelong Learning survey data from the third edition of the Global Report on Adult Learning and Education (GRALE III). In this article, particular attention is paid to exploring the specific effects of education systems on literacy skills gaps among different age groups. Key findings of this study indicate substantial intergenerational literacy gaps within countries as well as different patterns of literacy gaps across countries. Young generations generally outscore older adults in literacy skills, but feature bigger gaps when examined by gender and social origin. In addition, this study finds an interesting tendency for young generations to benefit from a system of Recognition, Validation and Accreditation (RVA) in closing literacy gaps by formal schooling at country level. This implies the potential of an RVA system for tackling educational inequality in initial schooling. The article concludes with suggestions for integrating literacy skills as a foundation of lifelong learning into national RVA frameworks and mechanisms at system level.

  14. Test Blanket Working Group's recent activities

    International Nuclear Information System (INIS)

    Vetter, J.E.

    2001-01-01

    The ITER Test Blanket Working Group (TBWG) has continued its activities during the period of extension of the EDA with a revised charter on the co-ordination of the development work performed by the Parties and by the JCT leading to a co-ordinated test programme on ITER for a DEMO-relevant tritium breeding blanket. This follows earlier work carried out until July 1998, which formed part of the ITER Final Design Report (FDR), completed in 1998. Whilst the machine parameters for ITER-FEAT have been significantly revised compared to the FDR, testing of breeding blanket modules remains a main objective of the test programme and the development of a reactor-relevant breeding blanket to ensure tritium fuel self-sufficiency is recognized a key issue for fusion. Design work and R and D on breeding blanket concepts, including co-operation with the other Contacting Parties of the ITER-EDA for testing these concepts in ITER, are included in the work plans of the Parties

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    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. (paper)

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

    International Nuclear Information System (INIS)

    Zacchia, F.; Daenner, W.; Stefanis, L. de; Ferrari, M.; Gerber, A.; Mustoe, J.

    1998-01-01

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

  17. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

    The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

  18. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1994-11-01

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.) [de

  19. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Palmer, B.J.F.

    1987-11-01

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

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

    International Nuclear Information System (INIS)

    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

  1. Pebble bed blanket design for deuterium burning tandem mirror reactors

    International Nuclear Information System (INIS)

    Grotz, S.P.; Dhir, V.K.

    1983-01-01

    The UCLA tandem mirror reactor, SATYR, was developed around the capability of tandem mirrors with thermal barriers to burn deuterium at reasonable efficiency levels. The pebble bed concept has been incorporated into our blanket design for the following reasons: 1) Large area-to-volume ratio for purposes of heat removal; 2) Large volume of structure for high thermal capacity thus increasing the safety margin during off-normal incidents; 3) Relatively inexpensive manufacturing costs because of large acceptable tolerances and lack of exotic materials (i.e., lithium). A simplified stress analysis of the blanket module was performed to optimize and simplify the design. The pre-specified stress intensity limitations used were based upon a 30-year predicted lifetime for each module. Along with stress analysis of the vessel a detailed thermal hydraulic analysis of the pebble bed has been completed. Parameters affecting the pebble bed design are fluidization velocity, pressure drop, heat transfer coefficient, thermally induced stress in the spheres and spatial variation of the power density. Although reasonable gross thermal efficiencies of the 2 designs has been achieved (28% for H 2 O and 39% for He) the high net recirculating power fraction for heating and neutral beams results in relatively low net plant efficiencies (21% and 27%). The results show that a blanket can be designed with good thermal efficiency and a relative-ly simple configuration. However, application of this concept to the high Q deuterium-tritium fuel cycle would have difficulties resulting from the need for continuous removal of the tritium. (orig./HP)

  2. Low activity blanket designs and heat transfer for experimental power reactors

    International Nuclear Information System (INIS)

    Fillo, J.; Tichler, P.; Lazareth, O.; Powell, J.

    1976-01-01

    Two minimum activity blanket designs are described, based on the ANL TEPR circular design parameters. A first wall loading (plasma on) of 1.0 MW(th)/m 2 has been assumed. The first option is composed of SAP (sintered aluminum product) modules. The oval shaped SAP shell, in which approximately 45 percent of the fusion energy is removed, is maintained at a temperature of approximately 400 0 C by a He coolant stream. The remaining 55 percent of the fusion energy is deposited in a thermally insulated hot interior (SiC and B 4 C) and removed by a separate He coolant, with exit temperature of 800 0 C. In the second option, the blanket is a thick graphite block structure (approximately 50 cm thickness) with SAP coolant tubes carrying He (50 atm) embedded deep within the graphite to minimize radiation damage. The neutron and gamma energy deposited in the graphite is radiated along internal slots and conducted through the graphite to the coolant tubes. To reduce surface evaporation above 2000 0 C, the blanket surface is radiatively cooled to a low temperature radiation sink, a bank of He cooled SAP tubes. Approximately 20 percent of the fusion energy is removed in this region, the remaining 80 percent in the primary graphite-aluminum blanket. Both blanket options are mounted on heavy Al backing plates, cooled by He, which are in turn supported from the fixed shield

  3. Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor

    Directory of Open Access Journals (Sweden)

    Bong Guen Hong

    2018-02-01

    Full Text Available A configuration of a fusion-driven transmutation reactor with a low aspect ratio tokamak-type neutron source was determined in a self-consistent manner by using coupled analysis of tokamak systems and neutron transport. We investigated the impact of blanket configuration on the characteristics of a fusion-driven transmutation reactor. It was shown that by merging the TRU burning blanket and tritium breeding blanket, which uses PbLi as the tritium breeding material and as coolant, effective transmutation is possible. The TRU transmutation capability can be improved with a reduced blanket thickness, and fast fluence at the first wall can be reduced.  Article History: Received: July 10th 2017; Received: Dec 17th 2017; Accepted: February 2nd 2018; Available online How to Cite This Article: Hong, B.G. (2018 Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor. International Journal of Renewable Energy Development, 7(1, 65-70. https://doi.org/10.14710/ijred.7.1.65-70

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

    International Nuclear Information System (INIS)

    Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

    2004-01-01

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

  5. Epoxy blanket protects milled part during explosive forming

    Science.gov (United States)

    1966-01-01

    Epoxy blanket protects chemically milled or machined sections of large, complex structural parts during explosive forming. The blanket uniformly covers all exposed surfaces and fills any voids to support and protect the entire part.

  6. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.

    1981-01-01

    The Tandem Mirror Reactor, with its cylindrical central cell, has led to numerous blanket designs taking advantage of the simple geometry. Also many new applications for fusion neutrons are now being considered. To the pure fusion electricity producers and hybrids producing fissile fuel, we are adding studies of synthetic fuel producers and fission-suppressed hybrids. The three blanket concepts presented are new ideas and should be considered illustrative of the breadth of Livermore's application studies. They are not meant to imply fully analyzed designs

  7. Fusion blanket high-temperature heat transfer

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1983-01-01

    Deep penetration of 14 MeV neutrons makes two-temperature region blankets feasible. A relatively low-temperature (approx. 300 0 C) metallic structure is the vacuum/coolant pressure boundary, while the interior of the blanket, which is a simple packed bed of nonstructural material, operates at very high temperatures (>1000 0 C). The water-cooled shell structure is thermally insulated from the steam-cooled interior. High-temperature steam can dramatically increase the efficiency of electric power generation, as well as produce hydrogen and oxygen-based synthetic fuels at high-efficiency

  8. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

    Tritium release from the candidate ceramic materials, Li 2 O, LiA10 2 , Li 2 SiO 3 , Li 4 SiO 4 and Li 2 ZrO 3 , is being investigated in many blanket programs. Factors that affect tritium release from the ceramic into the helium sweep gas stream include operating temperature, ceramic microstructure, tritium transport and solubility in the solid. A review is presented of the material properties studied and of the irradiation programs and the results are summarized. The ceramic breeder blanket concept is briefly reviewed

  9. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  10. Processing and waste disposal needs for fusion breeder blankets system

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1988-01-01

    We evaluated the waste disposal and recycling requirements for two types of fusion breeder blanket (solid and liquid). The goal was to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under U.S. Nuclear Regulatory Commission regulations. Described in this paper are the radionuclides expected in fusion blanket materials, plans for reprocessing and disposal of blanket components, and estimates for the operating costs involved in waste disposal. (orig.)

  11. Tritium inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Reiter, F.

    1990-01-01

    This report reviews studies of the transport of hydrogen isotopes in the DEMO relevant water-cooled Pb-17Li blanket to be tested in NET and in a self-cooled blanket which uses Pb-17Li or Flibe as a liquid breeder material and V or Fe as a first wall material. The time dependences of tritium inventory and permeation in these blankets and of deuterium and tritium recycling in the self-cooled blanket are presented and discussed

  12. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

    Engineering data on neutron irradiation performance of tritium breeders are needed to design the breeding blanket of fusion reactor. In this study, tritium release experiments of the breeders were carried out to examine the effects of various parameters (such as sweep gas flow rate, hydrogen content in sweep gas, irradiation temperature and thermal neutron flux) on tritium generation and release behavior. Lithium titanate (Li 2 TiO 3 ) is considered as a candidate tritium breeder in the blanket design of International Thermonuclear Experimental Reactor (ITER). As for the shape of the breeder material, a small spherical form is preferred to reduce the thermal stress induced in the breeder. Li 2 TiO 3 pebbles of about 170g in total weight and with 0.3 and 2 mm in diameter were manufactured by a wet process, and an assembly packed with the binary Li 2 TiO 3 pebbles was irradiated in Japan Materials Testing Reactor (JMTR). The tritium was generated in the Li 2 TiO 3 pebble bed and released from the pebble bed, and was swept downstream using the sweep gas for on-line analysis of tritium content. Concentration of total tritium and gaseous tritium (HT or T 2 gas) released from the Li 2 TiO 3 pebble bed were measured by ionization chambers, and the ratio of (gaseous tritium)/(total tritium) was evaluated. The sweep gas flow rate was changed from 100 to 900cm 3 /min, and hydrogen content in the sweep gas was changed from 100 to 10000 ppm. Furthermore, thermal neutron flux was changed using a window made of hafnium (Hf) neutron absorber. The irradiation temperature at an outer region of the Li 2 TiO 3 pebble bed was held between 200 and 400degC. The main results of this experiment are summarized as follows. 1) When the temperature at the outside edge of the Li 2 TiO 3 pebble bed exceeded 100degC, the tritium release from the Li 2 TiO 3 pebble bed started. The ratio of the tritium release rate and the tritium generation rate (normalized tritium release rate: R/G) reached

  13. Tritium inventory and permeation in the ITER breeding blanket

    International Nuclear Information System (INIS)

    Violante, V.; Tosti, S.; Sibilia, C.; Felli, F.; Casadio, S.; Alvani, C.

    2000-01-01

    A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5x10 -3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5x10 -5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity

  14. Tritium transport in HCLL and WCLL DEMO blankets

    Energy Technology Data Exchange (ETDEWEB)

    Candido, Luigi [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Utili, Marco [ENEA UTIS- C.R. Brasimone, Bacino del Brasimone, Camugnano, BO (Italy); Nicolotti, Iuri [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Zucchetti, Massimo, E-mail: massimo.zucchetti@polito.it [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy)

    2016-11-01

    Highlights: • Tritium inventories and tritium losses are the main output of the presented model for HCLL and WCLL. • A parametric study has been performed, to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and/or losses. • An improved design is needed, in order to reduce the radiological hazard related to tritium activity. According to test number 7, HCLL-BB could be able to have a tritium inventory of 33.05 g and losses of 19.55 Ci/d. • WCLL-BB shows a very low radiological risk, much lower than that suggested (inventory: 17.48 g, losses: 3.2 Ci/d). An ptimization study has been performed aiming to minimize the water flow rate for an upgraded design. • Both for HCLL and WCLL, the most critical parameters able to produce relevant variations in inventories and losses are the helium/water fraction, the CPS/WDS and the permeation reduction factors. - Abstract: The Helium-Cooled Lithium Lead (HCLL) and Water-Cooled Lithium Lead (WCLL) Breeding Blankets are two of the four blanket designs proposed for DEMO reactor. The study of tritium transport inside the blankets is fundamental to assess their preliminary design and safety features. A mathematical model has been derived, in a new form making makes easier to determine the most critical components as far as tritium losses and tritium inventories are concerned, and to model the tritium performance of the whole system. Two cases have been studied, the former with tritium generation rate constant in time and the latter considering a typical pulsed operation for a time span of 100 h. Tritium inventories and tritium losses are the main output of the model. Tritium concentrations, inventories and losses are initially calculated and compared for the two blankets, in a reference case without permeation barriers or cold traps. A parametric study to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and

  15. Cross-National Variations in Student Employment and Academic Performance: The Roles of National Context and International Law.

    Science.gov (United States)

    Byun, Soo-Yong; Henck, Adrienne; Post, David

    Most existing research indicates that working students perform more poorly than do full-time students on standardized achievement tests. However, we know there are wide international variations in this gap. This article shows that national and international contexts help to explain the gap in the academic performance between working and non-working middle-school students. We combined data from the 2003 Trends in International Mathematics and Science Study (TIMSS) eighth-grade assessment with the country specific information on socioeconomic and educational conditions, as well as the timing of each country's ratification of an international treaty regulating child labor. Our multilevel analyses show that, while student employment was generally negatively associated with academic performance, this negative association is smaller in countries that by 1995 had ratified the International Labour Organization's Convention No. 138 on child labor. These findings highlight the role of national and international policy in structuring the consequences of student employment for academic performance.

  16. Cross-National Variations in Student Employment and Academic Performance: The Roles of National Context and International Law*

    Science.gov (United States)

    Byun, Soo-yong; Henck, Adrienne; Post, David

    2014-01-01

    Most existing research indicates that working students perform more poorly than do full-time students on standardized achievement tests. However, we know there are wide international variations in this gap. This article shows that national and international contexts help to explain the gap in the academic performance between working and non-working middle-school students. We combined data from the 2003 Trends in International Mathematics and Science Study (TIMSS) eighth-grade assessment with the country specific information on socioeconomic and educational conditions, as well as the timing of each country's ratification of an international treaty regulating child labor. Our multilevel analyses show that, while student employment was generally negatively associated with academic performance, this negative association is smaller in countries that by 1995 had ratified the International Labour Organization's Convention No. 138 on child labor. These findings highlight the role of national and international policy in structuring the consequences of student employment for academic performance. PMID:25632163

  17. Intensive Care Unit Structure Variation and Implications for Early Mobilization Practices. An International Survey.

    Science.gov (United States)

    Bakhru, Rita N; McWilliams, David J; Wiebe, Douglas J; Spuhler, Vicki J; Schweickert, William D

    2016-09-01

    Early mobilization (EM) improves outcomes for mechanically ventilated patients. Variation in structure and organizational characteristics may affect implementation of EM practices. We queried intensive care unit (ICU) environment and standardized ICU practices to evaluate organizational characteristics that enable EM practice. We recruited 151 ICUs in France, 150 in Germany, 150 in the United Kingdom, and 500 in the United States by telephone. Survey domains included respondent characteristics, hospital and ICU characteristics, and ICU practices and protocols. We surveyed 1,484 ICU leaders and received a 64% response rate (951 ICUs). Eighty-eight percent of respondents were in nursing leadership roles; the remainder were physiotherapists. Surveyed ICUs were predominantly mixed medical-surgical units (67%), and 27% were medical ICUs. ICU staffing models differed significantly (P equipment were highly variable among respondents. International ICU structure and practice is quite heterogeneous, and several factors (multidisciplinary rounds, setting daily goals for patients, presence of a dedicated physiotherapist, country, and nurse/patient staffing ratio) are significantly associated with the practice of EM. Practice and barriers may be far different based upon staffing structure. To achieve successful implementation, whether through trials or quality improvement, ICU staffing and practice patterns must be taken into account.

  18. Spectrum of sequence variations in the FANCA gene: an International Fanconi Anemia Registry (IFAR) study.

    Science.gov (United States)

    Levran, Orna; Diotti, Raffaella; Pujara, Kanan; Batish, Sat D; Hanenberg, Helmut; Auerbach, Arleen D

    2005-02-01

    Fanconi anemia (FA) is an autosomal recessive disorder that is defined by cellular hypersensitivity to DNA cross-linking agents, and is characterized clinically by developmental abnormalities, progressive bone-marrow failure, and predisposition to leukemia and solid tumors. There is extensive genetic heterogeneity, with at least 11 different FA complementation groups. FA-A is the most common group, accounting for approximately 65% of all affected individuals. The mutation spectrum of the FANCA gene, located on chromosome 16q24.3, is highly heterogeneous. Here we summarize all sequence variations (mutations and polymorphisms) in FANCA described in the literature and listed in the Fanconi Anemia Mutation Database as of March 2004, and report 61 novel FANCA mutations identified in FA patients registered in the International Fanconi Anemia Registry (IFAR). Thirty-eight novel SNPs, previously unreported in the literature or in dbSNP, were also identified. We studied the segregation of common FANCA SNPs in FA families to generate haplotypes. We found that FANCA SNP data are highly useful for carrier testing, prenatal diagnosis, and preimplantation genetic diagnosis, particularly when the disease-causing mutations are unknown. Twenty-two large genomic deletions were identified by detection of apparent homozygosity for rare SNPs. In addition, a conserved SNP haplotype block spanning at least 60 kb of the FANCA gene was identified in individuals from various ethnic groups. (c) 2005 Wiley-Liss, Inc.

  19. International variation in adherence to referral guidelines for suspected cancer: a secondary analysis of survey data.

    Science.gov (United States)

    Nicholson, Brian D; Mant, David; Neal, Richard D; Hart, Nigel; Hamilton, Willie; Shinkins, Bethany; Rubin, Greg; Rose, Peter W

    2016-02-01

    Variation in cancer survival persists between comparable nations and appears to be due, in part, to primary care practitioners (PCPs) having different thresholds for acting definitively in response to cancer-related symptoms. To explore whether cancer guidelines, and adherence to them, differ between jurisdictions and impacts on PCPs' propensity to take definitive action on cancer-related symptoms. A secondary analysis of survey data from six countries (10 jurisdictions) participating in the International Cancer Benchmarking Partnership. PCPs' responses to five clinical vignettes presenting symptoms and signs of lung (n = 2), colorectal (n = 2), and ovarian cancer (n = 1) were compared with investigation and referral recommendations in cancer guidelines. Nine jurisdictions had guidelines covering the two colorectal vignettes. For the lung vignettes, although eight jurisdictions had guidelines for the first, the second was covered by a Swedish guideline alone. Only the UK and Denmark had an ovarian cancer guideline. Survey responses of 2795 PCPs (crude response rate: 12%) were analysed. Guideline adherence ranged from 20-82%. UK adherence was lower than other jurisdictions for the lung vignette covered by the guidance (47% versus 58%; P nations and poor guideline adherence does not explain differential survival. Guidelines that fail to cover high-risk presentations or that recommend non-definitive action may reduce definitive diagnostic action. © British Journal of General Practice 2016.

  20. INTOR first wall/blanket/shield activity

    International Nuclear Information System (INIS)

    Gohar, Y.; Billone, M.C.; Cha, Y.S.; Finn, P.A.; Hassanein, A.M.; Liu, Y.Y.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.

    1986-01-01

    The main emphasis of the INTOR first wall/blanket/shield (FWBS) during this period has been upon the tritium breeding issues. The objective is to develop a FWBS concept which produces the tritium requirement for INTOR operation and uses a small fraction of the first wall surface area. The FWBS is constrained by the dimensions of the reference design and the protection criteria required for different reactor components. The blanket extrapolation to commercial power reactor conditions and the proper temperature for power extraction have been sacrificed to achieve the highest possible local tritium breeding ratio (TBR). In addition, several other factors that have been considered in the blanket survey study include safety, reliability, lifetime fluence, number of burn cycles, simplicity, cost, and development issues. The implications of different tritium supply scenarios were discussed from the cost and availability for INTOR conditions. A wide variety of blanket options was explored in a preliminary way to determine feasibility and to see if they can satisfy the INTOR conditions. This survey and related issues are summarized in this report. Also discussed are material design requirements, thermal hydraulic considerations, structure analyses, tritium permeation through the first wall into the coolant, and tritium inventory

  1. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

    The primary function of beryllium in a fusion reactor blanket is neutron multiplication to enhance tritium breeding. However, because heat, tritium and helium will be generated in and/or transported through beryllium and because the beryllium is in contact with other blanket materials, the thermal, mechanical, tritium/helium and compatibility properties of beryllium are important in blanket design. In particular, tritium retention during normal operation and release during overheating events are safety concerns. Accommodating beryllium thermal expansion and helium-induced swelling are important issues in ensuring adequate lifetime of the structural components adjacent to the beryllium. Likewise, chemical/metallurgical interactions between beryllium and structural components need to be considered in lifetime analysis. Under accident conditions the chemical interaction between beryllium and coolant and breeding materials may also become important. The performance of beryllium in fusion blanket applications depends on fabrication variables and operational parameters. First the properties database is reviewed to determine the state of knowledge of beryllium performance as a function of these variables. Several design calculations are then performed to indicate ranges of fabrication and operation variables that lead to optimum beryllium performance. Finally, areas for database expansion and improvement are highlighted based on the properties survey and the design sensitivity studies

  2. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Anzidei, L.

    1994-01-01

    There are four breeding blanket concepts for a fusion DEMO reactor under development within the framework of the fusion technology programme of the European Union (EU). This paper describes the design of these concepts, the accompanying R + D programme and the status of the development. (authors). 8 figs., 1 tab

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

    International Nuclear Information System (INIS)

    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

  4. Ethnic variations in ulcerative colitis: Experience of an international hospital in Thailand.

    Science.gov (United States)

    Permpoon, Vibhakorn; Pongpirul, Krit; Anuras, Sinn

    2016-08-06

    To investigate the clinical characteristics, treatment, medication use, and treatment response in patients with ulcerative colitis (UC) across ethnic groups. This study retrospectively analyzed medical records of all 268465 patients who visited the Bumrungrad International Digestive Disease Center during 2005-2010. The demographics, clinical characteristics, medication use, results of investigations, and medical and surgical management for patients with UC were evaluated. Evaluation included sigmoidoscopy and colonoscopy performed in compliance with the American Society of Gastrointestinal Endoscopy practice guidelines. Patient ethnicities were categorized into seven groups: Thai, Oriental, South Asian (SA), Middle Eastern (ME), Caucasian, African, and Hispanic. UC pathological severity was classified into inactive, mild, moderate, and severe. Associations between categorical variables were analyzed using the χ(2) or Fischer's exact test. Associations between categorical and interval variables were analyzed using Student's t-test and/or analysis of covariance. UC was diagnosed in 371 of the 268465 patients: male 56.33%; ME 42%, Caucasian 23%, and Thai 19%. Annual incidence of UC was 82 cases per 100000 with wide ethnic variation, ranging from 29 to 206 cases per 100000 in Oriental and ME patients, respectively. Of the patients with UC, 16.71% had severe UC with highest incidence among the patients from ME (20.39%) and lowest among the Caucasian population (11.90%). ME had highest proportion of pancolitis (52.90%), followed by Caucasian (45.35%) and Asian (34.40%). Only 20.93% of Caucasian patients received steroid, compared with 26.40% and 27.10% of Asian and Middle Eastern, respectively (P = 0.732). Overall, 13.72% of UC patients did not respond to steroid therapy, with non-significantly higher proportions of non-responders among Asian and Middle Eastern patients (15.22% and 15.04%, respectively) (P = 0.781). On average, 5.93% underwent surgical management with

  5. Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

    1980-09-01

    A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

  6. Status of the European R and D on beryllium as multiplier material for breeder blankets

    International Nuclear Information System (INIS)

    Moeslang, A.; Boccaccini, L.V.; Rabaglino, E.; Piazza, G.; Cardella, A.; Sannen, L.; Scibetta, M.; Laan, J. van der; Hegeman, J.B.J.W.

    2004-01-01

    Within the international fusion community a variety of breeding blanket concepts are being considered, ranging from more conservative concepts to higher-risk concepts for fusion power reactors. In Europe, the Helium Cooled Pebble Bed (HCPB) blanket is one of the two reference concepts which will also be tested as Test Blanket Module (TBM) in ITER. In addition to the R and D for structural parts of the HCPB blanket, a considerable effort is devoted to the production and qualification of ceramic breeder and neutron multiplier (beryllium or beryllide) pebble beds. Since in the HCPB blanket pebbles made of lithium ceramics are foreseen, a high volume fraction of beryllium as a neutron multiplier to Li-based ceramic of about 4: l is needed. The typical loading conditions for beryllium are, with a neutron wall load of ∼12.5 MWa/m 2 and in ∼5 years lifetime: T min ∼300degC, T max ∼600-900degC, displacement damage ∼80 dpa, peak 4 He production ∼26000 appm and peak 3 H production ∼700 appm at the End-Of-Life. The behaviour of beryllium under irradiation is considered to be a key issue of the HCPB blanket, because of swelling due to helium bubbles and tritium retention. A large R and D programme on beryllium has been implemented in Europe, aimed at characterising and predicting the material behaviour before and under irradiation. An overview on experimental and modelling activities performed during the past 2 years is given with typical results on non-irradiated and irradiated Beryllium materials and pebble beds and the relevance of major results on future beryllium R and D is addressed. (author)

  7. International variations in primary care physician consultation time: a systematic review of 67 countries.

    Science.gov (United States)

    Irving, Greg; Neves, Ana Luisa; Dambha-Miller, Hajira; Oishi, Ai; Tagashira, Hiroko; Verho, Anistasiya; Holden, John

    2017-11-08

    To describe the average primary care physician consultation length in economically developed and low-income/middle-income countries, and to examine the relationship between consultation length and organisational-level economic, and health outcomes. This is a systematic review of published and grey literature in English, Chinese, Japanese, Spanish, Portuguese and Russian languages from 1946 to 2016, for articles reporting on primary care physician consultation lengths. Data were extracted and analysed for quality, and linear regression models were constructed to examine the relationship between consultation length and health service outcomes. One hundred and seventy nine studies were identified from 111 publications covering 28 570 712 consultations in 67 countries. Average consultation length differed across the world, ranging from 48 s in Bangladesh to 22.5 min in Sweden. We found that 18 countries representing about 50% of the global population spend 5 min or less with their primary care physicians. We also found significant associations between consultation length and healthcare spending per capita, admissions to hospital with ambulatory sensitive conditions such as diabetes, primary care physician density, physician efficiency and physician satisfaction. There are international variations in consultation length, and it is concerning that a large proportion of the global population have only a few minutes with their primary care physicians. Such a short consultation length is likely to adversely affect patient healthcare and physician workload and stress. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2017. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

  8. Variation in predicted internal concentrations in relation to PBPK model complexity for rainbow trout

    Energy Technology Data Exchange (ETDEWEB)

    Salmina, E.S.; Wondrousch, D. [UFZ Department of Ecological Chemistry, Helmholtz Centre for Environmental Research, Permoserstr. 15, 04318 Leipzig (Germany); Institute for Organic Chemistry, Technical University Bergakademie Freiberg, Leipziger Str. 29, 09596 Freiberg (Germany); Kühne, R. [UFZ Department of Ecological Chemistry, Helmholtz Centre for Environmental Research, Permoserstr. 15, 04318 Leipzig (Germany); Potemkin, V.A. [Department of Chemistry, South Ural State Medical University, Vorovskogo 64, 454048, Chelyabinsk (Russian Federation); Schüürmann, G. [UFZ Department of Ecological Chemistry, Helmholtz Centre for Environmental Research, Permoserstr. 15, 04318 Leipzig (Germany); Institute for Organic Chemistry, Technical University Bergakademie Freiberg, Leipziger Str. 29, 09596 Freiberg (Germany)

    2016-04-15

    The present study is motivated by the increasing demand to consider internal partitioning into tissues instead of exposure concentrations for the environmental toxicity assessment. To this end, physiologically based pharmacokinetic (PBPK) models can be applied. We evaluated the variation in accuracy of PBPK model outcomes depending on tissue constituents modeled as sorptive phases and chemical distribution tendencies addressed by molecular descriptors. The model performance was examined using data from 150 experiments for 28 chemicals collected from US EPA databases. The simplest PBPK model is based on the “K{sub ow}-lipid content” approach as being traditional for environmental toxicology. The most elaborated one considers five biological sorptive phases (polar and non-polar lipids, water, albumin and the remaining proteins) and makes use of LSER (linear solvation energy relationship) parameters to describe the compound partitioning behavior. The “K{sub ow}-lipid content”-based PBPK model shows more than one order of magnitude difference in predicted and measured values for 37% of the studied exposure experiments while for the most elaborated model this happens only for 7%. It is shown that further improvements could be achieved by introducing corrections for metabolic biotransformation and compound transmission hindrance through a cellular membrane. The analysis of the interface distribution tendencies shows that polar tissue constituents, namely water, polar lipids and proteins, play an important role in the accumulation behavior of polar compounds with H-bond donating functional groups. For compounds without H-bond donating fragments preferable accumulation phases are storage lipids and water depending on compound polarity. - Highlights: • For reliable predictions, models of a certain complexity should be compared. • For reliable predictions non-lipid fish tissue constituents should be considered. • H-donor compounds preferably accumulate in water

  9. International variation in adherence to referral guidelines for suspected cancer: a secondary analysis of survey data

    Science.gov (United States)

    Nicholson, Brian D; Mant, David; Neal, Richard D; Hart, Nigel; Hamilton, Willie; Shinkins, Bethany; Rubin, Greg; Rose, Peter W

    2016-01-01

    Background Variation in cancer survival persists between comparable nations and appears to be due, in part, to primary care practitioners (PCPs) having different thresholds for acting definitively in response to cancer-related symptoms. Aim To explore whether cancer guidelines, and adherence to them, differ between jurisdictions and impacts on PCPs’ propensity to take definitive action on cancer-related symptoms. Design and setting A secondary analysis of survey data from six countries (10 jurisdictions) participating in the International Cancer Benchmarking Partnership. Method PCPs’ responses to five clinical vignettes presenting symptoms and signs of lung (n = 2), colorectal (n = 2), and ovarian cancer (n = 1) were compared with investigation and referral recommendations in cancer guidelines. Results Nine jurisdictions had guidelines covering the two colorectal vignettes. For the lung vignettes, although eight jurisdictions had guidelines for the first, the second was covered by a Swedish guideline alone. Only the UK and Denmark had an ovarian cancer guideline. Survey responses of 2795 PCPs (crude response rate: 12%) were analysed. Guideline adherence ranged from 20–82%. UK adherence was lower than other jurisdictions for the lung vignette covered by the guidance (47% versus 58%; P <0.01) but similar (45% versus 46%) or higher (67% versus 38%; P <0.01) for the two colorectal vignettes. PCPs took definitive action least often when a guideline recommended a non-definitive action or made no recommendation. UK PCPs adhered to recommendations for definitive action less than their counterparts (P <0.01). There wasno association between jurisdictional guideline adherence and 1-year survival. Conclusion Cancer guideline content is variable between similarly developed nations and poor guideline adherence does not explain differential survival. Guidelines that fail to cover high-risk presentations or that recommend non-definitive action may reduce definitive

  10. Variation in predicted internal concentrations in relation to PBPK model complexity for rainbow trout

    International Nuclear Information System (INIS)

    Salmina, E.S.; Wondrousch, D.; Kühne, R.; Potemkin, V.A.; Schüürmann, G.

    2016-01-01

    The present study is motivated by the increasing demand to consider internal partitioning into tissues instead of exposure concentrations for the environmental toxicity assessment. To this end, physiologically based pharmacokinetic (PBPK) models can be applied. We evaluated the variation in accuracy of PBPK model outcomes depending on tissue constituents modeled as sorptive phases and chemical distribution tendencies addressed by molecular descriptors. The model performance was examined using data from 150 experiments for 28 chemicals collected from US EPA databases. The simplest PBPK model is based on the “K_o_w-lipid content” approach as being traditional for environmental toxicology. The most elaborated one considers five biological sorptive phases (polar and non-polar lipids, water, albumin and the remaining proteins) and makes use of LSER (linear solvation energy relationship) parameters to describe the compound partitioning behavior. The “K_o_w-lipid content”-based PBPK model shows more than one order of magnitude difference in predicted and measured values for 37% of the studied exposure experiments while for the most elaborated model this happens only for 7%. It is shown that further improvements could be achieved by introducing corrections for metabolic biotransformation and compound transmission hindrance through a cellular membrane. The analysis of the interface distribution tendencies shows that polar tissue constituents, namely water, polar lipids and proteins, play an important role in the accumulation behavior of polar compounds with H-bond donating functional groups. For compounds without H-bond donating fragments preferable accumulation phases are storage lipids and water depending on compound polarity. - Highlights: • For reliable predictions, models of a certain complexity should be compared. • For reliable predictions non-lipid fish tissue constituents should be considered. • H-donor compounds preferably accumulate in water, polar

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

    International Nuclear Information System (INIS)

    Chapin, D.L.; Green, L.; Lee, A.Y.; Culbert, M.E.; Kelly, J.L.

    1979-09-01

    The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO 2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li 2 O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

  12. Thermomechanical analysis of the DFLL test blanket module for ITER

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

    The finite element code is used to simulate two kinds of blanket design structure, which are SLL (Quasi-Static Lithium Lead) and DLL (Dual-cooled Lithium Lead) blanket concepts for the Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM) submitted to the ITER test blanket working group. The temperature and stress distributions have been presented for the two kinds of blanket structure on the basis of the structural design, thermal-hydraulic design and neutronics analysis. Also the mechanical performance is presented for the high temperature component of blanket structure according to the ITER Structural Design Criteria (ISDC). The rationality and feasibility of the two kinds of blanket structure design of DFLL-TBM have been analyzed based on the above results which also acted as the theoretical base for further optimized analysis. (authors)

  13. Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Testoni, P., E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy, Josep Plá n. 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Albanese, R. [Association Euratom/ENEA/CREATE, DIEL, Università Federico II di Napoli, Napoli 80125 (Italy); Lucca, F.; Roccella, M. [L.T. Calcoli S.a.S. Piazza Prinetti, 26/B, Merate, Lecco (Italy); Portone, A. [Fusion for Energy, Josep Plá n. 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Rubinacci, G. [Association Euratom/ENEA/CREATE, DIEL, Università Federico II di Napoli, Napoli 80125 (Italy); Ventre, S.; Villone, F. [Association Euratom/ENEA/CREATE, DAEIMI, Università di Cassino, Cassino 03043 (Italy)

    2013-10-15

    Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained.

  14. 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)

  15. Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

    International Nuclear Information System (INIS)

    Testoni, P.; Albanese, R.; Lucca, F.; Roccella, M.; Portone, A.; Rubinacci, G.; Ventre, S.; Villone, F.

    2013-01-01

    Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained

  16. Preliminary study of a blanket handling device and evaluation of the feasibility of eliminating the spread of radioactive contamination

    International Nuclear Information System (INIS)

    Leger, D.; Djerassi, H.; Maupou, M.; Charruyer, P.; Salpietro, E.

    1988-01-01

    A study concerning progress and future development of the BLANKET HANDLING DEVICE of NET-DN tokamak and the related potentialities against contamination dispersal during handling of internal segments. To prevent the dust dispersion during the mantainance operations, there are three options: a Tight-Intermediate Containment (TIC), a Containment Transfer Unit (CTU) or the dust fixation on the internal components. The design of the BHD takes account of multivarious dimensioning requirements (geometrical and dimensional constraints, including characteristics of the segments and torus), environmental and operational constraints (safety, lifetime, maintainability, cooling of Blanket segments, containment). The possible solutions concerning protection of special devices, during handling and travelling, are discussed

  17. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

  18. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

    The present report summarizes the design activities of the ITER first wall and shielding blanket conducted by the JA Home Team during this year (1994) in close contact with the JCT, and reported during the four Technical Meetings held at Garching ITER Co-center. These activities are based on the Task Agreement between the JCT and the JA Home Team. In the present report, a layered configuration composed of separate first walls, modular-type blanket modules and separate back plates has been proposed to realize reliable assembly and maintenance schemes as well as to realize reliable component designs under high surface heat loads, high neutron wall loading and electromagnetic loads during disruptions. Outline of the structural design, consideration on fabricability and maintainability, and the results of thermal, mechanical and electromagnetic analyses are described. (author)

  19. Flow balancing in liquid metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Morley, N.B.

    1995-01-01

    Non-uniform flow distribution between parallel channels is one of the most serious concerns for self-cooled liquid metal blankets with electrically insulated walls. We show that uncertainties in flow distribution can be dramatically reduced by relatively simple design modifications. Several design features which impose flow uniformity by electrically coupling parallel channels are surveyed. Basic mechanisms for ''flow balancing'' are described, and a particular self-regulating concept using discrete passive electrodes is proposed for the US ITER advanced blanket concept. Scoping calculations suggest that this simple technique can be very powerful in equalizing the flow, even with massive insulator failures in individual channels. More detailed analyses and experimental verification will be required to demonstrate this concept for ITER. (orig.)

  20. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li 2 ZrO 3 was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li 2 O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D- 3 He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view

  1. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

    Wei, Jianghua; Song, Yuntao; Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong

    2015-01-01

    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.

  2. 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.

  3. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Waganer, L.M.

    1985-01-01

    The economic impact of employing the highly ranked blankets in the Blanket Comparison and Selection Study (BCSS) was evaluated in the context of both a tokamak and a tandem mirror power reactor (TMR). The economic evaluation criterion was determined to be the cost of electricity. The influencing factors that were considered are the direct cost of the blankets and related systems; the annual cost of blanket replacement; and the performance of the blanket, heat transfer, and energy conversion systems. The technical and cost bases for comparison were those of the STARFIRE and Mirror Advanced Reactor Study conceptual design power plants. The economic evaluation results indicated that the nitrate-salt-cooled blanket concept is an economically attractive concept for either reactor type. The water-cooled, solid breeder blanket is attractive for the tokamak and somewhat less attractive for the TMR. The helium-cooled, liquidlithium breeder blanket is the least economically desirable of higher ranked concepts. The remaining self-cooled liquid-metal and the helium-cooled blanket concepts represent moderately attractive concepts from an economic standpoint. These results are not in concert with those found in the other BCSS evaluation areas (engineering feasibility, safety, and research and development (R and D) requirements). The blankets faring well economically had generally lower cost components, lower pumping power requirements, and good power production capability. On the other hand, helium- and lithium-cooled systems were preferred from the standpoints of safety, engineering feasibility, and R and D requirements

  4. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    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 2 and a particle heat flux of 1 MW/m 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

  5. Nuclear Analysis of an ITER Blanket Module

    Science.gov (United States)

    Chiovaro, P.; Di Maio, P. A.; Parrinello, V.

    2013-08-01

    ITER blanket system is the reactor's plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.

  6. 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.

  7. Effect of nature convection on heat transfer in the liquid LiPb blanket for FDS-II

    Energy Technology Data Exchange (ETDEWEB)

    Wang Hongyan; Chen Hongli [Huaibei Coal Industry Teachers Coll. (China). Dept. of Physics; Zhou Tao [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics

    2007-07-01

    The He-cooled liquid LiPb tritium breeder (SLL) blanket concept is one of options of the blanket design of the fusion power reactor (FDS-II). The SLL blanket could be developed relatively easily with lower LiPb outlet temperature and slower LiPb flow velocity that allows the utilization of relatively mature material technology. The velocity of the liquid LiPb in the blanket is very slowly only in order to extract tritium. The magnetohydrodynamic (MHD) flow and heat transfer become very complex resulting from the differential heating of walls of the channels, especially adjacent to the First Wall (FW), and internal heat sources inside of the liquid LiPb. It is necessary to analyse the effect of the buoyancy-driven LiPb MHD flow on heat transfer in the channels with electrically and thermally conducting walls adjacent to the FW. The nature convection of the liquid LiPb, due to thermal diffusion, in the poloidal channel adjacent to the FW in the presence of the strong magnetic field of the SLL blanket has been considered and studied. The specially numerical MHD code based on the computational fluid dynamic software has been developed for analysis of the buoyancy-driven MHD flow. The properties of buoyantly convective flows have been investigated for various thermal boundary conditions. The numerical analysis was performed for the effect of nature convection on heat transfer of the liquid LiPb MHD flow in the poloidal channel in the SLL blanket. For the strong temperature gradient in the blanket and internal heat flux of Liquid LiPb, the three-dimensional temperature distributions of the LiPb, the FW and other walls have been given. Finally, The effect of the ratio of MHD buoyancy on the heat transfer characteristics of the LiPb flow have been calculated and presented. (orig.)

  8. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

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

  9. 30th International School of Mathematics "G Stampacchia" : Equilibrium Problems and Variational Models "Ettore Majorana"

    CERN Document Server

    Giannessi, Franco; Maugeri, Antonino; Equilibrium Problems and Variational Models

    2000-01-01

    The volume, devoted to variational analysis and its applications, collects selected and refereed contributions, which provide an outline of the field. The meeting of the title "Equilibrium Problems and Variational Models", which was held in Erice (Sicily) in the period June 23 - July 2 2000, was the occasion of the presentation of some of these papers; other results are a consequence of a fruitful and constructive atmosphere created during the meeting. New results, which enlarge the field of application of variational analysis, are presented in the book; they deal with the vectorial analysis, time dependent variational analysis, exact penalization, high order deriva­ tives, geometric aspects, distance functions and log-quadratic proximal methodology. The new theoretical results allow one to improve in a remarkable way the study of significant problems arising from the applied sciences, as continuum model of transportation, unilateral problems, multicriteria spatial price models, network equilibrium...

  10. Sensitivities of Internal Combustion Automotive Engines to Variations in Fuel Properties

    Science.gov (United States)

    1982-02-01

    An assessment of the sensitivity of the automotive gasoline and diesel engines to variations in fuel properties has been made. The variables studied include H/C ratio, distillation range, aromatic content, ignition quality as determined by the octane...

  11. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

    Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

  12. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1980-01-01

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  13. Processing and waste disposal representative for fusion breeder blanket systems

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1987-01-01

    This study is an evaluation of the waste handling concepts applicable to fusion breeder systems. Its goal is to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under US Nuclear Regulatory regulations. The radionuclides expected in the materials used in fusion reactor blankets are described, as are plans for reprocessing and disposal of the components of different breeder blankets. An estimate of the operating costs involved in waste disposal is made

  14. Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) 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 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.

  15. Electromagnetic analysis of ITER shield blanket under VDE

    International Nuclear Information System (INIS)

    Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

    2010-01-01

    Electromagnetic force and torque of ITER shield blanket system and their surrounding major component under vertical displacement event (VDE) were calculated with finite element method. ANSYS APDL was used to simulate the shape and magnitude of plasmas current dynamically in the VDE course, and external magnetic field was imposed, then the induced current distribution inside the all conductor including the blanket was obtained from the calculation. The force and torque for every blanket module was obtained to assess the safety of blanket system under VDE. (authors)

  16. Minimum thickness blanket-shield for fusion reactors

    International Nuclear Information System (INIS)

    Karni, Y.; Greenspan, E.

    1989-01-01

    A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

  17. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

  18. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

  19. Child Maltreatment, Inflammation, and Internalizing Symptoms: Investigating the Roles of C-Reactive Protein, Gene Variation and Neuroendocrine Regulation

    Science.gov (United States)

    Cicchetti, Dante; Handley, Elizabeth D.; Rogosch, Fred A.

    2015-01-01

    Prior research has found inconsistent evidence regarding the association among childhood adversity, inflammation, and internalizing symptoms, perhaps because previous studies have yet to adequately integrate important factors such as the timing of the adversity, genetic variation, and other relevant processes such as neuroendocrine regulation. The aims of the present study were threefold: 1) Determine whether the effect of the timing of child maltreatment on C-reactive protein (CRP), an inflammatory marker, varies by CRP gene variation; 2) Explore whether links between salivary CRP and childhood internalizing symptoms depend on the presence and timing of maltreatment experiences; 3) Investigate the role of CRP in the relations between child neuroendocrine regulation and internalizing symptoms and examine whether these associations are moderated by the presence and timing of child maltreatment. Participants included a sample of 267 maltreated and 222 nonmaltreated children (M age= 9.72, SD=0.99; 52.4% male; 66% African-American) who attended a summer day camp research program designed for school-aged low-income children. Department of Human Services records were examined to determine the onset and recency of maltreatment for children in the maltreated group. Results indicated that among children with recent onset maltreatment, those with at least one A allele from CRP SNP rs1417938 evidenced significantly higher CRP levels compared to recently maltreated children carrying the TT genotype. Moreover, higher levels of CRP were associated with higher levels of internalizing symptoms only for recently maltreated children. Finally, we did not find support for salivary CRP as a mechanism in the relation between neuroendocrine regulation and childhood internalizing symptoms. Our findings highlight the importance of the timing of child maltreatment and have important implications for characterizing variability in inflammation and internalizing symptoms among youth. PMID

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

    International Nuclear Information System (INIS)

    Finn, P.A.

    1985-01-01

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

  1. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

    Terai, T.; Tanaka, S.; Sze, D.-K.

    2000-01-01

    Force-free helical reactor, FFHR, is a demo-relevant heliotron-type D-T fusion reactor based on the great amount of R and D results obtained in the LHD project. Since 1993, collaboration works have made great progress in design studies of FFHR with standing on the major advantage of current-less steady operation with no dangerous plasma disruptions. There are two types of reference designs, FFHR-1 and FFHR-2, where molten Flibe (LiF-BeF2) is utilized as tritium breeder and coolant. In this paper, we present the outline of FFHR blanket design and some related R and D topics focusing on Be utilization. Beryllium is used as a neutron multiplier in the design and Be pebbles are placed in the front part of the tritium breeding zone. In a Flibe blanket, HF (TF) generated due to nuclear transmutation will be a problem because of its corrosive property. Though nickel-based alloys are thought to be intact in such a corrosive environment, FFHR blanket design does not adopt the alloys because of their induced radioactivity. The present candidate materials for the structure are low-activated ferritic steel (JLF-1), V-4Cr-4Ti, etc. They are capable to be corroded by HF in the operation condition, and Be is expected to work as a reducing agent in the system as well. Whether Be pebbles placed in a Flibe flow can work well or not is a very important matter. From this point, Be solubility in Flibe, reaction rate of the Redox reaction with TF in the liquid and on the surface of Be pebbles under irradiation, flowing behavior of Flibe through a Be pebble bed, etc. should be investigated. In 1997, in order to establish more practical and new data bases for advanced design works, we started a collaboration work of R and D on blanket engineering, where the Be research above mentioned is included. Preliminary dipping-test of Be sheets and in-situ tritium release experiment from Flibe with Be sheets have got started. (orig.)

  2. Structural materials for fusion reactor blanket systems

    International Nuclear Information System (INIS)

    Bloom, E.E.; Smith, D.L.

    1984-01-01

    Consideration of the required functions of the blanket and the general chemical, mechanical, and physical properties of candidate tritium breeding materials, coolants, structural materials, etc., leads to acceptable or compatible combinations of materials. The presently favored candidate structural materials are the austenitic stainless steels, martensitic steels, and vanadium alloys. The characteristics of these alloy systems which limit their application and potential performance as well as approaches to alloy development aimed at improving performance (temperature capability and lifetime) will be described. Progress towards understanding and improving the performance of structural materials has been substantial. It is possible to develop materials with acceptable properties for fusion applications

  3. About possible technologies of creation nanostructures blankets

    International Nuclear Information System (INIS)

    Blednova, Zh.M.; Chaevskij, M.I.; Rusinov, P.O.

    2008-01-01

    Possible technologies of formation nanostructures blankets are considered: a method of thermal carrying over of weights in the conditions of a high gradient of temperatures; the combined method including cathode-plasma nitriding in the conditions of low pressure and drawing of nitride of the titan in a uniform work cycle; the combined method including high-frequency ionic nitriding and drawing of carbide of chrome by pyrolysis chrome and organic of connections in plasma of the decaying category. Possibility of formation layered nanostructures layers is shown.

  4. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

    Billone, M.C.; Dienst, W.; Noda, K.; Roux, N.

    1993-11-01

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

  5. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-1 through 4 and PULSAR 1 and 2. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. Also, the requirements of engineering and physics systems for a pulsed reactor were evaluated by the PULSAR design studies. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies

  6. Modelling of sludge blanket height and flow pattern in UASB reactors treating municipal wastewater

    International Nuclear Information System (INIS)

    Singh, K.S.; Viraraghavan, T.

    2002-01-01

    Two upflow anaerobic sludge blanket (UASB) reactors were started-up and operated for approximately 900 days to examine the feasibility of treating municipal wastewater under low temperature conditions. A modified solid distribution model was formulated by incorporating the variation of biogas production rate with a change in temperature. This model was used to optimize the sludge blanket height of UASB reactors for an effective operation of gas-liquid-solid (GLS) separation device. This model was found to simulate well the solid distribution as confirmed experimental observation of solid profile along the height of the reactor. Mathematical analysis of tracer curves indicated the presence of a mixed type of flow pattern in the sludge-bed zone of the reactor. It was found that the dead-zone and by-pass flow fraction were impacted by the change in operating temperatures. (author)

  7. Internal Variations in Health-care Federalism in Canada and the United States

    Directory of Open Access Journals (Sweden)

    Vandna Bhatia

    2014-12-01

    Full Text Available Federal systems are prone to dividing health benefits inconsistently across subnational jurisdictions. In this article, we examine how federalism intersects with economic and social factors, particularly gender and immigration status, to create structural barriers to accessing and receiving necessary healthcare. Drawing on insights from the historical institutionalist literature and the experiences of immigrant women in the Canadian and American health systems, we find significant subnational variations in access to health services and insurance coverage. Gaps in service – which are filled (if at all by costly, inaccessible private provision – are the product of piecemeal policymaking, as new programs and services are layered onto existing systems which are themselves outdated and anachronistic. Our analysis demonstrates the need to move beyond analyses of federal state architectures to an intersectional approach to better understand the differential negative impact of subnational variations on equity between social groups and their ability to access to basic health services.

  8. Failure maps for internally pressurized Zr-2.5% Nb pressure tubes with circumferential temperature variations

    International Nuclear Information System (INIS)

    Shewfelt, R.S.W.

    1986-01-01

    During some postulated loss-of-coolant accidents, the pressure tube temperature may rise before the internal pressure drops, causing the pressure tube to balloon. The temperature around the pressure tube circumference would likely be nonuniform, producing localized deformation that could possibly cause failure. The computer program, GRAD, was used to determine the circumferential temperature distribution required to cause an internally pressurized Zr-2.5% Nb pressure tube to fail before coming into full contact with its calandria tube. These results were used to construct failure maps. 7 refs

  9. International variation in management of screen-detected ductal carcinoma in situ of the breast

    DEFF Research Database (Denmark)

    Ponti, Antonio; Lynge, Elsebeth; James, Ted

    2014-01-01

    BACKGROUND: Ductal carcinoma in situ (DCIS) incidence has grown with the implementation of screening and its detection varies across International Cancer Screening Network (ICSN) countries. The aim of this survey is to describe the management of screen-detected DCIS in ICSN countries and to evalu...

  10. A statistical study of variations of internal gravity wave energy characteristics in meteor zone

    Science.gov (United States)

    Gavrilov, N. M.; Kalov, E. D.

    1987-01-01

    Internal gravity wave (IGW) parameters obtained by the radiometer method have been considered by many other researchers. The results of the processing of regular radiometeor measurements taken during 1979 to 1980 in Obninsk (55.1 deg N, 36.6 deg E) are presented.

  11. Situational Variations in Request and Apology Realization Strategies among International Postgraduate Students at Malaysian Universities

    Science.gov (United States)

    Muthusamy, Paramasivam; Farashaiyan, Atieh

    2016-01-01

    The present study attempted to describe the request, apology, and request mitigation strategies utilized by international postgraduate students in confronting different situations. In addition, it examined the effects of the situational factors of social distance, power, and imposition on the students' choice of request and apology strategies as…

  12. Two-phase-flow cooling concept for fusion reactor blankets

    International Nuclear Information System (INIS)

    Bender, D.J.; Hoffman, M.A.

    1977-01-01

    The new two-phase heat transfer medium proposed is a mixture of potassium droplets and helium which permits blanket operation at hih temperature and low pressure, while maintaining acceptable pumping power requirements, coolant ducting size, and blanket structure fractions. A two-phase flow model is described. The helium pumping power and the primary heat transfer loop are discussed

  13. Overview of the TFTB lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

    The Lithium Blanket Module (LBM) is an ∼ 80-cm 3 module, representative of a helium-cooled lithium oxide fusion reactor blanket module. This paper summarizes the design, development, and construction of the LBM, and indicates the present status of the LBM program

  14. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

    Kakudate, S.; Nakahira, M.; Oka, K.; Taguchi, K.; Obara, K.; Tada, E.; Shibanuma, K.; Tesini, A.; Haange, R.; Maisonnier, D.

    2001-01-01

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

  15. Accelerator driven heavy water blanket on circulating fuel

    International Nuclear Information System (INIS)

    Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

    1997-01-01

    A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

  16. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2005-03-01

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

  17. An assessment of the base blanket for ITER

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  18. Objectives and status of EUROfusion DEMO blanket studies

    Energy Technology Data Exchange (ETDEWEB)

    Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

    2016-11-01

    Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

  19. 18 CFR 284.303 - OCS blanket certificates.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false OCS blanket certificates. 284.303 Section 284.303 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... Pipelines on Behalf of Others § 284.303 OCS blanket certificates. Every OCS pipeline [as that term is...

  20. The TFTR lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bertone, P.C.; Creedon, R.L.; File, J.; Graumann, D.W.

    1985-01-01

    The Lithium Blanket Module (LBM) is an approximately 80X80X80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li 2 O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li 2 O pellets with satisfactory reproducibility were developed using purified Li 2 O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1-1nCi/g)

  1. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

    Cheng, E.T.; Engholm, B.A.; Su, S.D.

    1981-01-01

    The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design

  2. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

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

  3. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

    In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

  4. Neutron dosimetry for the TFTR Lithium-Blanket-Module program

    International Nuclear Information System (INIS)

    Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

    1981-01-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory

  5. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

    1971-01-01

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  6. Self-cooled liquid-metal blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Arheidt, K.; Barleon, L.

    1988-01-01

    A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

  7. Spatial Variation of Diapycnal Diffusivity Estimated From Seismic Imaging of Internal Wave Field, Gulf of Mexico

    OpenAIRE

    Dickinson, Nicholas; White, Nicholas Jeremiah; Caulfield, Colm-cille Patrick

    2017-01-01

    Bright reflections are observed within the upper 1000~m of the water column along a seismic reflection profile that traverses the northern margin of the Gulf of Mexico. Independent hydrographic calibration demonstrates that these reflections are primarily caused by temperature changes associated with different water masses that are entrained into the Gulf along the Loop Current. The internal wave field is analyzed by automatically tracking 1171 reflections, each of which is greater th...

  8. Towards better description of solar activity variation in the International Reference Ionosphere topside ion composition model

    Czech Academy of Sciences Publication Activity Database

    Truhlík, Vladimír; Bilitza, D.; Třísková, Ludmila

    2015-01-01

    Roč. 55, č. 8 (2015), s. 2099-2105 ISSN 0273-1177 R&D Projects: GA MŠk(CZ) LH11123 Institutional support: RVO:68378289 Keywords : ion composition * topside ionosphere * solar activity * empirical model * International Reference Ionosphere Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.409, year: 2015 http://www.sciencedirect.com/science/article/pii/S027311771400489X

  9. Variations of the internal pudendal artery as a congenital contributing factor to age at onset of erectile dysfunction in Japanese.

    Science.gov (United States)

    Kawanishi, Yasuo; Muguruma, Hiroshi; Sugiyama, Hiroaki; Kagawa, Junichirou; Tanimoto, Syuji; Yamanaka, Masahito; Kojima, Keiji; Numata, Akira; Kishimoto, Tomoteru; Nakanishi, Ryoichi; Kanayama, Hiro-omi

    2008-03-01

    To investigate the relationship between variations of the pelvic artery arrangement and the age at erectile dysfunction (ED) onset, as some men develop ED while relatively young, while others maintain erectile function into old age despite having cardiovascular diseases, thus congenital factors might be involved. We examined 290 units of internal iliac arteries (IIA) in 145 patients showing repeated incomplete erectile response to intracavernosal injections with prostaglandin E(1). Patients with cardiovascular risk factors, neurological disease or pelvic injury were excluded. The pelvic artery arrangement, evaluated by three-dimensional computed tomographic angiography, was classified anatomically into five types: Type 1 (normal or basic type), in which the internal pudendal artery (IPA) originates from the anterior trunk at the level between the linea terminalis and the major ischial notch; Type 2, the IPA originates from the anterior trunk of the IIA at the level of the major ischial notch or more distally; Type 3, the IPA originates directly from the IIA at a level proximal to the linea terminalis; Type 4, the IPA originates together with the superior and inferior gluteal artery within 1 cm of each other; and Type 5, the penile blood supply is dependent on arteries other than the IPA, such as the obturator artery. Among the 290 units, eight could not be classified due to poor image quality. There were no statistically significant differences in blood flow parameters among the types of IIAs, but there was a statistically significant difference in the IPA type at the age of onset of ED. Type 1 (153 units or 53%) anatomy, was more common in patients who developed ED at an advanced age. Types 2, 3 and 4 were more common in patients with onset of ED at an early age (log-rank test P < 0.001, P = 0.044, P < 0.001, respectively). Compared with patients with the common type of IIAs bilaterally, patients with any of the variations bilaterally are at risk of early onset of

  10. The international Genome sample resource (IGSR): A worldwide collection of genome variation incorporating the 1000 Genomes Project data.

    Science.gov (United States)

    Clarke, Laura; Fairley, Susan; Zheng-Bradley, Xiangqun; Streeter, Ian; Perry, Emily; Lowy, Ernesto; Tassé, Anne-Marie; Flicek, Paul

    2017-01-04

    The International Genome Sample Resource (IGSR; http://www.internationalgenome.org) expands in data type and population diversity the resources from the 1000 Genomes Project. IGSR represents the largest open collection of human variation data and provides easy access to these resources. IGSR was established in 2015 to maintain and extend the 1000 Genomes Project data, which has been widely used as a reference set of human variation and by researchers developing analysis methods. IGSR has mapped all of the 1000 Genomes sequence to the newest human reference (GRCh38), and will release updated variant calls to ensure maximal usefulness of the existing data. IGSR is collecting new structural variation data on the 1000 Genomes samples from long read sequencing and other technologies, and will collect relevant functional data into a single comprehensive resource. IGSR is extending coverage with new populations sequenced by collaborating groups. Here, we present the new data and analysis that IGSR has made available. We have also introduced a new data portal that increases discoverability of our data-previously only browseable through our FTP site-by focusing on particular samples, populations or data sets of interest. © The Author(s) 2016. Published by Oxford University Press on behalf of Nucleic Acids Research.

  11. Barcoding the Dendrobium (Orchidaceae Species and Analysis of the Intragenomic Variation Based on the Internal Transcribed Spacer 2

    Directory of Open Access Journals (Sweden)

    Xiaoyue Wang

    2017-01-01

    Full Text Available Many species belonging to the genus Dendrobium are of great commercial value. However, their difficult growth conditions and high demand have caused many of these species to become endangered. Indeed, counterfeit Dendrobium products are common, especially in medicinal markets. This study aims to assess the suitability of the internal transcribed spacer 2 (ITS2 region as a marker for identifying Dendrobium and to evaluate its intragenomic variation in Dendrobium species. In total, 29,624 ITS2 copies from 18 species were obtained using 454 pyrosequencing to evaluate intragenomic variation. In addition, 513 ITS2 sequences from 26 Dendrobium species were used to assess its identification suitability. The highest intragenomic genetic distance was observed in Dendrobium chrysotoxum (0.081. The average intraspecific genetic distances of each species ranged from 0 to 0.032. Phylogenetic trees based on ITS2 sequences showed that most Dendrobium species are monophyletic. The intragenomic and intraspecies divergence analysis showed that greater intragenomic divergence is mostly correlated with larger intraspecific variation. As a major ITS2 variant becomes more common in genome, there are fewer intraspecific variable sites in ITS2 sequences at the species level. The results demonstrated that the intragenomic multiple copies of ITS2 did not affect species identification.

  12. Barcoding the Dendrobium (Orchidaceae) Species and Analysis of the Intragenomic Variation Based on the Internal Transcribed Spacer 2.

    Science.gov (United States)

    Wang, Xiaoyue; Chen, Xiaochen; Yang, Pei; Wang, Lili; Han, Jianping

    2017-01-01

    Many species belonging to the genus Dendrobium are of great commercial value. However, their difficult growth conditions and high demand have caused many of these species to become endangered. Indeed, counterfeit Dendrobium products are common, especially in medicinal markets. This study aims to assess the suitability of the internal transcribed spacer 2 (ITS2) region as a marker for identifying Dendrobium and to evaluate its intragenomic variation in Dendrobium species. In total, 29,624 ITS2 copies from 18 species were obtained using 454 pyrosequencing to evaluate intragenomic variation. In addition, 513 ITS2 sequences from 26 Dendrobium species were used to assess its identification suitability. The highest intragenomic genetic distance was observed in Dendrobium chrysotoxum (0.081). The average intraspecific genetic distances of each species ranged from 0 to 0.032. Phylogenetic trees based on ITS2 sequences showed that most Dendrobium species are monophyletic. The intragenomic and intraspecies divergence analysis showed that greater intragenomic divergence is mostly correlated with larger intraspecific variation. As a major ITS2 variant becomes more common in genome, there are fewer intraspecific variable sites in ITS2 sequences at the species level. The results demonstrated that the intragenomic multiple copies of ITS2 did not affect species identification.

  13. Spatial Variation of Diapycnal Diffusivity Estimated From Seismic Imaging of Internal Wave Field, Gulf of Mexico

    Science.gov (United States)

    Dickinson, Alex; White, N. J.; Caulfield, C. P.

    2017-12-01

    Bright reflections are observed within the upper 1,000 m of the water column along a seismic reflection profile that traverses the northern margin of the Gulf of Mexico. Independent hydrographic calibration demonstrates that these reflections are primarily caused by temperature changes associated with different water masses that are entrained into the Gulf along the Loop Current. The internal wave field is analyzed by automatically tracking 1,171 reflections, each of which is greater than 2 km in length. Power spectra of the horizontal gradient of isopycnal displacement, ϕξx, are calculated from these tracked reflections. At low horizontal wave numbers (kxcpm), ϕξx∝kx-0.2±0.6, in agreement with hydrographic observations of the internal wave field. The turbulent spectral subrange is rarely observed. Diapycnal diffusivity, K, is estimated from the observed internal wave spectral subrange of each tracked reflection using a fine-scale parametrization of turbulent mixing. Calculated values of K vary between 10-8 and 10-4 m2 s-1 with a mean value of K˜4×10-6 m2 s-1. The spatial distribution of turbulent mixing shows that K˜10-7 m2 s-1 away from the shelf edge in the upper 300 m where stratification is strong. Mixing is enhanced by up to 4 orders of magnitude adjacent to the shoaling bathymetry of the continental slope. This overall pattern matches that determined by analyzing nearby suites of CTD casts. However, the range of values recovered by spectral analysis of the seismic image is greater as a consequence of significantly better horizontal resolution.

  14. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

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

  15. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

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

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

    International Nuclear Information System (INIS)

    Ishitsuka, E.

    2002-01-01

    Advanced solid breeding blanket design in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high dose of neutron irradiation. Therefore, the development of such advanced blanket materials is indispensable. In this paper, the cooperation activities among JAERI, universities and industries in Japan on the development of these advanced materials are reported. Advanced tritium breeding material to prevent the grain growth in high temperature had to be developed because the tritium release behavior degraded by the grain growth. As one of such materials, TiO 2 -doped Li 2 TiO 3 has been studied, and TiO 2 -doped Li 2 TiO 3 pebbles was successfully fabricated. For the advanced neutron multiplier, the beryllium intermetallic compounds that have high melting point and good chemical stability have been studied. Some characterization of Be 12 Ti was studied. The pebble fabrication study for Be 12 Ti was also performed and Be 12 Ti pebbles were successfully fabricated. From these activities, the bright prospect to realize the DEMO blanket by the application of TiO 2 -doped Li 2 TiO 3 and beryllium intermetallic compounds was obtained. (author)

  17. 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.

  18. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    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

  19. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

    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

  20. Neutronics analysis for aqueous self-cooled fusion reactor blankets

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Jaffa, R.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1986-06-01

    The tritium breeding performance of several Aqueous Self-Cooled Blanket (ASCB) configurations for fusion reactors has been evaluated. The ASCB concept employs small amounts of lithium compound dissolved in light or heavy water to serve as both coolant and breeding medium. The inherent simplicity of this concept allows the development of blankets with minimal technological risk. The tritium breeding performance of the ASCB concept is a critical issue for this family of blankets. Contrary to conventional blanket designs there will be a significant contribution to the tritium breeding ratio (TBR) in the water coolant/breeder of duct shields, and the 3-D TBR will therefore be similar to the 1-D TBR. The tritium breeding performance of an ASCB for a MARS-like-tandem reactor and an ASCB based breeding-shield for the Next European Torus (NET) are assessed. Two design options for the MARS-like blanket are discussed. One design employs a vanadium first wall, and zircaloy for the structural material. The trade-offs between light water and heavy water cooling options for this zircaloy blanket are discussed. The second design option for MARS relies on the use of a vanadium alloy as the stuctural material, and heavy water as the coolant. It is demonstrated that both design options lead to low-activation blankets that allow class C burial. The breeder-shield for NET consists of a water-cooled stainless steel shield

  1. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Naruse, Yuji; Yamaoka, Mitsuaki; Ohara, Atsushi; Ono, Kiyoshi; Kobayashi, Shigetada.

    1992-06-01

    Aqueous solution blanket using lithium salts such as LiNO 3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

  2. The effects of internal refractive index variation in near-infrared optical tomography: a finite element modelling approach

    International Nuclear Information System (INIS)

    Dehghani, Hamid; Brooksby, Ben; Vishwanath, Karthik; Pogue, Brian W; Paulsen, Keith D

    2003-01-01

    Near-infrared (NIR) tomography is a technique used to measure light propagation through tissue and generate images of internal optical property distributions from boundary measurements. Most popular applications have concentrated on female breast imaging, neonatal and adult head imaging, as well as muscle and small animal studies. In most instances a highly scattering medium with a homogeneous refractive index is assumed throughout the imaging domain. Using these assumptions, it is possible to simplify the model to the diffusion approximation. However, biological tissue contains regions of varying optical absorption and scatter, as well as varying refractive index. In this work, we introduce an internal boundary constraint in the finite element method approach to modelling light propagation through tissue that accounts for regions of different refractive indices. We have compared the results to data from a Monte Carlo simulation and show that for a simple two-layered slab model of varying refractive index, the phase of the measured reflectance data is significantly altered by the variation in internal refractive index, whereas the amplitude data are affected only slightly

  3. Advancement in tritium transport simulations for solid breeding blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Ying, Alice, E-mail: ying@fusion.ucla.edu [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Zhang, Hongjie [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Merrill, Brad J. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    In this paper, advancement on tritium transport simulations was demonstrated for a solid breeder blanket HCCR TBS, where multi-physics and detailed engineering descriptions are considered using a commercial simulation code. The physics involved includes compressible purge gas fluid flow, heat transfer, chemical reaction, isotope swamping effect, and tritium isotopes mass transport. The strategy adopted here is to develop numerical procedures and techniques that allow critical details of material, geometric and operational heterogeneity in a most complete engineering description of the TBS being incorporated into the simulation. Our application focuses on the transient assessment in view of ITER being pulsed operations. An immediate advantage is a more realistic predictive and design analysis tool accounting pulsed operations induced temperature variations which impact helium purge gas flow as well as Q{sub 2} composition concentration time and space evolutions in the breeding regions. This affords a more accurate prediction of tritium permeation into the He coolant by accounting correct temperature and partial pressure effects and realistic diffusion paths. The analysis also shows that by introducing by-pass line to accommodate ITER pulsed operations in the TES loop allows tritium extraction design being more cost effective.

  4. International variation in neighborhood walkability, transit, and recreation environments using geographic information systems: the IPEN adult study.

    Science.gov (United States)

    Adams, Marc A; Frank, Lawrence D; Schipperijn, Jasper; Smith, Graham; Chapman, James; Christiansen, Lars B; Coffee, Neil; Salvo, Deborah; du Toit, Lorinne; Dygrýn, Jan; Hino, Adriano Akira Ferreira; Lai, Poh-chin; Mavoa, Suzanne; Pinzón, José David; Van de Weghe, Nico; Cerin, Ester; Davey, Rachel; Macfarlane, Duncan; Owen, Neville; Sallis, James F

    2014-10-25

    The World Health Organization recommends strategies to improve urban design, public transportation, and recreation facilities to facilitate physical activity for non-communicable disease prevention for an increasingly urbanized global population. Most evidence supporting environmental associations with physical activity comes from single countries or regions with limited variation in urban form. This paper documents variation in comparable built environment features across countries from diverse regions. The International Physical Activity and the Environment Network (IPEN) study of adults aimed to measure the full range of variation in the built environment using geographic information systems (GIS) across 12 countries on 5 continents. Investigators in Australia, Belgium, Brazil, Colombia, the Czech Republic, Denmark, China, Mexico, New Zealand, Spain, the United Kingdom, and the United States followed a common research protocol to develop internationally comparable measures. Using detailed instructions, GIS-based measures included features such as walkability (i.e., residential density, street connectivity, mix of land uses), and access to public transit, parks, and private recreation facilities around each participant's residential address using 1-km and 500-m street network buffers. Eleven of 12 countries and 15 cities had objective GIS data on built environment features. We observed a 38-fold difference in median residential densities, a 5-fold difference in median intersection densities and an 18-fold difference in median park densities. Hong Kong had the highest and North Shore, New Zealand had the lowest median walkability index values, representing a difference of 9 standard deviations in GIS-measured walkability. Results show that comparable measures can be created across a range of cultural settings revealing profound global differences in urban form relevant to physical activity. These measures allow cities to be ranked more precisely than previously

  5. International variations and trends in renal cell carcinoma incidence and mortality.

    Science.gov (United States)

    Znaor, Ariana; Lortet-Tieulent, Joannie; Laversanne, Mathieu; Jemal, Ahmedin; Bray, Freddie

    2015-03-01

    Renal cell carcinoma (RCC) incidence rates are higher in developed countries, where up to half of the cases are discovered incidentally. Declining mortality trends have been reported in highly developed countries since the 1990s. To compare and interpret geographic variations and trends in the incidence and mortality of RCC worldwide in the context of controlling the future disease burden. We used data from GLOBOCAN, the Cancer Incidence in Five Continents series, and the World Health Organisation mortality database to compare incidence and mortality rates in more than 40 countries worldwide. We analysed incidence and mortality trends in the last 10 yr using joinpoint analyses of the age-standardised rates (ASRs). RCC incidence in men varied in ASRs (World standard population) from approximately 1/100,000 in African countries to >15/100,000 in several Northern and Eastern European countries and among US blacks. Similar patterns were observed for women, although incidence rates were commonly half of those for men. Incidence rates are increasing in most countries, most prominently in Latin America. Although recent mortality trends are stable in many countries, significant declines were observed in Western and Northern Europe, the USA, and Australia. Southern European men appear to have the least favourable RCC mortality trends. Although RCC incidence is still increasing in most countries, stabilisation of mortality trends has been achieved in many highly developed countries. There are marked absolute differences and opposing RCC mortality trends in countries categorised as areas of higher versus lower human development, and these gaps appear to be widening. Renal cell cancer is becoming more commonly diagnosed worldwide in both men and women. Mortality is decreasing in the most developed settings, but not in low- and middle-income countries, where access to and the availability of optimal therapies are likely to be limited. Copyright © 2014 European Association of

  6. International variations in clinical practice guidelines for palliative sedation: a systematic review.

    Science.gov (United States)

    Abarshi, Ebun; Rietjens, Judith; Robijn, Lenzo; Caraceni, Augusto; Payne, Sheila; Deliens, Luc; Van den Block, Lieve

    2017-09-01

    Palliative sedation is a highly debated medical practice, particularly regarding its proper use in end-of-life care. Worldwide, guidelines are used to standardise care and regulate this practice. In this review, we identify and compare national/regional clinical practice guidelines on palliative sedation against the European Association for Palliative Care (EAPC) palliative sedation Framework and assess the developmental quality of these guidelines using the Appraisal Guideline Research and Evaluation (AGREE II) instrument. Using the PRISMA criteria, we searched multiple databases (PubMed, CancerLit, CINAHL, Cochrane Library, NHS Evidence and Google Scholar) for relevant guidelines, and selected those written in English, Dutch and Italian; published between January 2000 and March 2016. Of 264 hits, 13 guidelines-Belgium, Canada (3), Ireland, Italy, Japan, the Netherlands, Norway, Spain, Europe, and USA (2) were selected. 8 contained at least 9/10 recommendations published in the EAPC Framework; 9 recommended 'pre-emptive discussion of the potential role of sedation in end-of-life care'; 9 recommended 'nutrition/hydration while performing sedation' and 8 acknowledged the need to 'care for the medical team'. There were striking differences in terminologies used and in life expectancy preceding the practice. Selected guidelines were conceptually similar, comparing closely to the EAPC Framework recommendations, albeit with notable variations. Based on AGREE II, 3 guidelines achieved top scores and could therefore be recommended for use in this context. Also, domains 'scope and purpose' and 'editorial independence' ranked highest and lowest, respectively-underscoring the importance of good reportage at the developmental stage. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

  7. Internal Variations in Empirical Oxygen Abundances for Giant H II Regions in the Galaxy NGC 2403

    Science.gov (United States)

    Mao, Ye-Wei; Lin, Lin; Kong, Xu

    2018-02-01

    This paper presents a spectroscopic investigation of 11 {{H}} {{II}} regions in the nearby galaxy NGC 2403. The {{H}} {{II}} regions are observed with a long-slit spectrograph mounted on the 2.16 m telescope at XingLong station of National Astronomical Observatories of China. For each of the {{H}} {{II}} regions, spectra are extracted at different nebular radii along the slit-coverage. Oxygen abundances are empirically estimated from the strong-line indices R23, N2O2, O3N2, and N2 for each spectrophotometric unit, with both observation- and model-based calibrations adopted into the derivation. Radial profiles of these diversely estimated abundances are drawn for each nebula. In the results, the oxygen abundances separately estimated with the prescriptions on the basis of observations and models, albeit from the same spectral index, systematically deviate from each other; at the same time, the spectral indices R23 and N2O2 are distributed with flat profiles, whereas N2 and O3N2 exhibit apparent gradients with the nebular radius. Because our study naturally samples various ionization levels, which inherently decline at larger radii within individual {{H}} {{II}} regions, the radial distributions indicate not only the robustness of R23 and N2O2 against ionization variations but also the sensitivity of N2 and O3N2 to the ionization parameter. The results in this paper provide observational corroboration of the theoretical prediction about the deviation in the empirical abundance diagnostics. Our future work is planned to investigate metal-poor {{H}} {{II}} regions with measurable T e, in an attempt to recalibrate the strong-line indices and consequently disclose the cause of the discrepancies between the empirical oxygen abundances.

  8. Timing is everything: international variations in historical sexual partnership concurrency and HIV prevalence.

    Directory of Open Access Journals (Sweden)

    Martina Morris

    Full Text Available BACKGROUND: Higher prevalence of concurrent partnerships is one hypothesis for the severity of the HIV epidemic in the countries of Southern Africa. But measures of the prevalence of concurrency alone do not adequately capture the impact concurrency will have on transmission dynamics. The importance of overlap duration and coital exposure are examined here. METHODOLOGY/PRINCIPAL FINDINGS: We conducted a comparison of data from three studies of sexual behavior carried out in the early 1990s in Uganda, Thailand and the US. Using cumulative concurrency measures, the three countries appeared somewhat similar. Over 50% of both Thai and Ugandan men reported a concurrency within the last three partnerships and over 20% reported a concurrency in the last year, the corresponding rates among US men were nearly 20% for Blacks and Hispanics, and about 10% for other racial/ethnic groups. Concurrency measures that were more sensitive to overlap duration, however, showed large differences. The point prevalence of concurrency on the day of interview was over 10% among Ugandan men compared to 1% for Thai men. Ugandan concurrencies were much longer duration - a median of about two years - than either the Thai (1 day or US concurrencies (4-9 months across all groups, and involved 5-10 times more coital risk exposure with the less frequent partner. In the US, Blacks and Hispanics reported higher prevalence, longer duration and greater coital exposure than Whites, but were lower than Ugandans on nearly every measure. Together, the differences in the prevalence, duration and coital exposure of concurrent partnerships observed align with the HIV prevalence differentials seen in these populations at the time the data were collected. CONCLUSIONS/SIGNIFICANCE: There were substantial variations in the patterns of concurrent partnerships within and between populations. More long-term overlapping partnerships, with regular coital exposure, were found in populations with

  9. Overview of EU activities on DEMO liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Malang, S.; Reimann, J.; Perujo, A.

    1994-01-01

    The present paper gives an overview of both design and experimental activities within the European Union (EU) concerning the development of liquid metal breeder blankets for DEMO. After several years of studies on breeding blankets, two blanket concepts are presently considered, both using the eutectic Pb-17Li: the dual-coolant concept and the water-cooled concept. The analysis of such concepts has permitted to identify the experimental areas where further data are required. Tritium control and MHD-issues are, at present, the activities on which is devoted the greatest effort within the EU. (authors). 4 figs., 4 tabs., 39 refs

  10. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-01-01

    The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

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

    International Nuclear Information System (INIS)

    Sze, D.K.; Cheng, E.T.

    1985-02-01

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

  12. Conceptual design of blanket structures for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

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

    International Nuclear Information System (INIS)

    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

  14. International variation in ethics committee requirements: comparisons across five Westernised nations

    Directory of Open Access Journals (Sweden)

    Nachson Israel

    2002-04-01

    Full Text Available Abstract Background Ethics committees typically apply the common principles of autonomy, nonmaleficence, beneficence and justice to research proposals but with variable weighting and interpretation. This paper reports a comparison of ethical requirements in an international cross-cultural study and discusses their implications. Discussion The study was run concurrently in New Zealand, UK, Israel, Canada and USA and involved testing hypotheses about believability of testimonies regarding alleged child sexual abuse. Ethics committee requirements to conduct this study ranged from nil in Israel to considerable amendments designed to minimise participant harm in New Zealand. Assessment of minimal risk is a complex and unreliable estimation further compounded by insufficient information on probabilities of particular individuals suffering harm. Estimating potential benefits/ risks ratio and protecting participants' autonomy similarly are not straightforward exercises. Summary Safeguarding moral/humane principles should be balanced with promotion of ethical research which does not impede research posing minimal risk to participants. In ensuring that ethical standards are met and research has scientific merit, ethics committees have obligations to participants (to meet their rights and protect them from harm; to society (to ensure good quality research is conducted; and to researchers (to treat their proposals with just consideration and respect. To facilitate meeting all these obligations, the preferable focus should be promotion of ethical research, rather than the prevention of unethical research, which inevitably results in the impediment of researchers from doing their work. How the ethical principles should be applied and balanced requires further consideration.

  15. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

    2016-01-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  16. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

    2016-11-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  17. Comparison over the nuclear analysis of the HCLL blanket for the European DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, Giacomo; Aubert, Julien [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [ENEA, UTFUS-TECN, Via E. Fermi 4, 00044 Frascati, Rome (Italy); Fischer, Ulrich [Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen, Karlsruhe (Germany)

    2016-11-01

    Highlights: • A complete nuclear analysis of the DEMO HCLL has been carried out at CEA with the TRIPOLI-4{sup ®} Monte Carlo code. • The DEMO tokamak model was generated by the CAD import tool McCad. • The HCLL blankets were implemented using a previous MCNP model developed at ENEA. • A good agreement is observed between the results obtained at CEA with TRIPOLI-4 and JEFF-3.1.1 and whose obtained at ENEA with MCNP and FENDL-2.1. - Abstract: This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4{sup ®} Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4{sup ®} representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4{sup ®} Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO.

  18. Comparison over the nuclear analysis of the HCLL blanket for the European DEMO

    International Nuclear Information System (INIS)

    Jaboulay, Jean-Charles; Aiello, Giacomo; Aubert, Julien; Villari, Rosaria; Fischer, Ulrich

    2016-01-01

    Highlights: • A complete nuclear analysis of the DEMO HCLL has been carried out at CEA with the TRIPOLI-4"® Monte Carlo code. • The DEMO tokamak model was generated by the CAD import tool McCad. • The HCLL blankets were implemented using a previous MCNP model developed at ENEA. • A good agreement is observed between the results obtained at CEA with TRIPOLI-4 and JEFF-3.1.1 and whose obtained at ENEA with MCNP and FENDL-2.1. - Abstract: This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4"® Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4"® representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4"® Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO.

  19. Evaluation of EM loads distribution on DEMO blanket segments and their effect on mechanical integrity

    International Nuclear Information System (INIS)

    Maione, Ivan Alessio; Zeile, Christian; Boccaccini, Lorenzo V.; Vaccaro, Alessandro

    2016-01-01

    Highlights: • Two DEMO 2015 ANSYS FEM models (for EM and structural analysis) have been implemented based on the EU-HCPB concept. • Lorentz’s forces have been calculated and their impact on the segment structure has been evaluated. • EM loads show a predominant total radial moment due to the high toroidal magnetic field (in comparison with the poloidal one). • A preliminary assessment of the primary stresses according the RCC-MRx code indicates the ability of the segments to resist the EM forces. - Abstract: This work is aimed to analyze the EM internal forces distribution on the blanket system (blankets modules and segment back supporting structure) of the EU PPPT DEMO 2015 reactor configuration. In order to validate their impact on the segment structure, an EM analysis is conducted using a simplified plasma central disruption. The calculated Lorentz’s forces distributions are then used as input for structural analyses focusing on the mechanical integrity of the segment back supporting structure. In particular, the electrical and structural assumptions used in this work are based on the HCPB blanket design developed at the Karlsruhe Institute of Technology. A preliminary assessment of the primary stresses according the design code RCC-MRx indicates the ability of the segments to resist the EM forces, where the lowest margin is given by the immediate plastic instability criterion on the inboard segment with 14%.

  20. 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.

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

    International Nuclear Information System (INIS)

    Panayotov, D.; Sardain, P.; Boccaccini, L.V.; Salavy, J.-F.; Cismondi, F.; Jourd'Heuil, L.

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Tesini, Alessandro

    2014-01-01

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

  3. Comparison of different fusion nuclear data libraries using the European INTOR blanket design

    International Nuclear Information System (INIS)

    Pelloni, S.; Stepanek, J.; Dudziak, D.

    1982-12-01

    The European Community International Tokamak Reactor (INTOR-EC) was used to investigate the influence of different cross-section libraries on the tritium breeding ratio. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, and the recently developed Swiss surface-flux code SURCU, for the Li 17 Pb 83 and Li 2 SiO 3 blanket designs. Nuclear data considered were from the DLC-37, VITAMIN-C (DLC-41) and Los Alamos-NJOY fusion libraries. In addition the reaction rates were estimated using the MACKLIB-IV response library. It is shown that very good agreement (within 0.5%) between the breeding ratios obtained using the VITAMIN-C and Los Alamos libraries could be obtained, whereas the corresponding values calculated using VITAMIN-C and MACKLIB-IV data sets collapsed into 25 neutron and 21 gamma groups differ up to 23%. It is found that this large discrepancy is due to the 6 Li(n, α) reaction cross sections in the low energy range between 4 and 0.03 eV. Furthermore, the collapsed DLC-37 library is not adequate for fusion blankets with a soft spectrum. It is important that greater care be given to preparation of broad group cross section sets, especially in the thermal energy region for blankets containing highly moderating materials. (Auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  5. Present status of irradiation tests on tritium breeding blanket for fusion reactor

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

    1994-01-01

    To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

  6. Ionosonde-based indices for improved representation of solar cycle variation in the International Reference Ionosphere model

    Science.gov (United States)

    Brown, Steven; Bilitza, Dieter; Yiǧit, Erdal

    2018-06-01

    A new monthly ionospheric index, IGNS, is presented to improve the representation of the solar cycle variation of the ionospheric F2 peak plasma frequency, foF2. IGNS is calculated using a methodology similar to the construction of the "global effective sunspot number", IG, given by Liu et al. (1983) but selects ionosonde observations based on hemispheres. We incorporated the updated index into the International Reference Ionosphere (IRI) model and compared the foF2 model predictions with global ionospheric observations. We also investigated the influence of the underlying foF2 model on the IG index. IRI has two options for foF2 specification, the CCIR-66 and URSI-88 foF2 models. For the first time, we have calculated IG using URSI-88 and assessed the impact on model predictions. Through a retrospective model-data comparison, results show that the inclusion of the new monthly IGNS index in place of the current 12-month smoothed IG index reduce the foF2 model prediction errors by nearly a factor of two. These results apply to both day-time and nightime predictions. This is due to an overall improved prediction of foF2 seasonal and solar cycle variations in the different hemispheres.

  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. Status of the EU test blanket systems safety studies

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-01-01

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  9. Status of the EU test blanket systems safety studies

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-10-15

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  10. Optimization of up-flow anaerobic sludge blanket reactor for ...

    African Journals Online (AJOL)

    Optimization of up-flow anaerobic sludge blanket reactor for treatment of composite ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... Granules grown in the bottom part of UASB reactor were more compact and tense ...

  11. Electromagnetic effects involving a tokamak reactor first wall and blanket

    International Nuclear Information System (INIS)

    Turner, L.R.; Evans, K. Jr.; Gelbard, E.; Prater, R.

    1980-01-01

    Four electromagnetic effects experienced by the first wall and blanket of a tokamak reactor are considered. First, the first wall provides reduction of the growth rate of vertical axisymmetric instability and stabilization of low mode number interval kink modes. Second, if a rapid plasma disruption occurs, a current will be induced on the first wall, tending to maintain the field formerly produced by the plasma. Third, correction of plasma movement can begin on a time scale much faster than the L/R time of the first wall and blanket. Fourth, field changes, especially those from plasma disruption or from rapid discharge of a toroidal field coil, can cause substantial eddy current forces on elements of the first wall and blanket. These effects are considered specifically for the first wall and blanket of the STARFIRE commercial reactor design study

  12. Blanket options for high-efficiency fusion power

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  13. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  14. Fusion blanket for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Taussig, R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperature (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by Ar) utilizing Li 2 O for tritium breeding. In this design, approx. 60% of the fusion energy is deposited in the high-temperature interior. The maximum Ar temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  15. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1981-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 deg C) of conventional structural materials such as stainless steels. In this project 'two-zone' blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 deg C leading to an overall efficiency estimate of 55 to 60% for this reference case. (author)

  16. Fusion-reactor blanket-material safety-compatibility studies

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

    1982-11-01

    Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO 2 , Li 2 ZrO 3 , Li 2 SiO 3 , Li 4 SiO 4 and LiTiO 3 ) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li 7 Pb 2 alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li 17 Pb 83 alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li 17 Pb 83 alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns

  17. Nuclear characteristics of D-D fusion reactor blankets, (1)

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao; Seki, Yasushi.

    1977-01-01

    Fusion reactors operating on the deuterium (D-D) cycle are considered promising for their freedom from tritium breeding in the blanket. In this paper, neutronic and photonic calculations are undertaken covering several blanket models of the D-D fusion reactor, using presently available data, with a view to comparing the nuclear characteristics of these models, in particular, the nuclear heating rates and their spatial distributions. Nine models are taken up for the study, embodying various combinations of coolant, blanket, structural and reflector materials. About 10 MeV is found to be a typical value for the total nuclear energy deposition per source neutron in the models considered here. The realization of high energy gain is contingent upon finding a favorable combination of blanket composition and configuration. The resulting implications on the thermal design aspect are briefly discussed. (auth.)

  18. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1987-01-01

    A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

  19. Blast venting through blanket material in the HYLIFE ICF reactor

    International Nuclear Information System (INIS)

    Liu, J.C.; Peterson, P.F.; Schrock, V.E.

    1992-01-01

    This work presents a numerical study of blast venting through various blanket configurations in the HYLIFE ICF reactor design. The study uses TSUNAMI -- a multi-dimensional, high-resolution, shock capturing code -- to predict the momentum exchange and gas dynamics for blast venting in complex geometries. In addition, the study presents conservative predictions of wall loading by gas shock and impulse delivered to the protective liquid blanket. Configurations used in the study include both 2700 MJ and 350 MJ fusion yields per pulse for 5 meter and 3 meter radius reactor chambers. For the former, an annular jet array is used for the blanket geometry, while in the latter, both annular jet array as well as slab geometries are used. Results of the study indicate that blast venting and wall loading may be manageable in the HYLIFE-II design by a judicious choice of blanket configuration

  20. Imploding-liner reactor nucleonic studies: the LINUS blanket

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1977-09-01

    Scoping nucleonic studies have been performed for a small imploding-liner fusion reactor concept. Tritium breeding ratio and time-dependent energy deposition rates were the primary parameters of interest in the study. Alloys of Pb and LiPb were considered for the liquid liner (blanket), and tritium breeding was found to be more than adequate with blankets less than 1 m thick. However, neutron leakages into the solid cylinder block surrounding the liquid liner are generally quite high, so considerable effort was concentrated on minimizing these values. Time-dependent calculations reveal that 89% of the energy is deposited in the blanket within 2 μs. Thus, LINUS's blanket should remain intact for the requisite neutron and gamma-ray lifetimes

  1. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

    Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

    1982-01-01

    Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

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

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-01-01

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

  3. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

  4. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-02-01

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

  5. Feasibility study of fusion breeding blanket concept employing graphite reflector

    International Nuclear Information System (INIS)

    Cho, Seungyon; Ahn, Mu-Young; Lee, Cheol Woo; Kim, Eung Seon; Park, Yi-Hyun; Lee, Youngmin; Lee, Dong Won

    2015-01-01

    Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

  6. Feasibility study of fusion breeding blanket concept employing graphite reflector

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seungyon, E-mail: sycho@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Woo; Kim, Eung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

  7. Overview of first wall/blanket/shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-04-01

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

  8. Applications of the Aqueous Self-Cooled Blanket concept

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.J.; Varsamis, G.; Wrisley, K.; Deutch, L.; Gierszewski, P.

    1986-01-01

    In this paper a novel water-cooled blanket concept is examined. This concept, designated the Aqueous Self-Cooled Blanket (ASCB), employs water with small amounts of dissolved fertile compounds as both the coolant and the breeding medium. The ASCB concept is reviewed and its application in three different contexts is examined: (1) power reactors; (2) near-term devices such as NET; and (3) fusion-fission hybrids

  9. Progress on DEMO blanket attachment concept with keys and pins

    International Nuclear Information System (INIS)

    Vizvary, Zsolt; Iglesias, Daniel; Cooper, David; Crowe, Robert; Riccardo, Valeria

    2015-01-01

    Highlights: • DEMO blanket attachment system with keys and pins (without using bolts). • Blanket segments are preloaded by progressively designed springs. • Blanket back plate flexibility has a major impact on spring design. • Mechanical analysis of other components indicates no unresolvable issues. • Thermal analysis indicates acceptable temperatures for the support system. - Abstract: The blanket attachment has to cope with gravity, thermal and electromagnetic loads, also it has to be installed and serviced by remote handling. Pre-stressed components suffer from stress relaxation in irradiated environments such as DEMO. To circumvent this problem pre-stressed component should be either avoided or shielded, and where possible keys and pins should be used. This strategy has been proposed for the DEMO multi-module segments (MMS). The blanket segments are held by two tapered keys each, designed to allow thermal expansions while providing contact with the vacuum vessel and to resist the poloidal and radial moments the latter being dominant at 9.1 MNm inboard and 15 MNm outboard. On the top of the blanket segment there is a pin which provides vertical support. At the bottom another vertical support has to lock them in position after installation and manage the pre-load on the segments. The pre-load is required to deal with the electromagnetic loads during disruption. This is provided by a set of springs, which require shielding as they are preloaded. These are sized to cope with the force (3 MN inboard, 1.4 MN outboard) due to halo currents and the toroidal moment which can reverse. Calculations show that the flexibility of the blanket segment itself plays a significant role in defining the required support system. The blanket segment acts as a preloaded spring and it has to be part of the attachment design as well.

  10. Electrical connectors for blanket modules in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I., E-mail: poddubnyyii@nikiet.ru [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Khomiakov, S.; Kolganov, V. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Sadakov, S.; Calcagno, B.; Chappuis, Ph.; Roccella, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Danilov, I.; Leshukov, A.; Strebkov, Y. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Ulrickson, M. [Sandia National Laboratories MS-1129, PO Box 5800, Albuquerque, NM 87185 (United States)

    2014-10-15

    Highlights: • Analysis of static and cyclic strength for L-shaped and Z-shaped ES has been performed. • Analysis results do show that for L-shaped ES static and cyclic strength criteria are not satisfied. • Static and cyclic strength criteria are met well by ES with Z-shaped elastic elements. • ES with Z-shaped elastic elements has been adopted as a new baseline design for ITER. - Abstract: Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak. The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was

  11. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

    Enoeda, M.; Sato, S.; Hatano, T.

    2001-01-01

    Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

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

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

    For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

    the energy conversion, on the other hand, it is the demerit to the structural limitation of the structural material of the FW which must remove the high surface heat flux from the plasma. To resolve this issue, the coolant path was selected to cool FWs of 4 modules first, and later the coolant was planned to cool the breeder region with higher temperature. From this flow path, the estimated highest temperature of the FW cooling is 360degC. By using this value, the thermo-mechanical performance was estimated to show the feasibility to the thermal stress and the internal coolant pressure. Also, TBR and thermal analysis was performed to search the acceptable dimensioning of the breeder layer and multiplier layer. As the result of the conceptual design, the basic feasibility was shown for such aspects as, heat removal, power generation, fuel production, neutron shielding and so on to the proposed DEMO blanket concept. Also, the other important issues such as, electro-magnetic performance and loads, corrosion of the supercritical water, tritium recovery system, power generation system, fabrication feasibility of the proposed blanket structure, remote handling system and so on, were preliminarily researched and identified as the issues to be clarified by R and D. (author)

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

    International Nuclear Information System (INIS)

    Enoeda, Mikio; Akiba, Masato; Ohara, Yoshihiro

    2001-12-01

    conversion, on the other hand, it is the demerit to the structural limitation of the structural material of the FW which must remove the high surface heat flux from the plasma. To resolve this issue, the coolant path was selected to cool FWs of 4 modules first, and later the coolant was planned to cool the breeder region with higher temperature. From this flow path, the estimated highest temperature of the FW cooling is 360degC. By using this value, the thermo-mechanical performance was estimated to show the feasibility to the thermal stress and the internal coolant pressure. Also, TBR and thermal analysis was performed to search the acceptable dimensioning of the breeder layer and multiplier layer. As the result of the conceptual design, the basic feasibility was shown for such aspects as, heat removal, power generation, fuel production, neutron shielding and so on to the proposed DEMO blanket concept. Also, the other important issues such as, electro-magnetic performance and loads, corrosion of the supercritical water, tritium recovery system, power generation system, fabrication feasibility of the proposed blanket structure, remote handling system and so on, were preliminarily researched and identified as the issues to be clarified by R and D. (author)

  15. Transmutation blanket design for a Tokamak system

    International Nuclear Information System (INIS)

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

    2011-01-01

    Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

  16. On the use of tin-lithium alloys as breeder material for blankets of fusion power plants

    International Nuclear Information System (INIS)

    Fuetterer, M.A.; Aiello, G.; Barbier, F.; Giancarli, L.; Poitevin, Y.; Sardain, P.; Szczepanski, J.; Li Puma, A.; Ruvutuso, G.; Vella, G.

    2000-01-01

    Tin-lithium alloys have several attractive thermo-physical properties, in particular high thermal conductivity and heat capacity, that make them potentially interesting candidates for use in liquid metal blankets. This paper presents an evaluation of the advantages and drawbacks caused by the substitution of the currently employed alloy lead-lithium (Pb-17Li) by a suitable tin-lithium alloy: (i) for the European water-cooled Pb-17Li (WCLL) blanket concept with reduced activation ferritic-martensitic steel as the structural material; (ii) for the European self-cooled TAURO blanket with SiC f /SiC as the structural material. It was found that in none of these blankets Sn-Li alloys would lead to significant advantages, in particular due to the low tritium breeding capability. Only in forced convection cooled divertors with W-alloy structure, Sn-Li alloys would be slightly more favorable. It is concluded that Sn-Li alloys are only advantageous in free surface cooled reactor internals, as this would make maximum use of the principal advantage of Sn-Li, i.e., the low vapor pressure

  17. Welding and cutting characteristics of blanket/first wall module to back plate for fusion experimental reactor

    International Nuclear Information System (INIS)

    Sato, Shinichi; Osaki, Toshio; Koga, Shinji

    1996-01-01

    The first wall and the blanket of the International Thermonuclear Experimental Reactor (ITER) are used under severe conditions such as the neutron irradiation by plasma, surface thermal load, the electromagnetic force at the time of plasma disruption and others. Consequently, from the viewpoint of the necessity for disassembling and maintenance, those are divided into modules in toroidal and poloidal directions. In this study, as to the welding of the back plate and the legs supporting blanket modules, which are installed in a vacuum vessel, the characteristic test paying attention to the deformation at the time of welding was carried out, and the optimal welding conditions and the characteristics of welding deformation and others were clarified. Moreover, when water jet method was used for cutting the welded parts of the supporting legs, the properties of the cut parts, the time for cutting and others were examined. The performance required for the welded parts of blanket modules with back plate is shown. The basic test of welding conditions using plate models, partial model test and whole model test are reported. The test of water jet cutting for the maintenance of shielding blanket modules is described. (K.I.)

  18. National economic and development indicators and international variation in prostate cancer incidence and mortality: an ecological analysis.

    Science.gov (United States)

    Neupane, Subas; Bray, Freddie; Auvinen, Anssi

    2017-06-01

    Macroeconomic indicators are likely associated with prostate cancer (PCa) incidence and mortality globally, but have rarely been assessed. Data on PCa incidence in 2003-2007 for 49 countries with either nationwide cancer registry or at least two regional registries were obtained from Cancer Incidence in Five Continents Vol X and national PCa mortality for 2012 from GLOBOCAN 2012. We compared PCa incidence and mortality rates with various population-level indicators of health, economy and development in 2000. Poisson and linear regression methods were used to quantify the associations. PCa incidence varied more than 15-fold, being highest in high-income countries. PCa mortality exhibited less variation, with higher rates in many low- and middle-income countries. Healthcare expenditure (rate ratio, RR 1.46, 95 % CI 1.45-1.47) and population growth (RR 1.15, 95 % CI 1.14-1.16), as well as computer and mobile phone density, were associated with a higher PCa incidence, while gross domestic product, GDP (RR 0.94, 95 % CI 0.93-0.95) and overall mortality (RR 0.72, 95 % CI 0.71-0.73) were associated with a low incidence. GDP (RR 0.55, 95 % CI 0.46-0.66) was also associated with a low PCa mortality, while life expectancy (RR 3.93, 95 % CI 3.22-4.79) and healthcare expenditure (RR 1.20, 95 % CI 1.09-1.32) were associated with an elevated mortality. Our results show that healthcare expenditure and, thus, the availability of medical resources are an important contributor to the patterns of international variation in PCa incidence. This suggests that there is an iatrogenic component in the current global epidemic of PCa. On the other hand, higher healthcare expenditure is associated with lower PCa death rates.

  19. Low variation in ribosomal DNA and internal transcribed spacers of the symbiotic fungi of leaf-cutting ants (Attini: Formicidae

    Directory of Open Access Journals (Sweden)

    Silva-Pinhati A.C.O.

    2004-01-01

    Full Text Available Leaf-cutting ants of the genera Atta and Acromyrmex (tribe Attini are symbiotic with basidiomycete fungi of the genus Leucoagaricus (tribe Leucocoprineae, which they cultivate on vegetable matter inside their nests. We determined the variation of the 28S, 18S, and 5.8S ribosomal DNA (rDNA gene loci and the rapidly evolving internal transcribed spacers 1 and 2 (ITS1 and ITS2 of 15 sympatric and allopatric fungi associated with colonies of 11 species of leafcutter ants living up to 2,600 km apart in Brazil. We found that the fungal rDNA and ITS sequences from different species of ants were identical (or nearly identical to each other, whereas 10 GenBank Leucoagaricus species showed higher ITS variation. Our findings suggest that Atta and Acromyrmex leafcutters living in geographic sites that are very distant from each other cultivate a single fungal species made up of closely related lineages of Leucoagaricus gongylophorus. We discuss the strikingly high similarity in the ITS1 and ITS2 regions of the Atta and Acromyrmex symbiotic L. gongylophorus studied by us, in contrast to the lower similarity displayed by their non-symbiotic counterparts. We suggest that the similarity of our L. gongylophorus isolates is an indication of the recent association of the fungus with these ants, and propose that both the intense lateral transmission of fungal material within leafcutter nests and the selection of more adapted fungal strains are involved in the homogenization of the symbiotic fungal stock.

  20. International variation in GP treatment strategies for subclinical hypothyroidism in older adults: a case-based survey.

    Science.gov (United States)

    den Elzen, Wendy P J; Lefèbre-van de Fliert, Anne A; Virgini, Vanessa; Mooijaart, Simon P; Frey, Peter; Kearney, Patricia M; Kerse, Ngaire; Mallen, Christian D; McCarthy, Vera J C; Muth, Christiane; Rosemann, Thomas; Russell, Audrey; Schers, Henk; Stott, David J; de Waal, Margot W M; Warner, Alex; Westendorp, Rudi G J; Rodondi, Nicolas; Gussekloo, Jacobijn

    2015-02-01

    There is limited evidence about the impact of treatment for subclinical hypothyroidism, especially among older people. To investigate the variation in GP treatment strategies for older patients with subclinical hypothyroidism depending on country and patient characteristics. Case-based survey of GPs in the Netherlands, Germany, England, Ireland, Switzerland, and New Zealand. The treatment strategy of GPs (treatment yes/no, starting-dose thyroxine) was assessed for eight cases presenting a woman with subclinical hypothyroidism. The cases differed in the patient characteristics of age (70 versus 85 years), vitality status (vital versus vulnerable), and thyroid-stimulating hormone (TSH) concentration (6 versus 15 mU/L). A total of 526 GPs participated (the Netherlands n = 129, Germany n = 61, England n = 22, Ireland n = 21, Switzerland n = 262, New Zealand n = 31; overall response 19%). Across countries, differences in treatment strategy were observed. GPs from the Netherlands (mean treatment percentage 34%), England (40%), and New Zealand (39%) were less inclined to start treatment than GPs in Germany (73%), Ireland (62%), and Switzerland (52%) (P = 0.05). Overall, GPs were less inclined to start treatment in 85-year-old than in 70-year-old females (pooled odds ratio [OR] 0.74 [95% confidence interval [CI] = 0.63 to 0.87]). Females with a TSH of 15 mU/L were more likely to get treated than those with a TSH of 6 mU/L (pooled OR 9.49 [95% CI = 5.81 to 15.5]). GP treatment strategies of older people with subclinical hypothyroidism vary largely by country and patient characteristics. This variation underlines the need for a new generation of international guidelines based on the outcomes of randomised clinical trials set within primary care. © British Journal of General Practice 2015.

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

    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 this 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. The primary results of the study are as follows: a) the lithium-lead blanket achieves a higher TBR with a smaller blanket thickness relative to the lithium blanket; b) the lithium blanket generates more energy per fusion neutron relative to the lithium-lead blanket; c) among the possible reflector materials, the carbon reflector produces the highest TBR; d) the high-Z reflector materials (Mo, Cu, W, or steel) generate more energy per fusion neutron and produce smaller TBRs relative to the carbon reflector; e) lithium-6 enrichment is required for the lithium-lead blanket to reduce the total blanket thickness; and f) the energy deposition per fusion neutron reaches a saturation as the blanket thickness, the fraction of the high-Z material in the reflector, or the reflector zone thickness increases (this allows one to design the blanket for a specific TBR without reducing the energy production)

  2. Manufacturing aspects in the design of the breeder unit for Helium Cooled Pebble Bed blankets

    International Nuclear Information System (INIS)

    Rey, J.; Ihli, T.; Filsinger, D.; Polixa, C.

    2007-01-01

    The breeding blanket programme has been in the focus of European fusion research for more than a decade. Recently, it has been driven by the EU Power Plant Conceptual Study (PPCS), investigating the potential of fusion energy in a future economic environment. On the way to the first commercial nuclear fusion reactor (DEMO) new studies for reactor in-vessel components have been initiated. One central focus is the design and manufacturing of the blankets that have to ensure the breeding process to maintain the fuel cycle and are also responsible for the extraction of the main part of the reactor heat for power generation. Two kinds are established: One is the Helium Cooled Pebble Bed (HCPB) and the other the Helium Cooled Liquid Lead (HCLL) blanket. Both designs employ three different cooling plate assemblies. The outer, cooled U-shaped shell, namely the First Wall (FW), with two caps builds the blanket box. The structural strength of the blanket box is realized by integrating Stiffening Grids (SG) that separate the equally spaced Breeder Unit (BU) and allow the box, in case of faulted conditions, to withstand an internal pressure of 8 MPa. The cooled SG constitute the side walls of the BU and are also cooled. The BU consists of a dedicated Cooling Plate (CP) assembly. In present studies about the fabrication of Cooling Plates two kinds of diffusion welding processes are focused on. One is based on a Hot Isostatic Gas Process (HIP). The second is a uni-axial Diffusion Welding Process (DWP). In both cases the bond between the two halves of the cooling plate structure is reached by controlled pressure and heat cycles. Approaching larger, realistic scaled components the uncertainty of ensuring uniform process parameters across the bonding zone increases the risk of defect sources and, therefore, makes it difficult to guarantee the required bonding penetration. This study presents an alternative manufacturing strategy. The premises for this strategy are the reduction of

  3. International variations in the gestational age distribution of births: an ecological study in 34 high-income countries.

    Science.gov (United States)

    Delnord, Marie; Mortensen, Laust; Hindori-Mohangoo, Ashna D; Blondel, Béatrice; Gissler, Mika; Kramer, Michael R; Richards, Jennifer L; Deb-Rinker, Paromita; Rouleau, Jocelyn; Morisaki, Naho; Nassar, Natasha; Bolumar, Francisco; Berrut, Sylvie; Nybo Andersen, Anne-Marie; Kramer, Michael S; Zeitlin, Jennifer

    2018-04-01

    Few studies have investigated international variations in the gestational age (GA) distribution of births. While preterm births (22-36 weeks GA) and early term births (37-38 weeks) are at greater risk of adverse health outcomes compared to full term births (39-40 weeks), it is not known if countries with high preterm birth rates also have high early term birth rates. We examined rate associations between preterm and early term births and mean term GA by mode of delivery onset. We used routine aggregate data on the GA distribution of singleton live births from up to 34 high-income countries/regions in 1996, 2000, 2004, 2008 and 2010 to study preterm and early term births overall and by spontaneous or indicated onset. Pearson correlation coefficients were adjusted for clustering in time trend analyses. Preterm and early term births ranged from 4.1% to 8.2% (median 5.5%) and 15.6% to 30.8% (median 22.2%) of live births in 2010, respectively. Countries with higher preterm birth rates in 2004-2010 had higher early term birth rates (r > 0.50, P birth rates suggest that common risk factors could underpin shifts in the GA distribution. Targeting modifiable population risk factors for delivery before 39 weeks GA may provide a useful preterm birth prevention paradigm.

  4. Progress on the Fabrication Methods Development for the Korean Test Blanket Module First Wall in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Cho, Seung Yon

    2010-01-01

    A Korean helium cooled molten lithium (HCML) test blanket module (TBM) has been designed to be tested in the International Thermonuclear Experimental Reactor (ITER) TBM and related fabrication methods have been developed especially for the purpose of joining. Since the first wall (FW) of the HCML TBM is composed of a beryllium (Be) as an armor material and a FMS as a structural one, joining with Be to FMS and FMS to FMS should be developed in order to fabricate it

  5. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  6. The transpiration cooled first wall and blanket concept

    International Nuclear Information System (INIS)

    Barleon, Leopold; Wong, Clement

    2002-01-01

    To achieve high thermal performance at high power density the EVOLVE concept was investigated under the APEX program. The EVOLVE W-alloy first wall and blanket concept proposes to use transpiration cooling of the first wall and boiling or vaporizing lithium (Li) in the blanket zone. Critical issues of this concept are: the Magnetohydrodynamic (MHD) pressure losses of the Li circuit, the evaporation through a capillary structure and the needed superheating of the Li at the first wall and blanket zones. Application of the transpiration concept to the blanket region results in the integrated transpiration cooling concept (ITCC) with either toroidal or poloidal first wall channels. For both orientations the routing of the liquid Li and the Li vapor has been modeled and the corresponding pressure losses have been calculated by varying the width of the supplying slot and the capillary diameter. The concept works when the sum of the active and passive pumping head is higher than the total system pressure losses and when the temperature at the inner side of the first wall does not override the superheating limit of the coolant. This cooling concept has been extended to the divertor design, and the removal of a surface heat flux of up to 10 MW/m 2 appears to be possible, but this paper will focus on the transpiration cooled first wall and blanket concept assessment

  7. Assessment of alkali metal coolants for the ITER blanket

    International Nuclear Information System (INIS)

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

    1994-01-01

    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 will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

  8. Tritium transport in the water cooled Pb-17Li blanket concept of DEMO

    International Nuclear Information System (INIS)

    Reiter, F.; Tominetti, S.; Perujo, A.

    1992-01-01

    The code TIRP has been used to calculate the time dependence of tritium inventory and tritium permeation into the coolant and into the first wall boxes in the water cooled Pb-17Li blanket concept of DEMO. The calculations have been performed for the martensitic steel MANET and the austenitic steel AISI 316L as blanket structure materials, for water or helium cooling and for convective or no motion of the liquid breeder in the blanket. Tritium inventories are rather low in blankets with MANET structure and higher in those with AISI 316L structure. Tritium permeation rates are too high in both blankets. Further calculations on tritium inventory and permeation are therefore presented for blankets with TiC permeation barriers of 1 μm thickness on various surfaces of the blanket structure and for blankets with any permeation barriers in function of their thickness, tritium diffusivities, tritium surface recombination rates and atomic densities. These last calculations have been performed for a blanket with coatings on the outer surfaces of the blanket and with a tritium residence time of 10 4 s and for a blanket with coatings on both sides of the cooling tubes and stagnant Pb-17Li in the blanket. The second case for a blanket with MANET structure presents a very interesting solution for tritium recovery by permeation into and pumping from the first wall boxes. (orig.)

  9. Status of blanket design for RTO/RC ITER

    International Nuclear Information System (INIS)

    Yamada, M.; Ioki, K.; Cardella, A.; Elio, F.; Miki, N.

    2000-01-01

    Design has progressed on the FW/blanket for the RTO/RC (reduced technical objective/ reduced cost) ITER. The basic functions and structures are the same as for the 1998 ITER design. However, design and fabrication methods of the FW/blanket have been improved to achieve ∝ 50% reduction of the construction cost compared to that for the 1998 ITER design. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the EDA (engineering design activity) is still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed. (orig.)

  10. Computation Method Comparison for Th Based Seed-Blanket Cores

    International Nuclear Information System (INIS)

    Kolesnikov, S.; Galperin, A.; Shwageraus, E.

    2004-01-01

    This work compares two methods for calculating a given nuclear fuel cycle in the WASB configuration. Both methods use the ELCOS Code System (2-D transport code BOXER and 3-D nodal code SILWER) [4] are compared. In the first method, the cross-sections of the Seed and Blanket, needed for the 3-D nodal code are generated separately for each region by the 2-D transport code. In the second method, the cross-sections of the Seed and Blanket, needed for the 3-D nodal code are generated from Seed-Blanket Colorsets (Fig.1) calculated by the 2-D transport code. The evaluation of the error introduced by the first method is the main objective of the present study

  11. Neutronics design aspects of reference ARIES-I fusion blanket

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1990-12-01

    A SiC composite blanket concept was recently conceived for a deuterium-tritium burning, 1000 MW(e) tokamak fusion reactor design, ARIES-I. SiC composite structural material was chosen due to its very low activation features. High blanket nuclear performance and thermal efficiency, adequate tritium breeding, and a low level of activation are important design requirements for the ARIES-I reactor. The major approaches, other than using SiC as structural material, in meeting these design requirements, are to employ beryllium, the only low activation neutron multiplying material, and isotopically tailored Li 2 ZrO 3 , a tritium breeding material stable at high temperature, as blanket materials. 5 refs., 4 figs., 2 tabs

  12. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

  13. High temperature blankets for the production of synthetic fuels

    International Nuclear Information System (INIS)

    Powell, J.R.; Steinberg, M.; Fillo, J.; Makowitz, H.

    1977-01-01

    The application of very high temperature blankets to improved efficiency of electric power generation and production of H 2 and H 2 based synthetic fuels is described. The blanket modules have a low temperature (300 to 400 0 C) structure (SS, V, Al, etc.) which serves as the vacuum/coolant pressure boundary, and a hot (>1000 0 C) thermally insulated interior. Approximately 50 to 70% of the fusion energy is deposited in the hot interior because of deep penetration by high energy neutrons. Separate coolant circuits are used for the two temperature zones: water for the low temperature structure, and steam or He for the hot interior. Electric generation efficiencies of approximately 60% and H 2 production efficiencies of approximately 50 to 70%, depending on design, are projected for fusion reactors using these high temperature blankets

  14. Heating an aquaculture pond with a solar pool blanket

    Energy Technology Data Exchange (ETDEWEB)

    Wisely, B; Holliday, J E; MacDonald, R E

    1982-01-01

    A floating solar blanket of laminated bubble plastic was used to heat a 0.11 ha seawater pond of 1.3 m depth. The covered pond maintained daily temperatures 6 to 9/sup 0/C above two controls. Local air temperatures averaged 14 to 19/sup 0/C. Oysters, prawns, seasquirts, and fish in the covered pond all survived. After three weeks, the blanket separated. This was the result of pond temperatures exceeding 30/sup 0/C, the maximum manufacturer's specification. Floating blankets fabricated to higher specifications would be useful for maintaining above-ambient temperatures in small ponds or tanks in temporary situations during cold winter months and might have a more permanent use.

  15. Solid breeder test blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ying, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)]. E-mail: ying@fusion.ucla.edu; Abdou, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Calderoni, P. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Sharafat, S. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Youssef, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); An, Z. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Abou-Sena, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Kim, E. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Reyes, S. [LANL, Livermore, CA (United States); Willms, S. [LANL, Los Alamos, NM (United States); Kurtz, R. [PNNL, Richland, WA (United States)

    2006-02-15

    This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration of fusion break-in phenomena and configuration scoping. Specific emphasis is placed on first wall structural response, evaluation of neutronic parameters, assessment of thermomechanical behavior and characterization of tritium release. The tests will be conducted with three unit cell arrays/sub-modules. The development approach includes: (1) design the unit cell/sub-module for low temperature operations and (2) refer to a reactor blanket design and use engineering scaling to reproduce key parameters under ITER wall loading conditions, so that phenomena under investigation can be measured at a reactor-like level.

  16. Direct LiT Electrolysis in a Metallic Fusion Blanket

    International Nuclear Information System (INIS)

    Olson, Luke

    2016-01-01

    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.

  17. The evolution of US helium-cooled blankets

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.

    1991-01-01

    This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors. (orig.)

  18. Status of fusion reactor blanket evaluation studies in France

    International Nuclear Information System (INIS)

    Carre, F.; Chevereau, G.; Gervaise, F.; Proust, E.

    1985-03-01

    In the frame of recent CEA studies aiming at the evaluation and at the comparison of various candidate blanket concepts in moderate power conditions (Psub(n) approximately 2 MW/m 2 ), the present work examines the neutronic and thermomechanical performances of a water cooled Li 17 Pb 83 tubular blanket and those of a helium cooled canister blanket taking advantage of the excellent breeding capability of composite Beryllium/LiAlO 2 (85/15%) breeder elements. The purpose of the following discussion is to justify the impetus for these reference concepts and to summarize the state of their evaluation studies updated by the continuous assimilation of calculations and experiments in progress

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

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Porfiri, T.

    1996-06-01

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

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

    International Nuclear Information System (INIS)

    Reed, C.B.; Haglund, R.C.; Miller, M.E.; Nasiatka, J.R.; Kirillov, I.R.; Ogorodnikov, A.P.; Preslitski, G.V.; Goloubovitch, G.P.; Xu, Zeng Yu

    1996-01-01

    The Vanadium/Lithium system has been the recent focus of ANL's Blanket Technology Pro-ram, and for the last several years, ANL's Liquid Metal Blanket activities have been carried out in direct support of the ITER (International Thermonuclear Experimental Reactor) breeding blanket task area. A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the Near the development of insulator coatings. Design calculations, Hua and Gohar, show that an electrically insulating layer is necessary to maintain an acceptably low magneto-hydrodynamic (MHD) pressure drop in the current ITER design. Consequently, the decision was made to convert Argonne's Liquid Metal EXperiment (ALEX) from a 200 degrees C NaK facility to a 350 degrees C lithium facility. The upgraded facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups at Hartmann numbers (M) and interaction parameters (N) in the range of 10 3 to 10 5 in lithium at 350 degrees C. Following completion of the upgrade work, a short performance test was conducted, followed by two longer multiple-hour, MHD tests, all at 230 degrees C. The modified ALEX facility performed up to expectations in the testing. MHD pressure drop and test section voltage distributions were collected at Hartmann numbers of 1000

  1. Overview of the TFTR Lithium Blanket Module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

    The LBM (Lithium Blanket Module) is an approximately cubic module, about 80 cm on each side, with construction representative of a helium-cooled lithium oxide fusion reactor blanket module. Measurements of neutron transport and tritium breeding in the LBM will be made in irradiation programs first with a point-neutron source, and subsequently with the D-D and D-T fusion-neutron sources of the TFTR. This paper summarizes the objectives of the LBM program, the design, development and construction of the LBM, and progress in the experimental tests

  2. ITER blanket module shield block design and analysis

    International Nuclear Information System (INIS)

    Mitin, D.; Khomyakov, S.; Razmerov, A.; Strebkov, Yu.

    2008-01-01

    This paper presents the alternative design of the shield block cooling path for a typical ITER blanket module with a predominantly sequential flow circuit. A number of serious disadvantages have been observed for the reference design, where the parallel flow circuit is used, which is inherent in the majority of blanket modules. The paper discusses these disadvantages and demonstrates the benefit of the alternative design based on the detailed design and the technological, hydraulic, thermal, structural and strength analyses, conducted for module no. 17

  3. Progress and achievements of the ITER L-4 blanket project

    International Nuclear Information System (INIS)

    Daenner, W.; Toschi, R.; Cardella, A.

    1999-01-01

    The L-4 Blanket Project embraces the R and D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R and D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined.The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification tests. (author)

  4. Ceramic BOT type blanket with poloidal helium cooling

    International Nuclear Information System (INIS)

    Cardella, A.; Daenenr, W.; Iseli, M.; Ferrari, M.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.

    1989-01-01

    This paper briefly describes the work done and results achieved over the past two years on the ceramic breeder BOT blanket with poloidal helium cooling. A conclusive remark on the brick/plate option described previously is followed by short descriptions of the low and high performance pebble bed options elaborated as alternatives for both NET and DEMO. The results show, togethre with those about the poloidal cooling of the First Wall, good prospects for this blanket type provided that the questions connected wiht an extensive use of beryllium find a satisfactor answer. (author). 5 refs.; 7 figs.; 1 tab

  5. Progress and achievements of the ITER L-4 blanket project

    International Nuclear Information System (INIS)

    Daenner, W.; Toschi, R.; Cardella, A.

    2001-01-01

    The L-4 Blanket Project embraces the R and D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R and D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined. The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification tests. (author)

  6. Limiter and first wall of the fusion reactor blanket

    International Nuclear Information System (INIS)

    Danilov, I.; Skladnov, K.; Kolganov, V.

    1994-01-01

    Previous designing of the first wall and limiter has allowed to determine their possible embodiment depending on the parameters and operation conditions of the blanket. As a rule limiter is a separate structure located on the plasma facing surface of the blanket assembly. Possible versions of the limiter/FW which may be considered: (1) limiters with mechanical attachment of the protective part; (2) limiters with the attachment with brazing; (3) limiters with common/separate cooling system; (4) limiter as a substitute of the FW. Generally the FW/limiter structure includes protective shield and its cooling system which consist of protective coating, heat accumulator, conductive layer and attachment locks

  7. Ceramic sphere-pac breeder design for fusion blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.J.; Sullivan, J.D.

    1991-01-01

    Randomly packed beds of ceramic spheres are a practical approach to surrounding fusion plasmas with tritium-breeding material. This paper examines the general properties of sphere-pac beds for application in fusion breeder blankets. The design considerations and models are reviewed for packing, tritium breeding and recovery, thermal conductivity, purge-gas pressure drop, mechanical behavior and fabrication. The design correlations are compared against available fusion ceramic data. Specific conclusions are that ternary (three-size) beds are not attractive for fusion blankets, and that the fusion spheres should be as large as possible subject primarily to packing constraints. (orig.)

  8. An aqueous lithium salt blanket option for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steiner, D.; Varsamis, G. (Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering and Engineering Physics); Deutsch, L.; Rathke, J. (Grumman Corp., Bethpage, NY (USA). Advanced Energy Systems); Gierszewski, P. (Canadian Fusion Fuels Technology Project (CFFTP), Mississauga, ON (Canada))

    1989-04-01

    An aqueous lithium salt blanket (ALSB) concept is proposed which could be the basis for either a power reactor blanket or a test module in an engineering test reactor. The design is based on an austenitic stainless steel structure, a beryllium multiplier, and a salt breeder concentration of about 32 g LiNO/sub 3/ per 100 cm/sup 3/ of H/sub 2/O. To limit tritium release rates, the salt breeder solution is separated from the water coolant circuit. The overall tritium system cost for a 2400 MW (fusion power) reactor is estimated to be 180 million Dollar US87 installed. (orig.).

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

    International Nuclear Information System (INIS)

    Hong, Bong Guen; In, S. R.; Bae, Y. D.

    2006-02-01

    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

  10. 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

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

    International Nuclear Information System (INIS)

    Wang, Q.; Henderson, D.L.

    1995-01-01

    Pulsed activation calculations have been performed on two blanket options considered as part of the ITER home team blanket trade-off study. The objective was to compare the activity, afterheat and waste disposal rating (WDR) results of a composite blanket-shield design for the continuous operation approximation to a pulsed operation case to determine whether the differences are at most the duty factor as predicted by the two nuclide chain model. Up to a cooling period of 100 years, the pulsed activity and afterheat values were below the continuous oepration results and well within (except for one afterheat value) the maximum deviation predicted by the two nuclide chain model. No differences in the WDR values were noted as they are, to a large extent, based on long-lived nuclides which are insensitive to short-term changes in the operation history. (orig.)

  12. A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael

    2016-01-01

    'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

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

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Seki, Yasushi; Minato, Akio; Kobayashi, Takeshi; Mori, Seiji; Kawasaki, Hiromitsu; Sumita, Kenji.

    1983-02-01

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

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

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Pinaev, S.S.; Muraviev, E.V.; Romanov, P.V.

    2005-01-01

    High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease magnetohydrodynamic resistance authors propose to form insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the insulating coatings characteristics ρδ is ∼ 10 -5 Ohm·m 2 for steels and 5,0x10 -6 - 5,0x10 -5 Ohm·m 2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steamgenerators and equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem of technology of lead and lead-bismuth coolants for power high temperature radioactive facilities has been solved. Accidents, emergency situations such as leakage of steamgenerators or depressurization of gas system in facilities with lead and lead-bismuth coolants have been explored and suppressed. (author)

  15. Design requirement on KALIMER blanket fuel assembly duct

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

    1998-03-01

    This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs

  16. Summary of the target-blanket breakout group

    Energy Technology Data Exchange (ETDEWEB)

    Capiello, M.; Bell, C. [Los Alamos National Laboratory, NM (United States); Barthold, W.

    1995-10-01

    This breakout group discussed a number of topics and issues pertaining to target and blanket concepts for accelerator-driven systems. This major component area is one marked by a broad spectrum of technical approaches. It is therefore less defined than other major component areas such as the accelerator and is at an earlier stage of technical needs and task specification. The working group did reach a number of general conclusions and recommendations that are summarized. The Conference and the Target/Blanket Breakout Group provided a first opportunity for people working on a variety of missions and concepts to get together and exchange information. A number of subcritical systems applicable for a spectrum of missions were proposed at the Conference and discussed in the Breakout Group. Missions included plutonium disposition, energy production, waste destruction, isotope production, and neutron scattering. The Target/Blanket Breakout Group also defined areas where parameters and data should be addressed as target/blanket design activities become more detailed and sophisticated.

  17. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  19. Thermal-hydraulic analysis of low activity fusion blanket designs

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.; Yu, W.S.

    1977-01-01

    The heat transfer aspects of fusion blankets are considered where: (a) conduction and (b) boiling and condensation are the dominant heat transfer mechanisms. In some cases, unique heat transfer problems arise and additional heat transfer data and analyses may be required

  20. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    Kleefeldt, K.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-12-01

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

  2. Performance evaluation on force control for ITER blanket installation

    Energy Technology Data Exchange (ETDEWEB)

    Aburadani, A., E-mail: aburadani.atsushi@jaea.go.jp [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Nakahira, M.; Hamilton, D.; Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation.

  3. Performance evaluation on force control for ITER blanket installation

    International Nuclear Information System (INIS)

    Aburadani, A.; Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S.; Nakahira, M.; Hamilton, D.; Tesini, A.

    2013-01-01

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation

  4. Tritium inventory in Li2ZrO3 blanket

    International Nuclear Information System (INIS)

    Nishikawa, M.; Baba, A.

    1998-01-01

    Recently, we have presented the way to estimate the tritium inventory in a solid breeder blanket considering effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reactions. It is reported in our previous paper that the estimated tritium inventory for a LiAlO 2 blanket agrees well with data observed in various in situ experiments when the effective diffusivity of tritium from the EXOTIC-6 experiment is used and that the better agreement is obtained when existence of some water vapor is assumed in the purge gas. The same way as used for a LiAlO 2 blanket is applied to a Li 2 ZrO 3 blanket in this study and the estimated tritium inventory shows a good agreement with data obtained in such in situ experiments as MOZART, EXOTIC-6 and TRINE experiments. (orig.)

  5. Examination of compression and resilience characteristics of fibrous insulation blankets

    International Nuclear Information System (INIS)

    Brislin, R.J.; Middleton, A.

    1979-08-01

    Load-deflection characteristics of alumina and alumino-silicate fibrous blankets were experimentally determined. Load retention and springback capability of combinations of these materials were measured in a 10,000-hour test at surface temperatures of 650 to 1000 0 C (1200 to 1832 0 F). Experimental results are presented and future testing plans are discussed

  6. Effects of buffer thickness on ATW blanket performances

    International Nuclear Information System (INIS)

    Yang, Won Sik

    2001-01-01

    This paper presents the preliminary results of target and buffer design studies for a lead-bismuth eutectic (LBE) cooled accelerator transmutation of waste (ATW) system, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using an 840 MWt LBE cooled ATW design, the effects of buffer thickness on the blanket performances have been studied. Varying the buffer thickness for a given blanket configuration, system performances have been estimated by a series of calculations using MCNPX and REBUS-3 codes. The effects of source importance change are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. As the irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. The results show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable

  7. Helium-cooled pebble bed test blanket module alternative design and fabrication routes

    International Nuclear Information System (INIS)

    Lux, M.

    2007-01-01

    According to first results of the recently started European DEMO study, a new blanket integration philosophy was developed applying so-called multi-module segments. These consist of a number of blanket modules flexibly mounted onto a common vertical manifold structure that can be used for replacing all modules in one segment at one time through vertical remote-handling ports. This principle gives new freedom in the design choices applied to the blanket modules itself. Based on the alternative design options considered for DEMO also the ITER test blanket module was newly analyzed. As a result of these activities it was decided to keep the major principles of the reference design like stiffening grid, breeder unit concept and perpendicular arrangement of pebble beds related to the First Wall because of the very positive results of thermo-mechanical and neutronics studies. The present paper gives an overview on possible further design optimization and alternative fabrication routes. One of the most significant improvements in terms of the hydraulic performance of the Helium cooled reactor can be reached with a new First Wall concept. That concept is based on an internal heat transfer enhancement technique and allows drastically reducing the flow velocity in the FW cooling channels. Small ribs perpendicular to the flow direction (transverse-rib roughness) are arranged on the inner surface of the First Wall cooling channels at the plasma side. In the breeder units cooling plates which are mostly parallel but bent into U-shape at the plasma-side are considered. In this design all flow channels are parallel and straight with the flow entering on one side of the parallel plate sections and exiting on the other side. The ceramic pebble beds are embedded between two pairs of such type of cooling plates. Different modifications could possibly be combined, whereby the most relevant discussed in this paper are (i) rib-cooled First Wall channels, (ii) U-bent cooling plates for

  8. Monte Carlo analysis of helium production in the ITER shielding blanket module

    International Nuclear Information System (INIS)

    Sato, S.

    1999-01-01

    In order to examine the shielding performances of the inboard blanket module in the international thermonuclear experimental reactor (ITER), shielding calculations have been carried out using a three-dimensional Monte Carlo method. The impact of radiation streaming through the front access holes and gaps between adjacent blanket modules on the helium gas production in the branch pipe weld locations and back plate have been estimated. The three-dimensional model represents an 18 sector of the overall torus region and includes the vacuum vessel, inboard blanket and back plate, plasma region, and outboard reflecting medium. And it includes the 1 m high inboard mid-plane module and the 20 mm wide gaps between adjacent modules. From the calculated results for the reference design, it has been found that the helium production at the plug of the branch pipe is four to five times higher than the design goal of 1 appm for a neutron fluence of 0.9 MW a m -2 at the inboard mid-plane first wall. Also, it has been found that the helium production at the back plate behind the horizontal gap is about three times higher than the design goal. In the reference design, the stainless steel (SS):H 2 O composition in the blanket module is 80:20%. Shielding calculations also have been carried out for the SS:H 2 O composition of 70:30, 60:40, 50:50 and 40:60%. From the evaluated results for their design, it has been found that the dependence of helium production on the SS:H 2 170 mm will reduce helium production to satisfy the design goal and not have a significant impact on weight limitations imposed by remote maintenance handling limitations. Also based on the calculated results, about 200 mm thick shields such as a key structure in the vertical gap are suggested to be installed in the horizontal gap as well to reduce the helium production at the back plate and to satisfy the design goal. (orig.)

  9. Low impact of dry conditions on the CO2 exchange of a Northern-Norwegian blanket bog

    International Nuclear Information System (INIS)

    Lund, Magnus; Parmentier, F J W; Bjerke, J W; Tømmervik, H; Drake, B G; Engelsen, O; Hansen, G H; Powell, T L; Silvennoinen, H; Weldon, S; Rasse, D P; Sottocornola, M

    2015-01-01

    Northern peatlands hold large amounts of organic carbon (C) in their soils and are as such important in a climate change context. Blanket bogs, i.e. nutrient-poor peatlands restricted to maritime climates, may be extra vulnerable to global warming since they require a positive water balance to sustain their moss dominated vegetation and C sink functioning. This study presents a 4.5 year record of land–atmosphere carbon dioxide (CO 2 ) exchange from the Andøya blanket bog in northern Norway. Compared with other peatlands, the Andøya peatland exhibited low flux rates, related to the low productivity of the dominating moss and lichen communities and the maritime settings that attenuated seasonal temperature variations. It was observed that under periods of high vapour pressure deficit, net ecosystem exchange was reduced, which was mainly caused by a decrease in gross primary production. However, no persistent effects of dry conditions on the CO 2 exchange dynamics were observed, indicating that under present conditions and within the range of observed meteorological conditions the Andøya blanket bog retained its C uptake function. Continued monitoring of these ecosystem types is essential in order to detect possible effects of a changing climate. (letter)

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

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakanhira, Masataka; Matsumoto, Yasuhiro; Shibanuma, K.

    2007-01-01

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

  11. 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.

  12. Evaluation of compost blankets for erosion control from disturbed lands.

    Science.gov (United States)

    Bhattarai, Rabin; Kalita, Prasanta K; Yatsu, Shotaro; Howard, Heidi R; Svendsen, Niels G

    2011-03-01

    Soil erosion due to water and wind results in the loss of valuable top soil and causes land degradation and environmental quality problems. Site specific best management practices (BMP) are needed to curb erosion and sediment control and in turn, increase productivity of lands and sustain environmental quality. The aim of this study was to investigate the effectiveness of three different types of biodegradable erosion control blankets- fine compost, mulch, and 50-50 mixture of compost and mulch, for soil erosion control under field and laboratory-scale experiments. Quantitative analysis was conducted by comparing the sediment load in the runoff collected from sloped and tilled plots in the field and in the laboratory with the erosion control blankets. The field plots had an average slope of 3.5% and experiments were conducted under natural rainfall conditions, while the laboratory experiments were conducted at 4, 8 and 16% slopes under simulated rainfall conditions. Results obtained from the field experiments indicated that the 50-50 mixture of compost and mulch provides the best erosion control measures as compared to using either the compost or the mulch blanket alone. Laboratory results under simulated rains indicated that both mulch cover and the 50-50 mixture of mulch and compost cover provided better erosion control measures compared to using the compost alone. Although these results indicate that the 50-50 mixtures and the mulch in laboratory experiments are the best measures among the three erosion control blankets, all three types of blankets provide very effective erosion control measures from bare-soil surface. Results of this study can be used in controlling erosion and sediment from disturbed lands with compost mulch application. Testing different mixture ratios and types of mulch and composts, and their efficiencies in retaining various soil nutrients may provide more quantitative data for developing erosion control plans. Copyright © 2010 Elsevier

  13. Structural analysis under the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Majumdar, S.

    1985-01-01

    Structural design procedures followed in the Blanket Comparison and Selection Study are briefly reviewed. The American Society of Mechanical Engineers Boilers and Pressure Vessels Code, Section III, Code Case N47 has been used as a design guide. Its relevance to fusion reactor applications, however, is open to question and needs to be evaluated in the future. The primary structural problem encountered in tokamak blanket designs is the high thermal stress due to surface heat flux, with fatigue being an additional concern for pulsed systems. The conflicting requirements of long erosion life and high surface heat flux capability imply that some form of stress relief in the first-wall region will be necessary. Simplified stress and fatigue crack growth analyses are presented to show that the use of orthogonally grooved first wall may be a potential solution for mitigating the thermal stress problem. A comparison of three structural alloys on the basis of both grooved and nongrooved first-wall designs is also presented. Other structural problems encountered in tokamak designs include stresses due to plasma disruptions, and magnetohydrodynamic (MHD) pressure drop in liquid-metal-cooled systems. In particular, it is shown that the maximum stress in the side wall of a uniform duct generated by MHD pressure drop cannot be reduced by increasing the wall thickness or by decreasing the span. In contract to tokamak blankets, tandem mirror blankets are far less severely stressed because of a much lower surface heat flux, coolant pressure, and also because of their axisymmetric geometry. Both blankets, however, will require detailed structural dynamics analysis to verify their ability to withstand seismic loadings if the heavy 17Li-83Pb is used as a coolant

  14. Overview and status of ITER internal components

    International Nuclear Information System (INIS)

    Merola, Mario; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

    2014-01-01

    Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine

  15. Overview and status of ITER internal components

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

    2014-10-15

    Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine.

  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. Distortion of the activation energy of high temperature internal friction background due to temperature dependence frequency variations

    International Nuclear Information System (INIS)

    Lambri, O.; Povolo, F.; Molinas, B.

    1991-01-01

    In this work, a study is made of how the variation of frequency with temperature affects an activation enthalpy. This effect is usually neglected, but in some cases like Cu-Au or Zry-4 (an alloy of nuclear interest base or Zr alloyed with Sn, Fe and Cr) such variation can rise up to as much as 16%/4/ and 37%/5/. (Author) [es

  18. Study on the temperature control mechanism of the tritium breeding blanket for CFETR

    Science.gov (United States)

    Liu, Changle; Qiu, Yang; Zhang, Jie; Zhang, Jianzhong; Li, Lei; Yao, Damao; Li, Guoqiang; Gao, Xiang; Wu, Songtao; Wan, Yuanxi

    2017-12-01

    The Chinese fusion engineering testing reactor (CFETR) will demonstrate tritium self- sufficiency using a tritium breeding blanket for the tritium fuel cycle. The temperature control mechanism (TCM) involves the tritium production of the breeding blanket and has an impact on tritium self-sufficiency. In this letter, the CFETR tritium target is addressed according to its missions. TCM research on the neutronics and thermal hydraulics issues for the CFETR blanket is presented. The key concerns regarding the blanket design for tritium production under temperature field control are depicted. A systematic theory on the TCM is established based on a multiplier blanket model. In particular, a closed-loop method is developed for the mechanism with universal function solutions, which is employed in the CFETR blanket design activity for tritium production. A tritium accumulation phenomenon is found close to the coolant in the blanket interior, which has a very important impact on current blanket concepts using water coolant inside the blanket. In addition, an optimal tritium breeding ratio (TBR) method based on the TCM is proposed, combined with thermal hydraulics and finite element technology. Meanwhile, the energy gain factor is adopted to estimate neutron heat deposition, which is a key parameter relating to the blanket TBR calculations, considering the structural factors. This work will benefit breeding blanket engineering for the CFETR reactor in the future.

  19. Preliminary analyses of neutronics schemes for three kinds waste transmutation blankets of fusion-fission hybrid

    International Nuclear Information System (INIS)

    Zhang Mingchun; Feng Kaiming; Li Zaixin; Zhao Fengchao

    2012-01-01

    The neutronics schemes of the helium-cooled waste transmutation blanket, sodium-cooled waste transmutation blanket and FLiBe-cooled waste transmutation blanket were preliminarily calculated and analysed by using the spheroidal tokamak (ST) plasma configuration. The neutronics properties of these blankets' were compared and analyzed. The results show that for the transmutation of "2"3"7Np, FLiBe-cooled waste transmutation blanket has the most superior transmutation performance. The calculation results of the helium-cooled waste transmutation blanket show that this transmutation blanket can run on a steady effective multiplication factor (k_e_f_f), steady power (P), and steady tritium production rate (TBR) state for a long operating time (9.62 years) by change "2"3"7Np's initial loading rate of the minor actinides (MA). (authors)

  20. International Geographic Variation in Event Rates in Trials of Heart Failure With Preserved and Reduced Ejection Fraction

    DEFF Research Database (Denmark)

    Kristensen, Søren L; Køber, Lars; Jhund, Pardeep S

    2015-01-01

    BACKGROUND: International geographic differences in outcomes may exist for clinical trials of heart failure and reduced ejection fraction (HF-REF), but there are few data for those with preserved ejection fraction (HF-PEF). METHODS AND RESULTS: We analyzed outcomes by international geographic reg...

  1. Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.; Embrechts, M.J.

    1987-11-01

    The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs

  2. Spatiotemporal variation of radon and carbon dioxide concentrations in an underground quarry: coupled processes of natural ventilation, barometric pumping and internal mixing.

    Science.gov (United States)

    Perrier, Frédéric; Richon, Patrick

    2010-04-01

    Radon-222 and carbon dioxide concentrations have been measured during several years at several points in the atmosphere of an underground limestone quarry located at a depth of 18 m in Vincennes, near Paris, France. Both concentrations showed a seasonal cycle. Radon concentration varied from 1200 to 2000 Bq m(-3) in summer to about 800-1400 Bq m(-3) in winter, indicating winter ventilation rates varying from 0.6 to 2.5 x 10(-6) s(-1). Carbon dioxide concentration varied from 0.9 to 1.0% in summer, to about 0.1-0.3% in winter. Radon concentration can be corrected for natural ventilation using temperature measurements. The obtained model also accounts for the measured seasonal variation of carbon dioxide. After correction, radon concentrations still exhibit significant temporal variation, mostly associated with the variation of atmospheric pressure, with coupling coefficients varying from -7 to -26 Bq m(-3) hPa(-1). This variation can be accounted for using a barometric pumping model, coupled with natural ventilation in winter, and including internal mixing as well. After correction, radon concentrations exhibit residual temporal variation, poorly correlated between different points, with standard deviations varying from 3 to 6%. This study shows that temporal variation of radon concentrations in underground cavities can be understood to a satisfactory level of detail using non-linear and time-dependent modelling. It is important to understand the temporal variation of radon concentrations and the limitations in their modelling to monitor the properties of natural or artificial underground settings, and to be able to assess the existence of new processes, for example associated with the preparatory phases of volcanic eruptions or earthquakes. Copyright 2009 Elsevier Ltd. All rights reserved.

  3. Spatiotemporal variation of radon and carbon dioxide concentrations in an underground quarry: coupled processes of natural ventilation, barometric pumping and internal mixing

    International Nuclear Information System (INIS)

    Perrier, Frederic; Richon, Patrick

    2010-01-01

    Radon-222 and carbon dioxide concentrations have been measured during several years at several points in the atmosphere of an underground limestone quarry located at a depth of 18 m in Vincennes, near Paris, France. Both concentrations showed a seasonal cycle. Radon concentration varied from 1200 to 2000 Bq m -3 in summer to about 800-1400 Bq m -3 in winter, indicating winter ventilation rates varying from 0.6 to 2.5 x 10 -6 s -1 . Carbon dioxide concentration varied from 0.9 to 1.0% in summer, to about 0.1-0.3% in winter. Radon concentration can be corrected for natural ventilation using temperature measurements. The obtained model also accounts for the measured seasonal variation of carbon dioxide. After correction, radon concentrations still exhibit significant temporal variation, mostly associated with the variation of atmospheric pressure, with coupling coefficients varying from -7 to -26 Bq m -3 hPa -1 . This variation can be accounted for using a barometric pumping model, coupled with natural ventilation in winter, and including internal mixing as well. After correction, radon concentrations exhibit residual temporal variation, poorly correlated between different points, with standard deviations varying from 3 to 6%. This study shows that temporal variation of radon concentrations in underground cavities can be understood to a satisfactory level of detail using non-linear and time-dependent modelling. It is important to understand the temporal variation of radon concentrations and the limitations in their modelling to monitor the properties of natural or artificial underground settings, and to be able to assess the existence of new processes, for example associated with the preparatory phases of volcanic eruptions or earthquakes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-15

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

  5. Fusion materials: Technical evaluation of the technology of vandium alloys for use as blanket structural materials in fusion power systems

    International Nuclear Information System (INIS)

    1993-01-01

    The Committee's evaluation of vanadium alloys as a structural material for fusion reactors was constrained by limited data and time. The design of the International Thermonuclear Experimental Reactor is still in the concept stage, so meaningful design requirements were not available. The data on the effect of environment and irradiation on vanadium alloys were sparse, and interpolation of these data were made to select the V-5Cr-5Ti alloy. With an aggressive, fully funded program it is possible to qualify a vanadium alloy as the principal structural material for the ITER blanket in the available 5 to 8-year window. However, the data base for V-5Cr-5Ti is United and will require an extensive development and test program. Because of the chemical reactivity of vanadium the alloy will be less tolerant of system failures, accidents, and off-normal events than most other candidate blanket structural materials and will require more careful handling during fabrication of hardware. Because of the cost of the material more stringent requirements on processes, and minimal historical worlding experience, it will cost an order of magnitude to qualify a vanadium alloy for ITER blanket structures than other candidate materials. The use of vanadium is difficult and uncertain; therefore, other options should be explored more thoroughly before a final selection of vanadium is confirmed. The Committee views the risk as being too high to rely solely on vanadium alloys. In viewing the state and nature of the design of the ITER blanket as presented to the Committee, h is obvious that there is a need to move toward integrating fabrication, welding, and materials engineers into the ITER design team. If the vanadium allay option is to be pursued, a large program needs to be started immediately. The commitment of funding and other resources needs to be firm and consistent with a realistic program plan

  6. Measurement of Internal Friction for Tungsten by the Curve Vibrating Method with Variation of Voltage and Temperature

    Directory of Open Access Journals (Sweden)

    Elin Yusibani

    2013-12-01

    Full Text Available Application of a curved vibrating wire method (CVM to measure gas viscosity has been widely used. A fine Tungsten wire with 50 mm of diameter is bent into a semi-circular shape and arranged symmetrically in a magnetic field of about 0.2 T. The frequency domain is used for calculating the viscosity as a response for forced oscillation of the wire. Internal friction is one of the parameter in the CVM which is has to be measured beforeahead. Internal friction coefficien for the wire material which is the inverse of the quality factor has to be measured in a vacuum condition. The term involving internal friction actually represents the effective resistance of motion due to all non-viscous damping phenomena including internal friction and magnetic damping. The testing of internal friction measurement shows that at different induced voltage and elevated temperature at a vacuum condition, it gives the value of internal friction for Tungsten is around 1 to 4 10-4.

  7. APT 3He target/blanket. Topical report

    International Nuclear Information System (INIS)

    1995-03-01

    The 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 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

  8. Experimental program for the Fast Breeder Blanket Facility, FBBF

    International Nuclear Information System (INIS)

    Ott, K.O.; Clikeman, F.M.; Johnson, R.H.; Borg, R.C.

    1976-01-01

    The work performed in the reporting period was primarily concerned with the development of the experimental program (Task A) and with the pre-analysis of future loadings and the impact upon the permanent loading of the two converter regions, which contain 4.8 percent enriched UO 2 rods. It appears necessary that a neutron poison (B 4 C) be placed in the converter (transformer) regions in order to hold, also for future loadings, the k/sub eff/ of a hypothetically flooded FBBF well below 1. Since it is planned to use the same welded converter regions for all experiments, the required B 4 C loading needs to be determined prior to the first blanket loading. Further the equipment needs have been identified (Task D), the 252 Cf-source has been requested on a loan basis (Task E). First discussions with ANL on blanket experiments have been initiated

  9. Rapid thermal cycling of new technology solar array blanket coupons

    Science.gov (United States)

    Scheiman, David A.; Smith, Bryan K.; Kurland, Richard M.; Mesch, Hans G.

    1990-01-01

    NASA Lewis Research Center is conducting thermal cycle testing of a new solar array blanket technologies. These technologies include test coupons for Space Station Freedom (SSF) and the advanced photovoltaic solar array (APSA). The objective of this testing is to demonstrate the durability or operational lifetime of the solar array interconnect design and blanket technology within a low earth orbit (LEO) or geosynchronous earth orbit (GEO) thermal cycling environment. Both the SSF and the APSA array survived all rapid thermal cycling with little or no degradation in peak performance. This testing includes an equivalent of 15 years in LEO for SSF test coupons and 30 years of GEO plus ten years of LEO for the APSA test coupon. It is concluded that both the parallel gap welding of the SSF interconnects and the soldering of the APSA interconnects are adequately designed to handle the thermal stresses of space environment temperature extremes.

  10. Stability properties of cold blanket systems for current driven modes

    International Nuclear Information System (INIS)

    Ohlsson, D.

    1977-12-01

    The stability problem of the boundary regions of cold blanket systems with induced currents parallel to the lines of force is formulated. Particular interest is focused on two types of modes: first electrostatic modes driven by the combined effects of a transverse resistivity gradient due to a spatially non-uniform electron temperature and a longitudinal current, second electromagnetic kink like modes driven by the torque arising from a transverse current density gradient and magnetic field perturbations. It is found that the combination of various dissipative and neutral gas effects introduces strong stabilizing effects within specific parameter ranges. For particular steady-state models investigated it is shown that these effects become of importance in laboratory plasmas at relatively high densities, low temperatures and moderate magnetic field strengths. Stability diagrams based on specific steady-state cold plasma blanket models will be presented

  11. 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.

  12. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    Jassby, D.L.; Leinoff, S.

    1979-12-01

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

  13. 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

  14. 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.

  15. Tritium breeding blanket device of D-T reactors

    International Nuclear Information System (INIS)

    Chevereau, G.

    1984-01-01

    This blanket device uses solid tritium breeding materials as those which include, in a known manner, near a neutron breeding plasma, a neutron multiplier medium and a tritium breeding medium, cooled by a cooling fluid circulation. This device is characterized by the fact that the association of the multiplier media and the tritium breeding media is realized by pellet alternated piling up of each of those both media, help in close contact on all their lateral surfaces [fr

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

    International Nuclear Information System (INIS)

    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 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 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 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

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

    International Nuclear Information System (INIS)

    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 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 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 2 O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  18. Evaluation of US demo helium-cooled blanket options

    International Nuclear Information System (INIS)

    Wong, C.P.C.; McQuillan, B.W.; Schleicher, R.W.

    1995-10-01

    A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed

  19. Heat Loads Due To Small Penetrations In Multilayer Insulation Blankets

    Science.gov (United States)

    Johnson, W. L.; Heckle, K. W.; E Fesmire, J.

    2017-12-01

    The main penetrations (supports and piping) through multilayer insulation systems for cryogenic tanks have been previously addressed by heat flow measurements. Smaller penetrations due to fasteners and attachments are now experimentally investigated. The use of small pins or plastic garment tag fasteners to ease the handling and construction of multilayer insulation (MLI) blankets goes back many years. While it has long been understood that penetrations and other discontinuities degrade the performance of the MLI blanket, quantification of this degradation has generally been lumped into gross performance multipliers (often called degradation factors or scale factors). Small penetrations contribute both solid conduction and radiation heat transfer paths through the blanket. The conduction is down the stem of the structural element itself while the radiation is through the hole formed during installation of the pin or fastener. Analytical models were developed in conjunction with MLI perforation theory and Fourier’s Law. Results of the analytical models are compared to experimental testing performed on a 10 layer MLI blanket with approximately 50 small plastic pins penetrating the test specimen. The pins were installed at ∼76-mm spacing inches in both directions to minimize the compounding of thermal effects due to localized compression or lateral heat transfer. The testing was performed using a liquid nitrogen boil-off calorimeter (Cryostat-100) with the standard boundary temperatures of 293 K and 78 K. Results show that the added radiation through the holes is much more significant than the conduction down the fastener. The results are shown to be in agreement with radiation theory for perforated films.

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

    International Nuclear Information System (INIS)

    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 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 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 2 O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  1. Choice of economical optimum blanket of hybrid reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blinkin, V L; Novikov, V M

    1981-01-01

    The economical effectiveness of symbiotic power systems depends on the choice of the correlation between energy production and fissile fuel production in blankets of controlled thermonuclear fusion reactor (CTR), what is investigated here. It is shown that the optimum value of this correlation essentially depends on the ratio between the specific costs for energy production in hybrid thermonuclear reactors and that in fission reactors as part of the symbiotic system.

  2. Development of insulating coatings for liquid metal blankets

    International Nuclear Information System (INIS)

    Malang, S.; Borgstedt, H.U.; Farnum, E.H.; Natesan, K.; Vitkovski, I.V.

    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

  3. Li2O-pebble type tritium breeding blanket for fusion experimental reactor, 1

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

    1984-01-01

    The fusion experimental reactor is the next stage device in Japan, which is planned to be constructed following the critical plasma experimental device JT-60 being constructed at present. The breeding blanket installed in nuclear fusion reactors is one of most important structures, and it is required to satisfy the fundamental performance of producing and continuously recovering tritium as the nuclear fusion fuel, and other requirement in good coordination. The Li 2 O pebble type breeding blanket that Kawasaki Heavy Industries Ltd. has examined is the concept for resolving the problems of the mass transfer and thermal stress cracking of Li 2 O, which are important in blanket design. In this paper, the concept and characteristics of this breeding blanket are discussed from the viewpoint of the breeding and continuous recovery of tritium, the ease of manufacture and the maintenance of soundness. The breeding blanket is composed of breeding region, tritium purge region, cooling region, plasma stabilizing conductors and blanket container. Li 2 O is excellent in its tritium breeding performance and heat conductivity. The functions required for the breeding blanket, the fundamental structure, the examples of breeding blanket concept, the selection of breeding blanket concept, the characteristics of Li 2 O pebble type blanket and its future prospect are described. (Kako, I.)

  4. Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

    2016-11-01

    Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

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

    International Nuclear Information System (INIS)

    Daly, E.; Pletzer, R.

    1992-04-01

    An 80-K Multi Layer Insulation (MLI) blanket in the interconnect region between magnets DD0019 and DD0027 in the Fermi National Accelerator Laboratory (FNAL) ER string was instrumented with temperature sensors to obtain the steady-state temperature gradient through the blanket after string cooldown. A thermal model of the 80-K blanket assembly was constructed to analyze the steady-state temperature gradient data. Estimates of the heat flux through the 80-K MLI blanket assembly and predicted temperature gradients were calculated. The thermal behavior of the heavy polyethylene terapthalate (PET) cover layers separating the shield and inner blanket and inner and outer blankets was derived empirically from the data. The results of the analysis predict a heat flux of 0.363--0.453 W/m 2 based on the 11 sets of data. These flux values are 33--46% below the 80-K MLI blanket heat leak budget of 0.676 W/m 2 . The effective thermal resistance of the two heavy PET cover layers between the shield and inner blanket was found to be 2.1 times that of a single PET spacer layer, and the effective resistance of the two heavy PET cover layers between the inner blanket and outer blanket was found to be 7 times that of a single PET spacer layer. Based on these results, the 80-K MLI blanket assembly appears to be performing more than adequately to meet the 80-K static IR heat leak budget. However, these results should not be construed as a verification of the 80-K static IR heat leak, since no actual heat leak was measured. The results have been used to improve the empirically based model data in the 80-K MLI blanket thermal model, which has previously not included the effects of heavy PET cover layers on 80-K MLI blanket thermal performance

  6. Low cost, high yield IFE reactors: Revisiting Velikhov's vaporizing blankets

    International Nuclear Information System (INIS)

    Logan, B.G.

    1992-01-01

    The performance (efficiency and cost) of IFE reactors using MHD conversion is explored for target blanket shells of various materials vaporized and ionized by high fusion yields (5 to 500 GJ). A magnetized, prestressed reactor chamber concept is modeled together with previously developed models for the Compact Fusion Advanced Rankine II (CFARII) MHD Balance-of-Plant (BoP). Using conservative 1-D neutronics models, high fusion yields (20 to 80 GJ) are found necessary to heat Flibe, lithium, and lead-lithium blankets to MHD plasma temperatures, at initial solid thicknesses sufficient to capture most of the fusion yield. Advanced drivers/targets would need to be developed to achieve a ''Bang per Buck'' figure-of-merit approx-gt 20 to 40 joules yield per driver $ for this scheme to be competitive with these blanket materials. Alternatively, more realistic neutronics models and better materials such as lithium hydride may lower the minimum required yields substantially. The very low CFARII BoP costs (contributing only 3 mills/kWehr to CoE) allows this type of reactor, given sufficient advances that non-driver costs dominate, to ultimately produce electricity at a much lower cost than any current nuclear plant

  7. Comparison of inventory of tritium in various ceramic breeder blankets

    International Nuclear Information System (INIS)

    Nishikawa, M.; Beloglazov, S.; Nakashima, N.; Hashimoto, K.; Enoeda, M.

    2002-01-01

    It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of breeder material, absorption of water vapor into bulk of the grain, and adsorption of water on surface of the grain, together with the isotope exchange reaction between hydrogen in purge gas and tritium on surface of breeder material and the isotope exchange reaction between water vapor in purge gas and tritium on surface, for estimation of the tritium inventory in a uniform ceramic breeder blanket under the steady-state condition. It has been also pointed out by the present authors that the water formation reaction on the surface of ceramic breeder materials at introduction of hydrogen can give effect on behavior of bred tritium and lithium transfer in blanket. The tritium inventory for various ceramic breeder blankets are compared in this study basing on adsorption capacity, absorption capacity, isotope exchange capacity, and isotope exchange reactions on the Li 2 O, LiAlO 2 , Li 2 ZrO 3 , Li 4 SiO 4 and Li 2 TiO 3 surface experimentally obtained by the present authors. Effect of each mass transfer steps on the shape of release curve of bred tritium at change of the operational conditions is also discussed from the observation at out pile experiment in KUR. (orig.)

  8. Space environment durability of beta cloth in LDEF thermal blankets

    Science.gov (United States)

    Linton, Roger C.; Whitaker, Ann F.; Finckenor, Miria M.

    1993-01-01

    Beta cloth performance for use on long-term space vehicles such as Space Station Freedom (S.S. Freedom) requires resistance to the degrading effects of the space environment. The major issues are retention of thermal insulating properties through maintaining optical properties, preserving mechanical integrity, and generating minimal particulates for contamination-sensitive spacecraft surfaces and payloads. The longest in-flight test of beta cloth's durability was on the Long Duration Exposure Facility (LDEF), where it was exposed to the space environment for 68 months. The LDEF contained 57 experiments which further defined the space environment and its effects on spacecraft materials. It was deployed into low-Earth orbit (LEO) in Apr. 1984 and retrieved Jan. 1990 by the space shuttle. Among the 10,000 plus material constituents and samples onboard were thermal control blankets of multilayer insulation with a beta cloth outer cover and Velcro attachments. These blankets were exposed to hard vacuum, thermal cycling, charged particles, meteoroid/debris impacts, ultraviolet (UV) radiation, and atomic oxygen (AO). Of these space environmental exposure elements, AO appears to have had the greatest effect on the beta cloth. The beta cloth analyzed in this report came from the MSFC Experiment S1005 (Transverse Flat-Plate Heat Pipe) tray oriented approximately 22 deg from the leading edge vector of the LDEF satellite. The location of the tray on LDEF and the placement of the beta cloth thermal blankets are shown. The specific space environment exposure conditions for this material are listed.

  9. Effect of blanket assembly shuffling on LMR neutronic performance

    International Nuclear Information System (INIS)

    Khalil, H.; Fujita, E.K.

    1987-01-01

    Neutronic analyses of advanced liquid-metal reactors (LMRs) have generally been performed with assemblies in different batches scatter-loaded but not shuffled among the core lattice positions between cycles. While this refueling approach minimizes refueling time, significant improvements in thermal performance are believed to be achievable by blanket assembly shuffling. These improvements, attributable to mitigation of the early-life overcooling of the blankets, include reductions in peak clad temperatures and in the temperature gradients responsible for thermal striping. Here the authors summarize results of a study performed to: (1) assess whether the anticipated gains in thermal performance can be realized without sacrificing core neutronic performance, particularly the burnup reactivity swing rho/sub bu/, which determines the rod ejection worth; (2) determine the effect of various blanket shuffling operations on reactor performance; and (3) determine whether shuffling strategies developed for an equilibrium (plutonium-fueled) core can be applied during the transition from an initial uranium-fueled core as is being considered in the US advanced LMR program

  10. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

    Schuller, M.J.

    1985-01-01

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

  11. NOEL: a no-leak fusion blanket concept

    International Nuclear Information System (INIS)

    Powell, J.R.; Yu, W.S.; Fillo, J.A.; Horn, F.L.; Makowitz, H.

    1980-01-01

    Analysis and tests of a no-leak fusion blanket concept (NOEL-NO External Leak) are described. Coolant cannot leak into the plasma chamber even if large through-cracks develop in the first wall. Blanket modules contain a two-phase material, A, that is solid (several cm thick) on the inside of the module shell, and liquid in the interior. The solid layer is maintained by imbedded tubes carrying a coolant, B, below the freezing point of A. Most of the 14-MeV neutron energy is deposited as heat in the module interior. The thermal energy flow from the module interior to the shell keeps the interior liquid. Pressure on the liquid A interior is greater than the pressure on B, so that B cannot leak out if failures occur in coolant tubes. Liquid A cannot leak into the plasma chamber through first wall cracks because of the intervening frozen layer. The thermal hydraulics and neutronics of NOEL blankets have been investigated for various metallic (e.g., Li, Pb 2 , LiPb, Pb) and fused salt choices for material A

  12. Neutronic investigation and activation calculation for CFETR HCCB blankets

    Science.gov (United States)

    Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

    2017-12-01

    The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10-4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

  13. Measurements relevant to simulating subcriticality in ADS facilities with blanket

    International Nuclear Information System (INIS)

    Titarenko, Yu. E.; Batyaev, V.F.; Borovlev, S.P.; Gladkikh, N.G.; Igumnov, M.M.; Legostaev, V.O.; Karpikhin, E.I.; Konev, V.N.; Kushnerev, Yu.T.; Popkov, V.N.; Ryazhsky, V.I.; Spiridonov, V.G.; Chernyavsky, E.V.; Shvedov, O.V.

    2009-10-01

    The work presents the results of determining the blanket subcriticality for a zero-power heavy water reactor MAKET at the Institute for Theoretical and Experimental Physics, Moscow. The blanket is hexagonal lattice made of 36 90%-enriched 235U fuel rods spaced 173mm apart. The subcriticality was varied from ∼0.3% to 5% by adjusting the heavy water level. The subcriticality values were calibrated using the dependence of reactivity on heavy water level. The pulsed neutron source technique was used to measure the temporal dependence of neutron field at different blanket points for the calibrated subcriticality values. The subciticality values obtained in terms of the 'inverse clock' formulae using the decay constants of the measured dependences proved to differ from the calibrated subcriticalities by not more than 7% at the average. The MCNP code-aided simulations of the experiment made has given the calibrated keff values at prescribed heavy water levels and led to the neutron field decay constants at given points, which differ on the average from their experimental values by not more than 7% too. (author)

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

    International Nuclear Information System (INIS)

    Proust, E.; Gervaise, F.; Carre, F.; Chevereau, G.; Doutriaux, D.

    1986-09-01

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

  15. Thermo-mechanical characterization of ceramic pebbles for breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lo Frano, Rosa, E-mail: rosa.lofrano@ing.unipi.it; Aquaro, Donato; Scaletti, Luca

    2016-11-01

    Highlights: • Experimental activities to characterize the Li{sub 4}SiO{sub 4}. • Compression tests of pebbles. • Experimental evaluation of thermal conductivity of pebbles bed at different temperatures. • Experimental test with/without compression load. - Abstract: An open issue for fusion power reactor is to design a suitable breeding blanket capable to produce the necessary quantity of the tritium and to transfer the energy of the nuclear fusion reaction to the coolant. The envisaged solution called Helium-Cooled Pebble Bed (HCPB) breeding blanket foresees the use of lithium orthosilicate (Li{sub 4}SiO{sub 4}) or lithium metatitanate (Li{sub 2}TiO{sub 3}) pebble beds. The thermal mechanical properties of the candidate pebble bed materials are presently extensively investigated because they are critical for the feasibility and performances of the numerous conceptual designs which use a solid breeder. This study is aimed at the investigation of mechanical properties of the lithium orthosilicate and at the characterization of the main chemical, physical and thermo-mechanical properties taking into account the production technology. In doing that at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa adequate experiments were carried out. The obtained results may contribute to characterize the material of the pebbles and to optimize the design of the envisaged fusion breeding blankets.

  16. Mirror hybrid reactor blanket and power conversion system conceptual design

    International Nuclear Information System (INIS)

    Schultz, K.R.; Backus, G.A.; Baxi, C.B.; Dee, J.B.; Estrine, E.A.; Rao, R.; Veca, A.R.

    1976-01-01

    The conceptual design of the blanket and power conversion system for a gas-cooled mirror hybrid fusion-fission reactor is presented. The designs of the fuel, blanket module and power conversion system are based on existing gas-cooled fission reactor technology that has been developed at General Atomic Company. The uranium silicide fuel is contained in Inconel-clad rods and is cooled by helium gas. The fuel is contained in 16 spherical segment modules which surround the fusion plasma. The hot helium is used to raise steam for a conventional steam cycle turbine generator. The details of the method of support for the massive blanket modules and helium ducts remain to be determined. Nevertheless, the conceptual design appears to be technically feasible with existing gas-cooled technology. A preliminary safety analysis shows that with the development of a satisfactory method of primary coolant circuit containment and support, the hybrid reactor could be licensed under existing Nuclear Regulatory Commission regulations

  17. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

    Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

  18. A study of sodium-cooled fast breeder reactor with thorium blanket for supply of U-233 to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Yoshida, H.; Nishimura, H.; Osugi, T.

    1978-08-01

    Symbiotic energy system between fast breeder reactor and thermal reactor would have a potential merit for nuclear proliferation problem. And when using HTGR as the thermal reactor in the system, the energy system appears to be promising as an energy system self-sufficient in fuels, which can generate both electricity and high temperature process heat. In the system the fast breeder reactor has to supply sufficient amount of fissile plutonium to keep the reactor going, and also produce U-233 necessary to the associated U-233 fuelled process heat production HTGR. Three types of LMFBR concepts with thorium blanket, conventional homogeneous core LMFBR, and axial and radial parfait heterogeneous core LMFBRs, have been investigated to find out suitable configurations of LMFBR for supply of U-233 to the HTGR with relatively high conversion ratio of 0.85, in the symbiotic energy system between LMFBR and HTGR. The investigation on LMFBR has been made on fuel sufficiency of the system, inherent safety such as sodium-void and Doppler coefficients, and fuel cycle cost. The followings were revealed; (1) Conventional homogeneous core LMFBR with thorium radial blanket well satisfies the condition of fuel sufficiency, if adequate radial blanket thickness is chosen. However, the sodium-void coefficient and fuel cycle cost are inferior to the other concepts. (2) Axial parfait heterogeneous core LMFBR can be regarded as one of the best LMFBR concepts installed in the symbiotic energy system, from the viewpoints of fuel sufficiency, inherent safety and fuel cycle cost. However, further investigations should be needed on reliability and operationability of the concept. (3) Radial parfait heterogeneous core LMFBR seems inadequate as the LMFBR in the system, because the configurations based on this concept does not satisfy plutonium and U-233 breedings, simultaneously. This LMFBR concept, however, has excellent breeding performance in the internal radial blanket. So further

  19. Thermohydraulics design and thermomechanics analysis of two European breeder blanket concepts for DEMO. Pt. 1 and Pt. 2. Pt. 1: BOT helium cooled solid breeding blanket. Pt. 2: Dual coolant self-cooled liquid metal blanket

    International Nuclear Information System (INIS)

    Norajitra, P.

    1995-06-01

    Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li 4 SiO 4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.) [de

  20. DEMO relevance of the test blanket modules in ITER-Application to the European test blanket modules

    International Nuclear Information System (INIS)

    Magnani, E.; Gabriel, F.; Boccaccini, L.V.; Li-Puma, A.

    2010-01-01

    Test blanket module (TBM) testing programme in ITER as a support to DEMO design is a very important step on the road map to commercial fusion reactors although it is an ambitious task. Finding as much as possible DEMO relevant tests in view of the future DEMO blanket design is therefore a major goal since ITER and DEMO environment and loading conditions are different. To clarify and quantify the meaning of the DEMO relevance, criteria using a structural, functional and behavioural representation of the breeding blanket acting as a system are investigated. Then, a three-step strategy is proposed to carry out TBM DEMO relevant tests associated with a TBM design modification strategy. Key parameters should intensively be used as target for TBM characterization and numerical code validation. When assessing the relevance, on the other hand, not only the actual difference between DEMO and ITER values should be considered, but also whether the analyzed phenomena have a threshold and a range of applicability, as numerical simulations are usually permitted within these limits. The proposed methodology is at the end applied to the design of the HCLL TBM breeding unit configuration.

  1. 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.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    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. (paper)

  3. Applications of the aqueous self-cooled blanket (ASCB) concept to the Next European Torus (NET)

    International Nuclear Information System (INIS)

    Embrechts, M.J.; Bogaerts, W.; Cardella, A.; Chazalon, M.; Danner, W.; Dinner, P.; Libin, B.

    1987-01-01

    The Aqueous Self-Cooled Blanket Concept (ASCB) leads to a low-technology blanket design that relies on just structural material and coolant with small amounts of lithium compound dissolved in the coolant to provide for tritium production. The application of the ASCB concept in NET is being considered as a driver blanket that would operate at low temperature and low pressure and provide a reliable environment for machine operation during the technology phase. Shielding and tritium production are the primary objectives for such a low-technology blanket. Net tritium breeding is not a design requirement per se for a driver blanket for NET. A DEMO relevant ASCB based blanket test module with (local) tritium self-sufficiency and energy recovery as primary objectives might also be tested in NET if future developments confirm their viability

  4. Will blocking historical drainage ditches increase carbon sequestration in upland blanket mires of Southwest England?

    Science.gov (United States)

    Le Feuvre, N.; Hartley, I.; Anderson, K.; Luscombe, D.; Grand-Clement, E.; Smith, D.; Brazier, R.

    2012-04-01

    Peat soils in the United Kingdom are estimated to store a minimum of 3,121Mt C (Lindsay, 2010). Despite being such a large carbon store the annual imbalance between uptake and release is small and susceptible to change in response to land management, atmospheric deposition and climate change. The upland blanket mires of Southwest England have been subject to extensive drainage and are particularly vulnerable to climate change as they lie at the lower edge of the peatland climatic envelope. The Mires-on-the-Moors project, funded by South West Water will restore over 2000 hectares of drained mire by April 2015. Herein, we question whether this restoration, which will block historical drainage ditches will allow the blanket bogs of Exmoor and Dartmoor National Parks to recover their ecohydrological functionality. We hypothesise that such mire restoration will increase the resilience of these ecosystems to climate change and will return these upland mires to peat forming/carbon sequestering systems. A method is proposed which aims to understand the processes driving gaseous carbon exchange and peat formation in an upland blanket bog and quantifies the effect restoration has on these processes. We propose to measure the spatial variation in gas fluxes with respect to structural features of the mire; drainage ditches and nanotopes. The role of vegetation; the community composition, phenology and health will be explored as well as environmental variables such as water table depths, temperature and photosynthetically active radiation. Importantly, the experiment will partition below ground respiration to assess the environmental controls and effect of restoration on autotrophic and heterotrophic respiration separately. Unusually, it will be possible to collect both pre- and post-restoration data for two experimental sites with existing intensive hydrological monitoring (baseline monitoring of water table depths at 15 minute timesteps has been in place for > 1 year at ca

  5. Composite beryllium-ceramics breeder pin elements for a gas cooled solid blanket

    International Nuclear Information System (INIS)

    Carre, F.; Chevreau, G.; Gervaise, F.; Proust, E.

    1986-06-01

    Helium coolant have main advantages compared to water for solid blankets. But limitations exist too and the development of attractive helium cooled blankets based on breeder pin assemblies has been essentially made possible by the derivation from recent CEA neutronic studies of an optimized composite beryllium/ceramics breeder arrangement. Description of the proposed toroidal blanket layout for Net is made together with the analysis of its main performance. Merits of the considered composite Be/ceramics breeder elements are discussed

  6. 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.

  7. Preliminary conceptual design of the blanket and power conversion system for the Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-01-01

    A conceptual design of a commercial Mirror Hybrid Reactor, optimized for 239 Pu production, has been completed. This design is the product of a joint effort by Lawrence Livermore Laboratory and General Atomic Company, and follows directly from earlier work on the Mirror Hybrid. This paper describes the blanket and power conversion system of the reactor design. Included are descriptions of the prestressed concrete reactor vessel that supports the magnets and contains the blanket and power conversion system components, the blanket module design, the blanket fuel design, and the power conversion system

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

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Y.; Tobita, K.; Utoh, H.; Hoshino, K.; Asakura, N.; Nakamura, M.; Tanigawa, H.; Mikio, E.; Tanigawa, H.; Nakamichi, M.; Hoshino, T., E-mail: someya.yoji@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

    2012-09-15

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

  9. Structural effects on fusion reactor blankets due to liquid metals in magnetic fields

    International Nuclear Information System (INIS)

    Lehner, J.R.; Reich, M.; Powell, J.R.

    1976-01-01

    The transient stress distribution caused in the blanket structure when the plasma current suddenly switches off in a time short compared to the L/R decay time of the liquid metal blanket was studied. Poloidal field of the plasma will induce a current to flow in the liquid metal and blanket walls. Since the resistance of the liquid lithium will be much less than that of the metal walls, the current can be considered as flowing around the blanket near the cross section perimeter, but in the lithium

  10. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H

    2001-11-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

  11. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

    2001-01-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

  12. 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.)

  13. 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.

  14. Calculations of tritium breeding ratio and inventory distributions of FEB blanket

    International Nuclear Information System (INIS)

    Deng Baiquan

    2001-01-01

    Based on the design features of FEB reactor blanket, the tritium breeding ratio and tritium concentrations in liquid lithium of each breeding zone have been calculated after 10 days full power operation for outboard blanket and one day operation for inboard blanket. The comparisons with the results calculated by Monte-Carlo code MORSE-CGT are made. Meanwhile the inventory in beryllium multiplier after one-year full power operation has also been estimated. An important conclusion has been drew the thermal hydraulic design should be careful to guarantee the blanket temperature should not rise as high as 680 degree C

  15. Japanese contribution to the design of primary module of shielding blanket in ITER-FEAT

    International Nuclear Information System (INIS)

    Kuroda, Toshimasa; Hatano, Toshihisa; Miki, Nobuharu; Hiroki, Seiji; Enoeda, Mikio; Ohmori, Junji; Akiba, Masato

    2003-02-01

    Japanese contributions to the design activity on the shielding blanket module consisting of the separable first wall and the shield block for ITER-FEAT are compiled. Temperature and stress distributions in the first wall and the shield block are analyzed and evaluated with 2-D and 3-D models for steady state and also for transient condition according to plasma ramp-up and ramp-down. While temperatures and stresses in the first wall satisfy their allowable values, those in a front part of the shield block exceed the allowable guideline. Based on this result, design improvements are suggested. Coolant flow and pressure distributions along the complicated coolant channel in the shield block are preliminary analyzed. Though heat removal is satisfactory in all coolant channels, back flows due to choking in coolant collectors are found. Design improvements to avoid the choking are suggested. Electromagnetic forces acting on blanket modules are analyzed with detailed 3-D models of solid elements for different disruption scenarios. The maximum moment around radial axis is 1.36 MNm on module no.5 under fast upward VDE, and the maximum moment around vertical axis is 1.47 MNm on module no.1 under fast downward VDE. The supporting beam of the first wall with welded attachment to the shield block is designed. Required welding thickness and support conditions to withstand electromagnetic forces are estimated. Strength of the shield block at the region mating the flexible cartridge is also estimated. Though the shield block surface attached by the flexible cartridge shows sufficient strength, the internal thread mating the Inconel bolt would need more length. In addition, water-to-water leak detection system in case main supply/return manifolds are located within the vacuum vessel is designed. By using Kr as the tracer material, the possibility of water-to-water leak detection and the concept of the detection system are shown. The design of the shielding blanket of ITER-FEAT has

  16. International

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    This rubric reports on 10 short notes about international economical facts about nuclear power: Electricite de France (EdF) and its assistance and management contracts with Eastern Europe countries (Poland, Hungary, Bulgaria); Transnuclear Inc. company (a 100% Cogema daughter company) acquired the US Vectra Technologies company; the construction of the Khumo nuclear power plant in Northern Korea plays in favour of the reconciliation between Northern and Southern Korea; the delivery of two VVER 1000 Russian reactors to China; the enforcement of the cooperation agreement between Euratom and Argentina; Japan requested for the financing of a Russian fast breeder reactor; Russia has planned to sell a floating barge-type nuclear power plant to Indonesia; the control of the Swedish reactor vessels of Sydkraft AB company committed to Tractebel (Belgium); the renewal of the nuclear cooperation agreement between Swiss and USA; the call for bids from the Turkish TEAS electric power company for the building of the Akkuyu nuclear power plant answered by three candidates: Atomic Energy of Canada Limited (AECL), Westinghouse (US) and the French-German NPI company. (J.S.)

  17. The role of the circle of Willis in internal carotid artery stenosis and anatomical variations: a computational study based on a patient-specific three-dimensional model.

    Science.gov (United States)

    Zhu, Guangyu; Yuan, Qi; Yang, Jian; Yeo, Joon Hock

    2015-11-25

    The aim of this study is to provide better insights into the cerebral perfusion patterns and collateral mechanism of the circle of Willis (CoW) under anatomical and pathological variations. In the current study, a patient-specific three-dimensional computational model of the CoW was reconstructed based on the computed tomography (CT) images. The Carreau model was applied to simulate the non-Newtonian property of blood. Flow distributions in five common anatomical variations coexisting with different degrees of stenosis in the right internal carotid artery (RICA) were investigated to obtain detailed flow information. With the development of stenosis in unilateral internal carotid artery (ICA), the cerebral blood supply decreased when the degree of stenosis increased. The blood supply of the ipsilateral middle cerebral artery (MCA) was most affected by the stenosis of ICA. The anterior communicating artery (ACoA) and ipsilateral posterior communicating artery (PCoA) functioned as the important collateral circulation channels when unilateral stenosis occurred. The blood flow of the anterior circulation and the total cerebral blood flow (CBF) reached to the minimum in the configuration of the contralateral proximal anterior cerebral artery (A1) absence coexisting with unilateral ICA stenosis. Communicating arteries provided important collateral channels in the complete CoW when stenosis in unilateral ICA occurred. The cross-flow in the ACoA is a sensitive indicator of the morphological change of the ICA. The collateral function of the PCoA on the affected side will not be fully activated until a severe stenosis occurred in unilateral ICA. The absence of unilateral A1 coexisting with the stenosis in the contralateral ICA could be the most dangerous configuration in terms of the total cerebral blood supply. The findings of this study would enhance the understanding of the collateral mechanism of the CoW under different anatomical variations.

  18. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    Veleckis, E.; Yonco, R.M.; Maroni, V.A.

    1980-01-01

    The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li 3 N, Li 2 O, and Li 2 C 2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

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

    International Nuclear Information System (INIS)

    Bohm, Tim; Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul; Ulrickson, Michael; Bullock, James

    2015-01-01

    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.

  20. 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.

  1. Modeling and experiments on tritium permeation in fusion reactor blankets

    Science.gov (United States)

    Holland, D. F.; Longhurst, G. R.

    The determination of tritium loss from helium-cooled fusion breeding blankets are discussed. The issues are: (1) applicability of present models to permeation at low tritium pressures; (2) effectiveness of oxide layers in reducing permeation; (3) effectiveness of hydrogen addition as a means to lower tritium permeation; and (4) effectiveness of conversion to tritiated water and subsequent trapping to reduce permeation. Theoretical models applicable to these issues are discussed, and results of experiments in two areas are presented; permeation of mixtures of hydrogen isotopes and conversion to tritiated water.

  2. Modeling and experiments on tritium permeation in fusion reactor blankets

    International Nuclear Information System (INIS)

    Holland, D.F.; Longhurst, G.R.

    1985-01-01

    Issues are discussed that are critical in determining tritium loss from helium-cooled fusion breeding blankets. These issues are: (a) applicability of present models to permeation at low tritium pressures, (b) effectiveness of oxide layers in reducing permeation, (c) effectiveness of hydrogen addition as a means to lower tritium permeation, and (d) effectiveness of conversion to tritiated water and subsequent trapping as a means to reduce permeation. The paper discusses theoretical models applicable to these issues, and presents results of experiments in two areas: permeation of mixtures of hydrogen isotopes and conversion to tritiated water

  3. Thermalhydraulics of flowing particle-bed-type fusion reactor blankets

    International Nuclear Information System (INIS)

    Nietert, R.E.; Abdelk-Khalik, S.I.

    1982-01-01

    An experimental investigation has been conducted to determine the heat transfer characteristics of gravity-flowing particle beds using a special heat transfer loop. Glass microspheres were allowed to flow by gravity at controlled rates through an electrically heated stainless steel tubular test section. Values of the local and average convective heat transfer coefficient as a function of the average bed velocity, particle size and heat flux were determined. Such information is necessary for the design of gravity-flowing particle-bed type fusion reactor-blankets and associated tritium recovery systems. (orig.)

  4. Reddening and blanketing of RR-Lyrae stars, ch. 3

    International Nuclear Information System (INIS)

    Lub, J.

    1977-01-01

    The effects of metal line blanketing and interstellar reddening upon the colours of the RR-Lyrae Stars are discussed. Due to the faintness of these stars in the ultraviolet W channel (at lambda 3720 A) the photometry is in most cases reduced to a four-colour VBLU photometry, i.e. there are only three colour indices available for the determination of the four quantities: interstellar reddening, effective temperature, atmospheric pressure (or effective gravity), and metal line strength which determine the energy distribution that was measured

  5. Conception of divertorless tokamak reactor with turbulent plasma blanket

    International Nuclear Information System (INIS)

    Nedospasov, A.V.; Tokar, M.Z.

    1980-01-01

    The results of the calculations presented here demonstrate that, with technically reasonable degree of the magnetic field stochastisation, the turbulent plasma blanket can take the place of a divertor. It performs the three main functions of the divertor: (a) the exhaust of the helium and unburned fuel; (b) weakening of the fast particle flux to the wall surface; and (c) essential reduction of the impurity content in the active zone of the reactor. Taking into account that plasma flows to the first wall along field lines, we may figuratively say that the first wall plays the role of a divertor in our conception. (orig.)

  6. Tritium Management In HCLL-PPCS Model AB Blanket

    International Nuclear Information System (INIS)

    Ricapito, I.; Aiello, A.; Benamati, G.; Utili, M.; Ciampichetti, A.; Zucchetti, M.

    2006-01-01

    One the main issues in the HCLL blanket development for a prototype fusion reactor is the technical feasibility of the bred tritium processing system. The basis of such concern lies in the very low tritium-Pb17Li Sieverts' constant, as measured by different scientists in the past years. In the PPCS reactor 650 g/d of tritium must be generated in the breeding blanket while less than 1 g/y of tritium has to be released to the environment through the secondary cooling circuit. As a consequence, CPS (Coolant Purification System) plays a fundamental role because it has to keep at an acceptable level the tritium partial pressure in the primary HCS (Helium Cooling Circuit) limiting, therefore, the tritium environmental release through leakage and permeation into the secondary cooling circuit. On the other hand, the He mass flow-rate to be processed by CPS is linear with the tritium permeation rate from the breeder into HCS. Therefore, with the above mentioned low Sieverts' constant values and the consequent high tritium partial pressure in the liquid metal, the possibility to keep acceptable the CPS capacity depends on a highly efficient and stable performance of tritium permeation barriers, to be applied not only on the blanket cooling plates but also on the steam generator walls. However, the experimental results on the tritium permeation barriers under relevant operative conditions were so far quite disappointing. The new data on the Sieverts' constant achieved at ENEA CR Brasimone, one order of magnitude higher than those founding the past, have a big impact in relaxing the above mentioned requirements for the tritium management in PPCS model AB reactor. Besides presenting and discussing these recent experimental results, an updated assessment of the tritium permeation rate from the liquid breeder into HCS through the cooling plates and from HCS into the environment through the steam generators is given in this paper. The consequent new constraints in terms of tritium

  7. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

    Tokuhiro, A.

    1991-11-01

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

  8. Thermal and mechanical design of WITAMIR-I blanket

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.N.

    1980-10-01

    The design philosophy of WITAMIR-I, a Wisconsin Tandem Mirror Reactor design study, uses the experience obtained from our previous tokamak studies and combines it with the unique features of the tandem mirror to obtain an attractive design of a TM power reactor. It is aimed at maximizing the strengths of the tandem mirror while mitigating its weaknesses. The end product should be a safe, reliable, maintainable and a relatively economic power reactor. The general description of the reactor, the plasma calculations, the magnet design, the neutronic calculations and the maintenance considerations are presented elsewhere. This paper presents the blanket design of this reactor study

  9. Sensitivity and uncertainty analysis of NET/ITER shielding blankets

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Gruppelaar, H.; Verschuur, K.A.

    1990-09-01

    Results are presented of sensitivity and uncertainty calculations based upon the European fusion file (EFF-1). The effect of uncertainties in Fe, Cr and Ni cross sections on the nuclear heating in the coils of a NET/ITER shielding blanket has been studied. The analysis has been performed for the total cross section as well as partial cross sections. The correct expression for the sensitivity profile was used, including the gain term. The resulting uncertainty in the nuclear heating lies between 10 and 20 per cent. (author). 18 refs.; 2 figs.; 2 tabs

  10. Evaluation of the activity levels in fusion reactor blankets

    International Nuclear Information System (INIS)

    Gruber, J.

    1977-05-01

    The activation of a fusion reactor blanket (316 SS or V-10Cr-10Ti as structure) with a minimum lithium inventory has been calculated for 0.83 MW/m 2 wall load. The resulting radiation levels and waste problems are discussed. The dose rate near the steel structure will always be higher than 0.1 rem/h due to its niobium content. After 200 to 100,000 years of decay the potential biological hazard originating from this high level fusion reactor waste (with plutonium recyclation). (orig.) [de

  11. Activation and afterheat analyses for the HCPB test blanket

    International Nuclear Information System (INIS)

    Pereslavtsev, P.; Fischer, U.

    2007-01-01

    The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in the framework of the European Fusion Technology Programme for performance tests in ITER. The recent development programme focussed on the detailed engineering design of the Test Blanket Module (TBM) and associated systems including the assessment of safety and licensing related issues with the objective to prepare for a preliminary Safety Report. To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20 degree ITER torus sector with an integrated TBM of the HCPB PI (Plant Integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a three (calendar) years period. It was simulated by a continuous irradiation for 3 years minus the last month and a discontinuous irradiation with 250 pulses (420 s pulse length, 1200 s power-off in between) over the last month. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER according to the M-DRG-1 irradiation scenario with a total first wall fluence of 0.3 MWa/m 2 . For both irradiation scenarios the radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity and afterheat of the TBM, its constituting components and materials including their

  12. Feasibility analysis of vacuum sieve tray for tritium extraction in the HCLL test blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Okino, Fumito, E-mail: fumito.okino@iae.kyoto-u.ac.jp [Kyoto University Institute of Advanced Energy, 611-0011 Gokasho, Uji, Kyoto (Japan); Calderoni, Pattrick [Fusion For Energy, 08019 Barcelona (Spain); Kasada, Ryuta; Konishi, Satoshi [Kyoto University Institute of Advanced Energy, 611-0011 Gokasho, Uji, Kyoto (Japan)

    2016-11-01

    Highlights: • The authors discovered faster mass transport on a droplet falling in a vacuum. • Primary cause of the hydrogen release from droplet is by the oscillation of a droplet. • The spherical oscillation induces the internal advection and enhances mass transfer. • This assumption agreed with previous experimental results. - Abstract: This paper describes the quantitative analysis for the design of a tritium extraction system that uses liquid PbLi droplets in vacuum (Vacuum Sieve Tray, VST), for application to the ITER helium-cooled lithium lead (HCLL) test blanket system (TBS). The parametric dependences of tritium extraction efficiency from the main geometrical features such as initial droplet velocity, nozzle head height, nozzle diameter, and flow rate are discussed. With nozzle diameters between 0.4 and 0.6 mm, extraction efficiency is estimated from 0.77 to 0.96 at the falling height of 0.5 m, with flow rate between 0.2 and 1.0 kg/s. The device has a height of 1.6 m, within the external dimensions of the HCLL Test Blanket Module (TBM), and no additional pumping power is required. The attained results are considered attractive not only for ITER, but also in view of the application of the VST concept as a candidate tritium extraction system for the European Union's demonstration fusion reactor (DEMO). The extraction efficiency of a single droplet column, which is the basis of the design analysis presented, has been validated experimentally with hydrogen. However, further experiments are required on an integrated system with size relevant to the proposed HCLL-TBS design to validate system-level effects, particularly regarding the desorption process in an array of multiple droplets.

  13. Variations in Target Volume Definition for Postoperative Radiotherapy in Stage III Non-Small-Cell Lung Cancer: Analysis of an International Contouring Study

    International Nuclear Information System (INIS)

    Spoelstra, Femke; Senan, Suresh; Le Pechoux, Cecile; Ishikura, Satoshi; Casas, Francesc; Ball, David; Price, Allan; De Ruysscher, Dirk; Soernsen de Koste, John R. van

    2010-01-01

    Purpose: Postoperative radiotherapy (PORT) in patients with completely resected non-small-cell lung cancer with mediastinal involvement is controversial because of the failure of earlier trials to demonstrate a survival benefit. Improved techniques may reduce toxicity, but the treatment fields used in routine practice have not been well studied. We studied routine target volumes used by international experts and evaluated the impact of a contouring protocol developed for a new prospective study, the Lung Adjuvant Radiotherapy Trial (Lung ART). Methods and Materials: Seventeen thoracic radiation oncologists were invited to contour their routine clinical target volumes (CTV) for 2 representative patients using a validated CD-ROM-based contouring program. Subsequently, the Lung ART study protocol was provided, and both cases were contoured again. Variations in target volumes and their dosimetric impact were analyzed. Results: Routine CTVs were received for each case from 10 clinicians, whereas six provided both routine and protocol CTVs for each case. Routine CTVs varied up to threefold between clinicians, but use of the Lung ART protocol significantly decreased variations. Routine CTVs in a postlobectomy patient resulted in V 20 values ranging from 12.7% to 54.0%, and Lung ART protocol CTVs resulted in values of 20.6% to 29.2%. Similar results were seen for other toxicity parameters and in the postpneumectomy patient. With the exception of upper paratracheal nodes, protocol contouring improved coverage of the required nodal stations. Conclusion: Even among experts, significant interclinician variations are observed in PORT fields. Inasmuch as contouring variations can confound the interpretation of PORT results, mandatory quality assurance procedures have been incorporated into the current Lung ART study.

  14. Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

    International Nuclear Information System (INIS)

    Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

    1985-01-01

    Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

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

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Ito, Akira; Taguchi, Kou; Takiguchi, Yuji; Takahashi, Hiroyuki; Tada, Eisuke

    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)

  16. Annual report of the CTR Blanket Engineering research facility in 1996

    International Nuclear Information System (INIS)

    1998-02-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1996. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Bond, A.

    1984-01-01

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

  18. 75 FR 46911 - Certain Woven Electric Blankets from the People's Republic of China: Amended Final Determination...

    Science.gov (United States)

    2010-08-04

    ... Blankets from the People's Republic of China: Amended Final Determination of Sales at Less Than Fair Value... than fair value (``LTFV'') in the antidumping investigation of certain woven electric blankets (``woven... From the People's Republic of China: Final Determination of Sales at Less Than Fair Value, 75 FR 38459...

  19. Annual report of the CTR Blanket Engineering research facility in 1992

    International Nuclear Information System (INIS)

    1993-08-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1992. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

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

    International Nuclear Information System (INIS)

    1995-09-01

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

  1. 77 FR 76013 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-12-26

    ... marketing supplies of LNG. Sempra is a customer of the Cameron Terminal. On June 22, 2012, FE issued DOE/FE... DEPARTMENT OF ENERGY [FE Docket No. 12-155-LNG] Sempra LNG Marketing, LLC; Application for Blanket..., by Sempra LNG Marketing, LLC (Sempra LNG Marketing), requesting blanket authorization to export...

  2. 78 FR 4400 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2013-01-22

    ... in the business of purchasing and marketing supplies of LNG, and is a customer of the Cameron... DEPARTMENT OF ENERGY [FE Docket No. 12-161-LNG] Eni USA Gas Marketing LLC; Application for Blanket..., by Eni USA Gas Marketing LLC (Eni USA Gas Marketing), requesting blanket authorization to export...

  3. Annual report of the CTR blanket engineering research facility in 1993

    International Nuclear Information System (INIS)

    1994-08-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1993. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  4. Heat-pipe liquid-pool-blanket concept for the Tandem Mirror Reactor

    International Nuclear Information System (INIS)

    Hoffman, M.A.; Werner, R.W.; Johnson, G.L.

    1981-01-01

    The blanket concept for the tandem mirror reactor described in this paper was developed to produce the medium temperature heat (approx. 850 to 950 K) for the General Atomic sulfur-iodine thermochemical process for producing hydrogen. This medium temperature heat from the blanket constitutes about 81% of the total power output of the fusion reactor

  5. 77 FR 31004 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-05-24

    ... 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 Docket...

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

    International Nuclear Information System (INIS)

    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

  7. Design study of blanket structure based on a water-cooled solid breeder for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Youji; Tobita, Kenji; Utoh, Hiroyasu; Tokunaga, Shinji; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Sakamoto, Yoshiteru

    2015-10-15

    Highlights: • Neutronics design of a water-cooled solid mixed breeder blanket was presented. • The blanket concept achieves a self-sufficient supply of tritium by neutronics analysis. • The overall outlet coolant temperature was 321 °C, which is in the acceptable range. - Abstract: Blanket concept with a simplified interior for mass production has been developed using a mixed bed of Li{sub 2}TiO{sub 3} and Be{sub 12}Ti pebbles, coolant conditions of 15.5 MPa and 290–325 °C and cooling pipes without any partitions. Considering the continuity with the ITER test blanket module option of Japan and the engineering feasibility in its fabrication, our design study focused on a water-cooled solid breeding blanket using the mixed pebbles bed. Herein, we propose blanket segmentation corresponding to the shape and dimension of the blanket and routing of the coolant flow. Moreover, we estimate the overall tritium breeding ratio (TBR) with a torus configuration, based on the segmentation using three-dimensional (3D) Monte Carlo N-particle calculations. As a result, the overall TBR is 1.15. Our 3D neutronics analysis for TBR ensures that the blanket concept can achieve a self-sufficient supply of tritium.

  8. Potential and problems of an aqueous lithium salt solution blanket for NET

    International Nuclear Information System (INIS)

    Kuechle, M.; Bojarsky, E.; Dorner, S.; Fischer, U.; Reimann, J.; Reiser, H.

    1987-07-01

    The report describes design studies on a water cooled in-vessel shield blanket for NET and its modification into an aqueous lithium salt blanket. The shield blankets are exchangable against breeding blankets and fulfill their shielding and heat removal functions. Emphasis is on simplicity and reliability. The water cooled shield is a large steel container in the shape of the blanket segment which is filled by water and containes a grid structure of poloidally arranged steel plates. The water flows several times in poloidal direction through the channels formed by the steel plates and is thereby heated up from 40degC to 70degC. When the water is replaced by an aqueous lithium salt solution the shield can be converted into a tritium breeding blanket without any design modification or invessel component replacement. When compared with other concepts this blanket has the advantage that the solution can replace water cooling also in the divertor and in segments dedicated to plasma heating and diagnostics, what increases the coverage considerably. Extensive three-dimensional neutronics calculations were done which, together with literature studies on candidate materials, corrosion, and tritium recovery led to a first assessment of the concept. There is an indication that no major corrosion problems are to be expected in the low temperature region envisaged. Tritium recovery capital costs were estimated to be in the 20 MECU to 50 MECU range and tritium breeding ratio is comparable to the best breeding blanket. (orig./GG) [de

  9. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Blanket certificates for unbundled sales services. 284.284 Section 284.284 Conservation of Power and Water Resources... Sales by Interstate Pipelines § 284.284 Blanket certificates for unbundled sales services. (a...

  10. Rotating liquid blanket for a toroidal fusion reator

    International Nuclear Information System (INIS)

    Moir, R.W.

    1987-01-01

    A novel blanket concept is presented for toroidal geometry in which many of the limitations imposed by a first wall are avoided by not having a first wall in the usual sense. The blanket consists of a rapidly rotating, low-vapor-pressure liquid that has a sharp boundary with the vacuum region. Nozzles inject ja continuous layer of cool liquid on the inner surface. The noncentricity of the plasma is maintained so that the plasma scrape-off region intersects the rotating liqid in a localized region. This noncentricity allows sufficient space so that the scrape-off plasma layer will not bombard the nozzles, whch penetrate through the rotating liquid. This liquid ''first wall'' is bombarded by the plasma, resulting in heat deposition, sputtering, and evaporation during the short time before the exposed liquid is covered by fresh, cool liquid from the nozzles. The advantages of this reactor concept appear to be very high wall loadings (speculated to be over 10 MW/m 2 ) and long component lifetime, both crucial economic factors. The nozzles are designed for easy replacement. The reactor's disatvantage is its enormous potential for plasma contamination by impurities. (orig.)

  11. Performance of silvered Teflon (trademark) thermal control blankets on spacecraft

    Science.gov (United States)

    Pippin, Gary; Stuckey, Wayne; Hemminger, Carol

    1993-01-01

    Silverized Teflon (Ag/FEP) is a widely used passive thermal control material for space applications. The material has a very low alpha/e ratio (less than 0.1) for low operating temperatures and is fabricated with various FEP thicknesses (as the Teflon thickness increases, the emittance increases). It is low outgassing and, because of its flexibility, can be applied around complex, curved shapes. Ag/FEP has achieved multiyear lifetimes under a variety of exposure conditions. This has been demonstrated by the Long Duration Exposure Facility (LDEF), Solar Max, Spacecraft Charging at High Altitudes (SCATHA), and other flight experiments. Ag/FEP material has been held in place on spacecraft by a variety of methods: mechanical clamping, direct adhesive bonding of tapes and sheets, and by Velcro(TM) tape adhesively bonded to back surfaces. On LDEF, for example, 5-mil blankets held by Velcro(TM) and clamping were used for thermal control over 3- by 4-ft areas on each of 17 trays. Adhesively bonded 2- and 5-mil sheets were used on other LDEF experiments, both for thermal control and as tape to hold other thermal control blankets in place. Performance data over extended time periods are available from a number of flights. The observed effects on optical properties, mechanical properties, and surface chemistry will be summarized in this paper. This leads to a discussion of performance life estimates and other design lessons for Ag/FEP thermal control material.

  12. Strategy for the development of EU Test Blanket Systems instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Calderoni, P., E-mail: Pattrick.Calderoni@f4e.europa.eu; Ricapito, I.; Poitevin, Y.

    2013-10-15

    Highlights: • We developed a strategy for the development of instrumentation for EU ITER TBSs. • TBSs instrumentation functions: safety, operation and scientific mission. • Described activities are in support of ITER design review process. -- Abstract: The instrumentation of the HCLL and HCPB Test Blanket System is fundamental in ensuring that ITER safety and operational requirements are satisfied as well as in enabling the scientific mission of the TBM program. It carries out three essential functions: (i) safety, intended as compliance with ITER requirements toward public and workers protection; (ii) system control, intended as compliance with ITER operational requirements and investment protection; and (iii) scientific mission, intended as validating technology and predictive tools for blanket concepts relevant to fusion energy systems. This paper describes the strategy for instrumentation development by providing details of the following five steps to be implemented in procured activities in the short to mid-term (3–4 years): (i) provide mapping of sensors requirements based on critical review of preliminary design data; (ii) develop functional specifications for TBS sensors based on the analysis of operative conditions in the various ITER buildings in which they are located; (iii) assess availability of commercial sensors against developed specifications; (iv) develop prototypes when no available solution is identified; and (v) perform single effect tests for the most critical solicitations and post-test examination of commercial products and prototypes. Examples of technology assessment in two technical areas are included to reinforce and complement the strategy description.

  13. 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.

  14. Radwaste management aspects of the test blanket systems in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Laan, J.G. van der, E-mail: JaapG.vanderLaan@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Canas, D. [CEA, DEN/DADN, centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Chaudhari, V. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Iseli, M. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Kawamura, Y. [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Petit, P. [European Commission, DG ENER, Brussels (Belgium); Pitcher, C.S.; Torcy, D. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Ugolini, D. [Fusion for Energy, Barcelona (Spain); Zhang, H. [China Nuclear Energy Industry Corporation, Beijing 100032 (China)

    2016-11-01

    Highlights: • Test Blanket Systems are operated in ITER to test tritium breeding technologies. • The in-vessel parts of TBS become radio-active during the ITER nuclear phase. • For each TBM campaign the TBM, its shield and the Pipe Forests are removed. • High tritium contents and novel materials are specific TBS radwaste features. • A preliminary assessment confirmed RW routing, provided its proper conditioning. - Abstract: Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment.

  15. Prevalence of angina in women versus men: a systematic review and meta-analysis of international variations across 31 countries.

    Science.gov (United States)

    Hemingway, Harry; Langenberg, Claudia; Damant, Jacqueline; Frost, Chris; Pyörälä, Kalevi; Barrett-Connor, Elizabeth

    2008-03-25

    In the absence of previous international comparisons, we sought to systematically evaluate, across time and participant age, the sex ratio in angina prevalence in countries that differ widely in the rate of mortality due to myocardial infarction. We searched MEDLINE and EMBASE until February 2006 for healthy population studies published in any language that reported the prevalence of angina (Rose questionnaire) in women and men. We obtained myocardial infarction mortality rates from the World Health Organization. A total of 74 reports of 13,331 angina cases in women and 11,511 cases in men from 31 countries were included. Angina prevalence varied widely across populations, from 0.73% to 14.4% (population weighted mean 6.7%) in women and from 0.76% to 15.1% (population weighted mean 5.7%) in men, and was strongly correlated within populations between the sexes (r=0.80, Pethnic groups than among whites. This sex ratio did not differ significantly by participant's age, the year the survey began, or the sex ratio for mortality due to myocardial infarction. Over time and at different ages, independent of diagnostic and treatment practices, women have a similar or slightly higher prevalence of angina than men across countries with widely differing myocardial infarction mortality rates.

  16. Development of filler wires for welding of reduced activation ferritic martenstic steel for India's test blanket module of ITER

    International Nuclear Information System (INIS)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K.

    2011-01-01

    Highlights: → Weld microstructure produced by RAFMS filler wires are free from delta ferrite. → Cooling rates of by weld thermal cycles influences the presence of delta ferrite. → Weld parameters modified with higher pre heat temperature and high heat input. → PWHT optimized based on correlation of hardness between base and weld metals. → Optimised mechanical properties achieved by proper tempering of the martensite. - Abstract: Indigenous development of reduced activation ferritic martensitic steel (RAFMS) has become mandatory to India to participate in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFMS is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFMS filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFMS. Purpose of this study is to develop filler wires that can be directly used for both tungsten inert gas welding (TIG) and narrow gap tungsten inert gas welding (NG-TIG), which reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, autogenous welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using TIG process at various heat inputs with a preheat temperature of 250 deg. C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimised to qualify the filler wires without the presence of delta-ferrite in

  17. Development of filler wires for welding of reduced activation ferritic martensitic steel for India's test blanket module of ITER

    International Nuclear Information System (INIS)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K.

    2010-01-01

    Indigenous development of reduced activation ferritic-martensitic (RAFM) steel has become necessary for India as a participant in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFM steel is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFM steel filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFM steel. The purpose of this study is to develop filler wires that can be directly used for both gas tungsten arc welding (GTAW) and for narrow-gap gas tungsten arc welding (NG-GTAW) that reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser-MIG welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using GTAW process at various heat inputs with a preheat temperature of 250 C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some amount of delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimized to qualify the filler wires without the presence of delta-ferrite in the weld metal and with optimized mechanical properties. Results showed that the weld metals are free from delta-ferrite. Tensile properties at ambient temperature and at 500 C are well above the specified values, and are much higher than the base metal values. Ductile Brittle Transition Temperature (DBTT) has been evaluated as -81 C based on the 68 J criteria. The present study highlights the basis and methodology

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

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi; Tesini, Alessandro

    2008-01-01

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

  19. Evaluation of potential blanket concepts for a Demonstration Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Chapin, D.L.; Chi, J.W.H.; Kelly, J.L.

    1978-01-01

    An evaluation has been made of several different blanket concepts for use in a near-term Demonstration Tokamak Hybrid Reactor (DTHR), whose main objective would be to produce a significant amount of fissile fuel while demonstrating the feasibility of the tokamak hybrid reactor concept. The desirability of a simple design using proven technology plus a proliferation resistant fuel cycle led to the selection of a low temperature and pressure water-cooled, zircaloy clad ThO 2 blanket concept to breed 233 U. The nuclear performance and thermal-hydraulics characteristics of the blanket were evaluated to arrive at a consistent design. The blanket was found to be feasible for producing a significant amount of fissile fuel even with the limited operating conditions and blanket coverage in the DTHR

  20. Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie

    1998-01-01

    Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

  1. The impact of tritium solubility and diffusivity on inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Caorlin, M.; Gervasini, G.; Reiter, F.

    1988-01-01

    The authors reviewed hydrogen solubility and diffusivity data for liquid lithium-based compounds which are potential breeding blanket materials in NET-type fusion devices. These data have been used to assess tritium permeation and inventory in separately cooled NET blankets and in self cooled blankets with a vanadium first wall. The results for the separately cooled NET-liquid breeder show that tritium permeation is negligible for lithium, a serious problem for Pb-17Li and a critical one for Flibe. The total tritium inventory is lowest in lithium, high in Pb-17Li and very high in Flibe. The high tritium partial pressure for Flibe or Pb-17Li can be reduced in a self cooled blanket with a vanadium first wall. Permeation into the plasma reduces the blanket tritium inventory and permeation. Tritium recovery can be combined with the plasma exhaust

  2. Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

    International Nuclear Information System (INIS)

    Swami, H.L.; Danani, C.

    2015-01-01

    In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

  3. Phylogeny of Alternaria fungi known to produce host-specific toxins on the basis of variation in internal transcribed spacers of ribosomal DNA.

    Science.gov (United States)

    Kusaba, M; Tsuge, T

    1995-10-01

    The internal transcribed spacer regions (ITS1 and ITS2) of ribosomal DNA from Alternaria species, including seven fungi known to produce host-specific toxins, were analyzed by polymerase chain reaction-amplification and direct sequencing. Phylogenetic analysis of the sequence data by the Neighbor-joining method showed that the seven toxin-producing fungi belong to a monophyletic group together with A. alternata. In contract, A. dianthi, A. panax, A. dauci, A. bataticola, A. porri, A. sesami and A. solani, species that can be morphologically distinguished from A. alternata, could be clearly separated from A. alternata by phylogenetic of the ITS variation. These results suggest that Alternaria pathogens which produce host-specific toxins are pathogenic variants within a single variable species, A. alternata.

  4. Update on Pneumocystis carinii f. sp. hominis Typing Based on Nucleotide Sequence Variations in Internal Transcribed Spacer Regions of rRNA Genes

    Science.gov (United States)

    Lee, Chao-Hung; Helweg-Larsen, Jannik; Tang, Xing; Jin, Shaoling; Li, Baozheng; Bartlett, Marilyn S.; Lu, Jang-Jih; Lundgren, Bettina; Lundgren, Jens D.; Olsson, Mats; Lucas, Sebastian B.; Roux, Patricia; Cargnel, Antonietta; Atzori, Chiara; Matos, Olga; Smith, James W.

    1998-01-01

    Pneumocystis carinii f. sp. hominis isolates from 207 clinical specimens from nine countries were typed based on nucleotide sequence variations in the internal transcribed spacer regions I and II (ITS1 and ITS2, respectively) of rRNA genes. The number of ITS1 nucleotides has been revised from the previously reported 157 bp to 161 bp. Likewise, the number of ITS2 nucleotides has been changed from 177 to 192 bp. The number of ITS1 sequence types has increased from 2 to 15, and that of ITS2 has increased from 3 to 14. The 15 ITS1 sequence types are designated types A through O, and the 14 ITS2 types are named types a through n. A total of 59 types of P. carinii f. sp. hominis were found in this study. PMID:9508304

  5. Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

    Science.gov (United States)

    Xiaokang, ZHANG; Songlin, LIU; Xia, LI; Qingjun, ZHU; Jia, LI

    2017-11-01

    The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR). Some updating of neutronics analyses was needed, because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket, including the optimization of radial build-up and customized structure for each blanket module. A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses. The tritium breeding capability, nuclear heating power, radiation damage, and decay heat were calculated by the MCNP and FISPACT code. The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency. The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW. The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60, respectively. The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module #3. The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time. The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.

  6. Socio-economic variation in price minimizing behaviors: findings from the International Tobacco Control (ITC) Four Country Survey.

    Science.gov (United States)

    Licht, Andrea S; Hyland, Andrew J; O'Connor, Richard J; Chaloupka, Frank J; Borland, Ron; Fong, Geoffrey T; Nargis, Nigar; Cummings, K Michael

    2011-01-01

    This paper examines how socio-economic status (SES) modifies how smokers adjust to changes in the price of tobacco products through utilization of multiple price minimizing techniques. Data come from the International Tobacco Control Policy Evaluation (ITC) Four Country Survey, nationally representative samples of adult smokers and includes respondents from Canada, the United States, the United Kingdom and Australia. Cross-sectional analyses were completed among 8,243 respondents (7,038 current smokers) from the survey wave conducted between October 2006 and February 2007. Analyses examined predictors of purchasing from low/untaxed sources, using discount cigarettes or roll-your-own (RYO) tobacco, purchasing cigarettes in cartons, and engaging in high levels of price and tax avoidance at last purchase. All analyses tested for interactions with SES and were weighted to account for changing and under-represented demographics. Relatively high levels of price and tax avoidance behaviors were present; 8% reported buying from low or untaxed source; 36% used discount or generic brands, 13.5% used RYO tobacco, 29% reported purchasing cartons, and 63% reported using at least one of these high price avoidance behaviors. Respondents categorized as having low SES were approximately 26% less likely to report using low or untaxed sources and 43% less likely to purchase tobacco by the carton. However, respondents with low SES were 85% more likely to report using discount brands/RYO compared to participants with higher SES. Overall, lower SES smokers were 25% more likely to engage in at least one or more tax avoidance behaviors compared to their higher SES counterparts. Price and tax avoidance behaviors are relatively common among smokers of all SES strata, but strategies differed with higher SES groups more likely to report traveling to a low-tax location to avoid paying higher prices, purchase duty free tobacco, and purchase by cartons instead of packs all of which were less

  7. Socio-Economic Variation in Price Minimizing Behaviors: Findings from the International Tobacco Control (ITC Four Country Survey

    Directory of Open Access Journals (Sweden)

    Nigar Nargis

    2011-01-01

    Full Text Available This paper examines how socio-economic status (SES modifies how smokers adjust to changes in the price of tobacco products through utilization of multiple price minimizing techniques. Data come from the International Tobacco Control Policy Evaluation (ITC Four Country Survey, nationally representative samples of adult smokers and includes respondents from Canada, the United States, the United Kingdom and Australia. Cross-sectional analyses were completed among 8,243 respondents (7,038 current smokers from the survey wave conducted between October 2006 and February 2007. Analyses examined predictors of purchasing from low/untaxed sources, using discount cigarettes or roll-your-own (RYO tobacco, purchasing cigarettes in cartons, and engaging in high levels of price and tax avoidance at last purchase. All analyses tested for interactions with SES and were weighted to account for changing and under-represented demographics. Relatively high levels of price and tax avoidance behaviors were present; 8% reported buying from low or untaxed source; 36% used discount or generic brands, 13.5% used RYO tobacco, 29% reported purchasing cartons, and 63% reported using at least one of these high price avoidance behaviors. Respondents categorized as having low SES were approximately 26% less likely to report using low or untaxed sources and 43% less likely to purchase tobacco by the carton. However, respondents with low SES were 85% more likely to report using discount brands/RYO compared to participants with higher SES. Overall, lower SES smokers were 25% more likely to engage in at least one or more tax avoidance behaviors compared to their higher SES counterparts. Price and tax avoidance behaviors are relatively common among smokers of all SES strata, but strategies differed with higher SES groups more likely to report traveling to a low-tax location to avoid paying higher prices, purchase duty free tobacco, and purchase by cartons instead of packs all of

  8. Socio-Economic Variation in Price Minimizing Behaviors: Findings from the International Tobacco Control (ITC) Four Country Survey

    Science.gov (United States)

    Licht, Andrea S.; Hyland, Andrew J.; O’Connor, Richard J.; Chaloupka, Frank J.; Borland, Ron; Fong, Geoffrey T.; Nargis, Nigar; Cummings, K. Michael

    2011-01-01

    This paper examines how socio-economic status (SES) modifies how smokers adjust to changes in the price of tobacco products through utilization of multiple price minimizing techniques. Data come from the International Tobacco Control Policy Evaluation (ITC) Four Country Survey, nationally representative samples of adult smokers and includes respondents from Canada, the United States, the United Kingdom and Australia. Cross-sectional analyses were completed among 8,243 respondents (7,038 current smokers) from the survey wave conducted between October 2006 and February 2007. Analyses examined predictors of purchasing from low/untaxed sources, using discount cigarettes or roll-your-own (RYO) tobacco, purchasing cigarettes in cartons, and engaging in high levels of price and tax avoidance at last purchase. All analyses tested for interactions with SES and were weighted to account for changing and under-represented demographics. Relatively high levels of price and tax avoidance behaviors were present; 8% reported buying from low or untaxed source; 36% used discount or generic brands, 13.5% used RYO tobacco, 29% reported purchasing cartons, and 63% reported using at least one of these high price avoidance behaviors. Respondents categorized as having low SES were approximately 26% less likely to report using low or untaxed sources and 43% less likely to purchase tobacco by the carton. However, respondents with low SES were 85% more likely to report using discount brands/RYO compared to participants with higher SES. Overall, lower SES smokers were 25% more likely to engage in at least one or more tax avoidance behaviors compared to their higher SES counterparts. Price and tax avoidance behaviors are relatively common among smokers of all SES strata, but strategies differed with higher SES groups more likely to report traveling to a low-tax location to avoid paying higher prices, purchase duty free tobacco, and purchase by cartons instead of packs all of which were less

  9. Cross-section uncertainty study of the NET shielding blanket

    International Nuclear Information System (INIS)

    Jaeger, J.F.

    1990-11-01

    The Next European Torus (NET) is foreseen as the next step in the European development towards the controlled use of thermonuclear fusion. Detail design of the shielding blanket protecting the peripherals, more especially the super-conducting coils, is well advanced. A cross-section uncertainty, i.e. a study of the expected inaccuracy due to the nuclear cross-section data, has been done for the neutron-gamma reactions in the insulation of the coils for such a design. As an extension of previous work on the NET shielding blanket (e.g. MCNP calculations), it was deemed necessary to estimate the accuracy attainable with transport codes in view of the uncertainties in microscopic cross-sections. The code used, SENSIBL, is based on perturbation theory and uses covariance files, COVFILS-2, for the cross-section data. This necessitates forward and adjoint flux calculations with a transport code (e.g. ONEDANT, TRISM) and folding the information contained in these coupled fluxes with the accuracy estimates of the evaluators of the ENDF/B-V files. Transport, P 5 S 12 , calculations were done with the ONEDANT code, for a shielding blanket design with 714 MW plasma fusion power. Several runs were done to obtain well converged forward and adjoint fluxes (ca. 1%). The forward and adjoint integral responses agree to 2%, which is consistent with the above accuracy. The n-γ response was chosen as it is typical of the general accuracy and is available for all materials considered. The present version of SENSIBL allows direct use of the geometric files of ONEDANT (or TRISM) which simplifies the input. Covariance data is not available at present in COVFILS-2 for all of the materials considered. Only H, C, N, O, Al, Si, Fe, Ni, and Pb could be considered, the big absentee being copper. The resulting uncertainty for the neutron-gamma reactions in the insulation of the coil was found to be 17%. Simulating copper by aluminium produces a negligible increase in the uncertainty, mainly

  10. Helium Loop for the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

    2006-01-01

    In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

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

    International Nuclear Information System (INIS)

    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. (author)

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

    International Nuclear Information System (INIS)

    Lee, Jeong-Hun; Kim, Geon-Woo; Park, Goon-Cherl; Cho, Hyoung-Kyu; Im, Kihak

    2014-01-01

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

  13. The lithium blanket program at the LOTUS facility

    International Nuclear Information System (INIS)

    File, J.; Haldy, P.A.; Quanci, J.

    1987-01-01

    An experimental program of neutron transport studies of the lithium Blanket Module (LBM) carried out with the LOTUS point-neutron source at the Ecole Polytechnique Federale de Lausanne (EPTL), Switzerland has been concluded. The major objectives of this program are to perform a series of neutron transport and tritium breeding experiments to qualify the LBM for future experiments on toroidal fusion devices such as TFTR to perform neutron multiplier experiments on the LBM employing various materials in a removable slab geometry; and, to compare the experimental results of radiation dosimetry and tritium breeding with the calculations of two and three dimensional neutron transport codes. An overview of the results from the radiation dosimetry and tritium assay are presented and compared to the two and three dimensional neutron transport codes

  14. 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.

  15. Development of packagings for 'MONJU' blanket fuel assemblies

    International Nuclear Information System (INIS)

    Shibata, Kan; Ouchi, Yuichiro; Matsuzaki, Masaaki; Okuda, Yoshihisa

    1995-01-01

    Blanket assemblies for prototype Fast Breeder Reactor 'MONJU' are made at commercial fuel fabrication plants capable of handling deplete Uranium in Japan. For the purpose of transport the assemblies are inserted into a packaging that is set horizontally at the fabrication plants because of compatibility with equipment installed at the plants. On the other hand, the assemblies must be taken out from the packaging set vertically at 'MONJU' due to compatibility. For this reason development of a new packaging, which makes it possible to take assemblies in and out both horizontally and vertically, is needed to carry out transport of assemblies for reload. The development and fabrication of the packagings, taking about two years, were completed in March 1995. The packagings were used in transport of assemblies in June 1995 for the first change. This report introduces the outline of the packaging and confirmation tests done in the process of development. (author)

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Li Chunjing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China)], E-mail: lcj@ipp.ac.cn; Huang Qunying; Wu Qingsheng; Liu Shaojun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Lei Yucheng [Jiangsu University, Zhenjiang, Jiangsu, 212013 (China); Muroga, Takeo; Nagasaka, Takuya [National Institute for Fusion Science, Toki, Jifu, 509-5292 (Japan); Zhang Jianxun [Xi' an Jiaotong University, Xi' an, Shanxi, 710049 (China); Li Jinglong [Northwestern Polytechnical University, Xi' an, Shanxi, 710072 (China)

    2009-06-15

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

  18. 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.

  19. Possible association of mucous blanket integrity with postirradiation colonization resistance

    International Nuclear Information System (INIS)

    Walker, R.I.; Brook, I.; Costerton, J.W.; MacVittie, T.; Myhal, M.L.

    1985-01-01

    Radiation-induced infections can be associated with changes in colonization potential of the intestine. Since the mucous blanket, which overlays the epithelium, is a major mucosal structure and is heavily colonized by microorganisms, we examined the status of the mucus after radiation and evaluated susceptibility to intestinal challenge with bacteria. A downward shift (2.5 X 10(8) cells/g to 5.3 X 10(5)) of total facultatively anaerobic bacteria of the ileum of C3HeB/FeJ mice was detected by 3 days post exposure to 10 Gy 60Co. Numbers of flora returned to normal by 11 days after radiation. Scanning electron microscopy was used to show that the loss of bacteria could be associated with major disruptions of the continuity of the mucous blanket. The pathogen Pseudomonas aeruginosa adhered to mouse mucous films used in in vitro assays. When irradiated mice were challenged orally with 1 X 10(5) P. aeruginosa on days 1, 2, or 3 after irradiation, a progressive increase in susceptibility was seen, but no animals died before Day 4 postirradiation. Sensitivity to subcutaneous (sc) challenge with Pseudomonas also increased by Day 3 and was probably due largely to the profound neutropenia observed. Immunoglobulin G (Gamimmune), which protected burned mice infected with Pseudomonas, was ineffectual in treatment of 7 or 10 Gy irradiated mice challenged either orally or sc with the organism. The ileal mucosal barrier was compromised after radiation in ways which could facilitate epithelial colonization, an event which combined with other immunological and physiological decrements in this model can compromise the effectiveness of therapeutic modalities

  20. Vibration damage testing of thermal barrier fibrous blanket insulation

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.

    1984-01-01

    GA Technologies is engaged in a long-term, multiphase program to determine the vibration characteristics of thermal barrier components leading to qualification of assemblies for High Temperature Gas-Cooled Reactor (HTGR) service. The phase of primary emphasis described herein is the third in a series of acoustic tests and uses as background the more elemental tests preceding it. Two sizes of thermal barrier coverplates with one fibrous blanket insulation type were tested in an acoustic environment at sound pressure levels up to 160 dB. Three tests were conducted using sinusoidal and random noise for up to 200 h duration at room temperature. Frequent inspections were made to determine the progression of degradation using definition of stages from a prior test program. Initially the insulation surface adjacent to the metallic seal sheets (noise side) assumed a chafed or polished appearance. This was followed by flattening of the as-received pillowed surface. This stage was followed by a depression being formed in the vicinity of the free edge of the coverplate. Next, loose powder from within the blanket and from fiber erosion accumulated in the depression. Prior experience showed that the next stage of deterioration exhibited a consolidation of the powder to form a local crust. In this test series, this last stage generally failed to materialize. Instead, surface holes generated by solid ceramic particulates (commonly referred to as 'shot') constituted the stage following powdering. With the exception of some manufacturing-induced anomalies, the final stage, namely, gross fiber breakup, did not occur. It is this last stage that must be prevented for the thermal barrier to maintain its integrity. (orig./GL)

  1. Assessing impact of blanket interventions for MAM prevention

    International Nuclear Information System (INIS)

    Grais, Rebecca F.; Isanaka, I; Langendorf, C; Roederer, T

    2014-01-01

    Full text: Blanket interventions for MAM prevention (Blanket supplementary feeding programming (BSFP)) provide a supplementary food ration often accompanied by a basic medical treatment and prevention package to a vulnerable population for a defined period in a defined geographic location. There is little strong evidence on the impact of BSFP on rates of malnutrition and mortality, and scare guidance on program monitoring and evaluation to improve the implementation of specific programs. Assessing the impact of BSFP has been fraught with difficulty. Their isolated impact is difficult, if not often impossible to disentangle from larger care and prevention packages, the objectives of BSFP may vary by context, implementing agency, time and geography. Various and often multiple co-morbidities among children in the targeted group complicate matters further with respect to impact assessment. This leads to difficulties in generalizing results from one context to another and the need for more complex metrics to guide operational decision-making. Ideally, impact or effectiveness of BSFP should be addressed in a research framework where appropriate and complete data is collected in order to address specific questions. The gold standard is the conduct of randomized studies including a control group. These studies have been scarce as they may be perceived as either rarely feasible or not ethical or both. However, as generating evidence on impact of BSFP is essential to provide operational guidance, these studies should be encouraged through a diversity of robust, yet creative and pragmatic, methodological approaches. As a case study, a series of studies conducted over the past decade are reviewed in the same location in Niger highlighting the lessons learned. (author)

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

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Mori, Takamasa; Kosako, Kazuaki; Oyama, Yukio; Nakamura, Tomoo

    1989-10-01

    Phase IIa and IIb experiments of JAERI/U.S. Collaborative Program on Fusion Blanket Neutronics have been performed using the FNS facility at JAERI. The phase IIa experimental systems consist of the Li 2 O test region, the rotating neutron target and the Li 2 CO 3 container. In phase IIb, a beryllium layer is added to the inner wall to investigate a multiplier effect. Measured parameters are source characteristics by a foil activation method and spectrum measurements using both NE-213 and proton recoil counters. The measurements inside the Li 2 O region included tritium production rates, reaction rate by foil activation and neutron spectrum measurements. Analysis for these parameters was performed by using two dimensional discrete ordinate codes DOT3.5 and DOT-DD, and a Monte Carlo code MORSE-DD. The nuclear data used were based on JENDL3/PR1 and PR2. ENDF/B-IV, V and the FNS file were used as activation cross sections. The configurations analysed for the test region were a reference, a beryllium front and a beryllium sandwiched systems in phase IIa, and a reference and a beryllium front with first wall systems in phase IIb. This document describes the results of analysis and comparison between the calculations and the measurements. The prediction accuracy of key parameters in a fusion reactor blanket are examined. The tritium production rates can be well predicted in the reference systems but are fairly underestimated in the system with a beryllium multiplier. Details of experiments and the experimental techniques are described separately in the another report. (author)

  3. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    Sato, Satoshi; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

    2003-01-01

    In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li 2 TiO 3 blocks with a 6 Li enrichment of 40 and 95%, and beryllium blocks. Sample pellets of Li 2 TiO 3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test speciments simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steels F82H were irradiated as typical fusion materials. The effective cross-sections for incident neutron flux to calculate the radioactive nuclei ( 56 Co, 184 Re, 48 V, 206 Bi, 65 Zn and 51 Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections remarkably increases with coming closer to polyethylene board that was a substitute of water. As a result of the present study, it has become clear that the sequential reaction rates are important factors to accurately evaluate the induced activity in fusion reactors design. (author)

  4. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    Sato, S.; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

    2003-01-01

    In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li 2 TiO 3 blocks with a 6 Li enrichment of 40 and 95%, and beryllium blocks. Sample pellets of Li 2 TiO 3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross- sections for incident neutron flux to calculate the radioactive nuclei ( 56 Co, 184 Re, 48 V, 206 Bi, 65 Zn and 51 Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections remarkably increases with coming closer to polyethylene board that was a substitute of water. As a result of the present study, it has become clear that the sequential reaction rates are important factors to accurately evaluate the induced activity in fusion reactors design. (author)

  5. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1984-08-01

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

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

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

  7. Peningkatan mutu blanket karet alam melalui proses predrying dan penyemprotan asap cair

    Directory of Open Access Journals (Sweden)

    Afrizal Vachlepi

    2017-06-01

    Full Text Available Most of Indonesian rubber products SIR 20 are made from the material of raw rubber obtained from smallholders. However, the quality of this material is not good enough. Thus, quality improvement has to be carried out by manufacturers. The liquid smoke used during the blanket hanging process can improve the quality of the rubber products SIR 20. This research aimed to determine and study the effects of liquid smoke spraying and blanket hanging duration on the drying factor, the dry rubber content, technical quality, vulcanization characteristics, and physical properties of vulcanized natural rubber. Treatments consisted of various hanging duration (6, 8, and 10 days, and without hanging and spraying (with and without spraying of liquid smoke. The results showed that the spraying of liquid smoke on natural rubber blankets could improve the technical quality of the natural rubber, especially the values of Po and PRI. The spraying of liquid smoke could reduce the blanket hanging duration to 6-8 days. The blankets sprayed with liquid smoke had the optimum cure time of around 15 minutes and 19 seconds and the scorch time of around 3 minutes and 22 seconds. These values indicated that the vulcanization characteristics of blankets which were sprayed with liquid smoke were generally better than those of blankets which were not sprayed with liquid smoke

  8. Progress in blanket designs using SiCf/SiC composites

    International Nuclear Information System (INIS)

    Giancarli, L.; Golfier, H.; Nishio, S.; Raffray, R.; Wong, C.; Yamada, R.

    2002-01-01

    This paper summarizes the most recent design activities concerning the use of SiC f /SiC composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the TAURO blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium-lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and ARIES-AT blankets are essentially formed by a SiC f /SiC box acting as a container for the lithium-lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. The DREAM blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li 2 O (or other lithium ceramics) as breeder material and of SiC as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R and D on SiC f /SiC

  9. 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.

  10. Joint Markov Blankets in Feature Sets Extracted from Wavelet Packet Decompositions

    Directory of Open Access Journals (Sweden)

    Gert Van Dijck

    2011-07-01

    Full Text Available Since two decades, wavelet packet decompositions have been shown effective as a generic approach to feature extraction from time series and images for the prediction of a target variable. Redundancies exist between the wavelet coefficients and between the energy features that are derived from the wavelet coefficients. We assess these redundancies in wavelet packet decompositions by means of the Markov blanket filtering theory. We introduce the concept of joint Markov blankets. It is shown that joint Markov blankets are a natural extension of Markov blankets, which are defined for single features, to a set of features. We show that these joint Markov blankets exist in feature sets consisting of the wavelet coefficients. Furthermore, we prove that wavelet energy features from the highest frequency resolution level form a joint Markov blanket for all other wavelet energy features. The joint Markov blanket theory indicates that one can expect an increase of classification accuracy with the increase of the frequency resolution level of the energy features.

  11. Advanced methods comparisons of reaction rates in the Purdue Fast Breeder Blanket Facility

    International Nuclear Information System (INIS)

    Hill, R.N.; Ott, K.O.

    1988-01-01

    A review of worldwide results revealed that reaction rates in the blanket region are generally underpredicted with the discrepancy increasing with penetration; however, these results vary widely. Experiments in the large uniform Purdue Fast Breeder Blanket Facility (FBBF) blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50 group cross sections), a consistent Calculated/Experimental (C/E) drop-off was observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the blanket is necessary for agreement with experiments. The usefulness of refined group constant generation utilizing specialized weighting spectra and transport theory methods in correcting this discrepancy was analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The surprising result was that transport methods had no effect on the blanket deviations; thus, transport theory considerations do not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by some approximations which are applied in all current methods. 27 refs., 3 figs., 1 tab

  12. Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Mistrangelo, C., E-mail: chiara.mistrangelo@kit.edu; Bühler, L.

    2014-10-15

    Highlights: • MHD flows in model geometries of DCLL blanket manifolds. • Study of velocity, pressure distributions and flow partitioning in parallel ducts. • Flow partitioning affected by 3D MHD pressure drop and velocity distribution in the expanding zone. • Reduced pressure drop in a continuous expansion compared to a sudden expansion. - Abstract: An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure. Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules. In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.

  13. Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

    2016-11-01

    Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

  14. A coupled systems code-CFD MHD solver for fusion blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Wolfendale, Michael J., E-mail: m.wolfendale11@imperial.ac.uk; Bluck, Michael J.

    2015-10-15

    Highlights: • A coupled systems code-CFD MHD solver for fusion blanket applications is proposed. • Development of a thermal hydraulic systems code with MHD capabilities is detailed. • A code coupling methodology based on the use of TCP socket communications is detailed. • Validation cases are briefly discussed for the systems code and coupled solver. - Abstract: The network of flow channels in a fusion blanket can be modelled using a 1D thermal hydraulic systems code. For more complex components such as junctions and manifolds, the simplifications employed in such codes can become invalid, requiring more detailed analyses. For magnetic confinement reactor blanket designs using a conducting fluid as coolant/breeder, the difficulties in flow modelling are particularly severe due to MHD effects. Blanket analysis is an ideal candidate for the application of a code coupling methodology, with a thermal hydraulic systems code modelling portions of the blanket amenable to 1D analysis, and CFD providing detail where necessary. A systems code, MHD-SYS, has been developed and validated against existing analyses. The code shows good agreement in the prediction of MHD pressure loss and the temperature profile in the fluid and wall regions of the blanket breeding zone. MHD-SYS has been coupled to an MHD solver developed in OpenFOAM and the coupled solver validated for test geometries in preparation for modelling blanket systems.

  15. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    Directory of Open Access Journals (Sweden)

    Kooyman Timothée

    2017-01-01

    Full Text Available Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long- and short-term neutron and gamma source is carried out whereas in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

  16. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    Science.gov (United States)

    Kooymana, Timothée; Buiron, Laurent; Rimpault, Gérald

    2017-09-01

    Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long and short term neutron and gamma source is carried out while in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

  17. Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

    International Nuclear Information System (INIS)

    Kooyman, T.; Buiron, L.; Rimpault, G.

    2017-01-01

    Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long- and short-term neutron and gamma source is carried out whereas in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing. (authors)

  18. Analysis of the separation of protium from blanket tritium-product streams

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1981-07-01

    The case is considered in which the blanket product stream has been purified to the point where only protium, tritium, and a small quantity of deuterium remain. A cryogenic distillation cascade concept developed specifically to handle this enrichment problem is shown. The concept is based on a series of distillation columns and equilibrators capable of producing a protium-rich stream containing less than 1000 appm T and a tritium-rich stream containing less than 2000 appm H. It is envisioned that both of these streams could be blended with streams of comparable composition in the mainstream position of the fuel cycle without further processing. The computational analysis of the cascade was based on a fixed arrangement of columns and equilibrators and a fixed number of theoretical plates per columns, since these features are less easily varied in an actual system than reflux ratios and flow rates. In order to test the flexibility of this conceptual enruchment system to adjust to variations of the H/T ratio in the feed, H/T values of 0.333, 1.00, and 3.00 were investigated

  19. Study of MHD problems in liquid metal blankets of fusion reactors

    International Nuclear Information System (INIS)

    Michael, I.

    1984-12-01

    This study describes in a concise form the state of knowledge regarding MHD problems to be expected in case of use of liquid metal in the blankets of fusion reactors with magnetic confinement. MHD pressure losses and MHD friction coefficients in the straight channel, in bent sections and in case of variation of the channel cross section play a major role because the high MHD flow resistances call for high pumping powers. Influencing the velocity profile transverse to the main flow direction of the liquid metal by application of an external, strong magnetic field bears consequences on the release and transport of corrosion products in the liquid metal circuit and on the heat transfer. Possibilities of reducing the MHD effects are discussed. However, it becomes obvious that an account of the lack of experimental results there are still major gaps in the knowledge of MHD effects occurring in strong magnetic fields. These gaps can be greatly reduced by implementation of an experimental program as proposed in this report. (orig.) [de

  20. Feasibility study of incore fission chamber application for neutron flux measurements on the NET blanket

    International Nuclear Information System (INIS)

    Bertalot, L.

    1987-01-01

    A feasibility study has been carried out on the use of in-core fission chambers as neutron diagnostic tools to perform neutron flux measurements on the blanket component of NET. The high neutron and gamma fluxes and the severe thermal-mechanical and magnetic conditions of the blanket structure have been taken into account in this analysis. Preliminary design criteria and specifications of an in-core detector are presented for NET application. A research and development programme is outlined which aims to obtain more information on the tecnological constraints arising from the severe conditions of the NET blanket