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Sample records for iter blanket maintenance

  1. ITER breeding blanket design

    SciTech Connect

    Gohar, Y.; Cardella, A.; Ioki, K.; Lousteau, D.; Mohri, K.; Raffray, R.; Zolti, E.

    1995-12-31

    A breeding blanket design has been developed for ITER to provide the necessary tritium fuel to achieve the technical objectives of the Enhanced Performance Phase. It uses a ceramic breeder and water coolant for compatibility with the ITER machine design of the Basic Performance Phase. Lithium zirconate and lithium oxide am the selected ceramic breeders based on the current data base. Enriched lithium and beryllium neutron multiplier are used for both breeders. Both forms of beryllium material, blocks and pebbles are used at different blanket locations based on thermo-mechanical considerations and beryllium thickness requirements. Type 316LN austenitic steel is used as structural material similar to the shielding blanket. Design issues and required R&D data are identified during the development of the design.

  2. ITER shielding blanket

    SciTech Connect

    Strebkov, Y.; Blinov, Y.; Avsjannikov, A.

    1994-12-31

    A version of ITER shielding blanket design is presented. Main features of this proposal are: Cu based alloy as structure material of the first wall - integrated in the blanket segment box structure and 316L SS as material of both back/side walls of the box and the shield structure elements; water of medium pressure (up to 4 MPa) as coolant with toroidal direction of flow; two variants of beryllium protection tiles joining (either permanent joints by mince of solid diffusion bonding or demountable attachment with compliant layer; for last versions designs options of tiles attachment units are given). Problems of manufacturing of such blanket segment including its assembly sequence are considered in details. Results of stress problem analysis for thermal, pressure, and electromagnetic loads will be given in this report also.

  3. US solid breeder blanket design for ITER

    SciTech Connect

    Gohar, Y.; Attaya, H.; Billone, M.; Lin, C.; Johnson, C.; Majumdar, S.; Smith, D. ); Goranson, P.; Nelson, B.; Williamson, D.; Baker, C. ); Raffray, A.; Badawi, A.; Gorbis, Z.; Ying, A.; Abdou, M. ); Sviatoslavsky, I.; Blanchard, J.; Mogahed, E.; Sawan, M.; Kulcinski, G. )

    1990-09-01

    The US blanket design activity has focused on the developments and the analyses of a solid breeder blanket concept for ITER. The main function of this blanket is to produce the necessary tritium required for the ITER operation and the test program. Safety, power reactor relevance, low tritium inventory, and design flexibility are the main reasons for the blanket selection. The blanket is designed to operate satisfactorily in the physics and the technology phases of ITER without the need for hardware changes. Mechanical simplicity, predictability, performance, minimum cost, and minimum R D requirements are the other criteria used to guide the design process. The design aspects of the blanket are summarized in this paper. 2 refs., 7 figs., 3 tabs.

  4. Thermomechanical analysis of the ITER breeding blanket

    SciTech Connect

    Majumdar, S.; Gruhn, H.; Gohar, Y.; Giegerich, M.

    1997-03-01

    Thermomechanical performance of the ITER breeding blanket is an important design issue because it requires first, that the thermal expansion mismatch between the blanket structure and the blankets internals (such as, beryllium multiplier and tritium breeders) can be accommodated without creating high stresses, and second, that the thermomechanical deformation of various interfaces within the blanket does not create high resistance to heat flow and consequent unacceptably high temperatures in the blanket materials. Thermomechanical analysis of a single beryllium block sandwiched between two stainless steel plates was carried out using the finite element code ABAQUS to illustrate the importance of elastic deformation on the temperature distributions. Such an analysis for the whole ITER blanket needs to be conducted in the future. Uncertainties in the thermomechanical contact analysis can be reduced by bonding the beryllium blocks to the stainless steel plates by a thin soft interfacial layer.

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

    NASA Astrophysics Data System (ADS)

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

    SciTech Connect

    L. C. Cadwallader; C. P. C. Wong; M. Abdou; B. B. Morely; B.J Merrill

    2014-10-01

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

  7. Solid breeder blanket option for the ITER conceptual design

    SciTech Connect

    Gohar, Y.; Attaya, H.; Billone, M.C.; Finn, P.; Majumdar, S.; Turner, L.R.; Baker, C.C.; Nelson, B.E.; Raffray, R.; Oak Ridge National Lab., TN; California Univ., Los Angeles, CA )

    1989-10-01

    A solid-breeder water-cooled blanket option was developed for ITER based on a multilayer configuration. The blanket uses beryllium for neutron multiplication and lithium oxide for tritium breeding. The material forms are sintered products for both material with 0.8 density factor. The lithium-6 enrichment is 90%. This blanket has the capability to accommodate a factor of two change in the neutron wall loading without violating the different design guidelines. 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. At the same time, the reliability and the safety aspects of the blanket are enhanced by the use of a low-pressure coolant and the separation of the tritium purge lines from the coolant system. The blanket modules are made by hot vacuum forming and diffusion bonding a double wall structure with integral cooling channels. The different aspects of the blanket design including tritium breeding, nuclear heat deposition, activation analyses, thermal-hydraulics, tritium inventory, structural analyses, and water coolant conditions are summarized in this paper. 12 refs., 2 figs., 1 tab.

  8. An assessment of the base blanket for ITER

    SciTech Connect

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

    1991-12-31

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

  9. An assessment of the base blanket for ITER

    SciTech Connect

    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.

  10. ITER solid breeder blanket materials database

    SciTech Connect

    Billone, M.C.; Dienst, W.; Flament, T.; Lorenzetto, P.; Noda, K.; Roux, N.

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 2}) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized.

  11. Assessment of alkali metal coolants for the ITER blanket

    SciTech Connect

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

    1994-12-31

    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.

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

    SciTech Connect

    Wang, Q.; Henderson, D.L.

    1994-12-31

    The International Thermonuclear Experimental Reactor (ITER) project began a new design phase called the Engineering Design Activity (EDA) which started in July 1992. A variety of blanket designs options were analyzed as a part of the U.S. ITER home team blanket option trade-off study (BOTS) which began in May 1993. The options considered were a self-cooled Li/V blanket, a helium cooled Li/V blanket and a water cooled 316 SS nonbreeding shield option. Detailed activation, dose rate and waste disposal rating calculations have been performed for these different ITER blanket design options based on a fluence of 3.0 MWa/m{sup 2} and an average neutron wall loading of 2.0 MW/m{sup 2}. A continuous operation assumption was utilized in the analysis. The results of this work are presented in this conference.

  13. Assessment of alkali metal coolants for the ITER blanket

    SciTech Connect

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

    1994-06-01

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

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

    SciTech Connect

    Not Available

    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. ITER R & D on breeder blanket materials*1

    NASA Astrophysics Data System (ADS)

    Smith, D. L.; Daenner, W.; Kalinin, G.; Yoshida, H.

    1991-03-01

    The International Thermonuclear Experimental Reactor (ITER) activity includes specific R&D in critical areas to support the ITER design. Several of the critical tasks defined as part of the ITER validating R&D are materials related. This paper presents a summary of the breeder/blanket materials related R&D conducted by the four participating parties; viz., European Community, Japan, Soviet Union and the United States. The current effort includes several subtasks in each of the following four task areas: (1) ceramic breeder materials, (2) lead-lithium breeder, (3) aqueous salt solutions, (4) neutron multiplier materials. The objective and scope of each subtask is stated. Although this work was initiated in late 1988, important results have been obtained in all four areas. The work on ceramic breeders includes properties measurements, compatibility studies, and tritium recovery experiments with Li 2O and selected ternary ceramics. Investigations on the lead-lithium breeder have focused primarily on properties, compatibility, and tritium extraction from the 17Li-83Pb eutectic alloy. The work on the aqueous salt breeders includes corrosion, solution chemistry, and radiolysis effects in a radiation environment. Investigations defined under the neutron multiplier task are focused on beryllium; however, lead in the lithium lead alloy also serves as a neutron multiplier. The schedule and planned future work in each area are summarized.

  16. ITER Test Blanket Module Error Field Simulation Experiments

    NASA Astrophysics Data System (ADS)

    Schaffer, M. J.

    2010-11-01

    Recent experiments at DIII-D used an active-coil mock-up to investigate effects of magnetic error fields similar to those expected from two ferromagnetic Test Blanket Modules (TBMs) in one ITER equatorial port. The largest and most prevalent observed effect was plasma toroidal rotation slowing across the entire radial profile, up to 60% in H-mode when the mock-up local ripple at the plasma was ˜4 times the local ripple expected in front of ITER TBMs. Analysis showed the slowing to be consistent with non-resonant braking by the mock-up field. There was no evidence of strong electromagnetic braking by resonant harmonics. These results are consistent with the near absence of resonant helical harmonics in the TBM field. Global particle and energy confinement in H-mode decreased by <20% for the maximum mock-up ripple, but <5% at the local ripple expected in ITER. These confinement reductions may be linked with the large velocity reductions. TBM field effects were small in L-mode but increased with plasma beta. The L-H power threshold was unaffected within error bars. The mock-up field increased plasma sensitivity to mode locking by a known n=1 test field (n = toroidal harmonic number). In H-mode the increased locking sensitivity was from TBM torque slowing plasma rotation. At low beta, locked mode tolerance was fully recovered by re-optimizing the conventional DIII-D ``I-coils'' empirical compensation of n=1 errors in the presence of the TBM mock-up field. Empirical error compensation in H-mode should be addressed in future experiments. Global loss of injected neutral beam fast ions was within error bars, but 1 MeV fusion triton loss may have increased. The many DIII-D mock-up results provide important benchmarks for models needed to predict effects of TBMs in ITER.

  17. ITER blanket and shield studies for high aspect ratio design option

    SciTech Connect

    Gohar, Y.; Attaya, H.; Billone, M.; Johnson, C.; Kopasz, J.; Majumdar, S.; Mattas, R.; Smith, D.L. ); Lousteau, D.; Nelson, B.; Williamson, D. ); Raffray, A.; Badawi, A.; Ying, A.; Abdou, M. ); El-Guebaly, L.; Sawan, M.; Sviatoslavsky, I.; Mogahed, E.; Kulcinski, G. (Wisconsin U

    1991-01-01

    The attractiveness of the high aspect ratio design (HARD) option for ITER has motivated a study to assess the blanket and shield design performance for this configuration relative to the ITER reference design. The blanket and shield have been configured to take an advantage of the HARD option. The layered solid breeder blanket concept with water-coolant and steel-structure and the water-steel shield have been used. The changes in the neutron wall loading distribution, the mechanical design, the net tritium breeding ratio, the total tritium inventory, and the nuclear heating profiles are evaluated. The tradeoff between the net tritium breeding ratio, and the fuel operating cost is analyzed. The mechanical design and the structural interaction between the first wall and the blanket is studied. 4 refs., 3 figs., 5 tabs.

  18. Neutronics and shielding results of the U.S. ITER blanket trade-off study

    SciTech Connect

    Sawan, M.E.; El-Guebaly, L.A.

    1994-12-31

    The first formal phase of the International Thermonuclear Experimental Reactor (ITER) project was the Conceptual Design Activity (CDA) which was conducted from May 1988 until December 1990. The CDA blanket utilized Li{sub 2}O solid breeder, beryllium neutron multiplier, 316 SS structure and low temperature water coolant. The ITER project embarked on a new phase called the Engineering Design Activity (EDA) which started in July 1992. A variety of blanket design options have been considered since the start of the EDA. The US ITER home team started a blanket option trade-off-study (BOTS) in May 1993. The options considered included a self-cooled Li/V option, ferritic steel designs with helium cooled first wall and liquid metal (Li or LiPb) cooled blanket, the CDA design option, and a water cooled 316/SS nonbreeding shield option. Detailed neutronics and shielding analysis have been performed for the different ITER blanket design options considered in this study. The shielding performance was determined for these options and inboard shielding space requirements were compared. Peak first wall (FW) nuclear performance parameters were calculated in addition to power density distribution in the blanket required for the thermal hydraulics calculations. Peak vacuum vessel (VV) nuclear performance parameters have also been determined. The overall tritium breeding ratio (TBR) was determined for each of the design options. A double wall Inconel 625 VV is used with water cooled 316 SS balls. For the design options with liquid metal in the blanket, the use of tungsten carbide balls cooled by NaK is considered. The minimum total thickness of FW/B/S/W is maintained at 1 m in the inboard region.

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

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

    DOE PAGES

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

    2016-11-03

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

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

    SciTech Connect

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

    2016-11-03

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

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

    SciTech Connect

    Not Available

    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.

  3. Modeling and Simulation of the ITER First Wall/Blanket Primary Heat Transfer System

    SciTech Connect

    Ying, Alice; Popov, Emilian L

    2011-01-01

    ITER inductive power operation is modeled and simulated using a thermal-hydraulics system code (RELAP5) integrated with a 3-D CFD (SC-Tetra) code. The Primary Heat Transfer System (PHTS) functions are predicted together with the main parameters operational ranges. The control algorithm strategy and derivation are summarized as well. The First Wall and Blanket modules are the primary components of PHTS, used to remove the major part of the thermal heat from the plasma. The modules represent a set of flow channels in solid metal structure that serve to absorb the radiation heat and nuclear heating from the fusion reactions and to provide shield for the vacuum vessel. The blanket modules are water cooled. The cooling is forced convective with constant blanket inlet temperature and mass flow rate. Three independent water loops supply coolant to the three blanket sectors. The main equipment of each loop consists of a pump, a steam pressurizer and a heat exchanger. A major feature of ITER is the pulsed operation. The plasma does not burn continuously, but on intervals with large periods of no power between them. This specific feature causes design challenges to accommodate the thermal expansion of the coolant during the pulse period and requires active temperature control to maintain a constant blanket inlet temperature.

  4. Test Strategy for the European HCPB Test Blanket Module in ITER

    SciTech Connect

    Boccaccini, L.V.; Meyder, R.; Fischer, U.

    2005-05-15

    According to the European Blanket Programme two blanket concepts, the Helium Cooled Pebble Bed (HCPB) and a Helium Cooled Lithium Lead (HCLL) will be tested in ITER. During 2004 the test blanket modules (TBM) of both concepts were redesigned with the goal to use as much as possible similar design options and fabrication techniques for both types in order to reduce the European effort for TBM development. The result is a robust TBM box being able to withstand 8 MPa internal pressure in case of in-box LOCA; the TBM box consists of First wall (FW), caps, stiffening grid and manifolds. The box is filled with typically 18 and 24 breeding units (BU), for HCPB and HCLL respectively. A breeding unit has about 200 mm in poloidal and toroidal direction and about 400 mm in radial direction; the design is adapted to contain and cooling ceramic breeder/beryllium pebble beds for the HCPB and eutectic Lithium-Lead for the HCLL.The use of a new material, EUROFER, and the innovative design of these Helium Cooled components call for a large qualification programme before the installation in ITER; availability and safety of ITER should not be jeopardised by a failure of these components. Fabrication technologies especially in the welding processes (diffusion welding, EB, TIG, LASER) need to be tested in the manufacturing of large mock-ups; an extensive out-of-pile programme in Helium facility should be foreseen for the verification of the concept from basic helium cooling functions (uniformity of flow in parallel channels, heat transfer coefficient in FW, etc.) up to the verification of large portions of the TBM design under relevant ITER loading.In ITER the TBM will have the main objective to collect information that will contribute to the final design of DEMO blankets. A strategy has been proposed in 2001 that leads to the tests in ITER 4 different Test Blanket Modules (TBM's) type during the first 10 years of ITER operation. For the new HCPB design this strategy is confirmed with

  5. Measurements of breakaway reaction between beryllium and water vapor for ITER blanket design

    NASA Astrophysics Data System (ADS)

    Yoshida, Hiroshi; Enoeda, Mikio; Ashibe, Kusuo; Ono, Kiyoshi

    1992-06-01

    Beryllium will be utilized as the neutron multiplier in the ITER breeding blanket. As part of Japanese contribution of the water-cooled blanket design, preliminary study was performed to investigate beryllium breakaway reaction, which is one of safety issues of ITER under high temperature conditions. Thermogravimetric measurements in the temperature range 550 - 750 C were carried out under helium gas flow containing water vapor of 7.6 and 0.76 Torr. The test samples were prepared from commercially available hot-pressed and hot-rolled beryllium plates. Characterization of the surface reaction product was performed by macro and microscopic observations, and x-ray diffraction analysis. Linear and parabolic rate laws were found for the dominant reaction steps in the preceding period of the breakaway reaction. Microstructure of the surface reaction layer formed in all exposure conditions revealed brittle structure, which was composed of blister and microcracks. The rate equations were obtained for the preceding step of breakaway reaction.

  6. Feasibility study of ITER vacuum vessel and blanket back plate assembly plan

    SciTech Connect

    Nakahira, Masataka; Koizumi, Kouichi; Itou, Yutaka; Tada, Eisuke; Kanamori, Naokazu

    1996-12-31

    Design efforts have been made to develop the assembly procedure of the Vacuum Vessel (VV) and the Blanket (BL) Back Plate (BP) of the International Thermonuclear Experimental Reactor (ITER). This procedure must mitigate accumulation of welding distortion of the thick wall structure, handle and assemble massive structures such as the BP within narrow space and certain accuracy, and minimize assembly schedule. As a result, candidate procedures have been developed which rely on symmetric welding, limited rotation and reasonable schedule. Based on these procedures, the detail design of jigs and tools has been started. 2 refs., 5 figs., 1 tab.

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

    SciTech Connect

    Kasza, K.E.

    1992-02-01

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

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

    SciTech Connect

    Kasza, K.E.

    1992-02-01

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

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

    SciTech Connect

    Reed, C.B.; Haglund, R.C.; Miller, M.E.

    1995-08-01

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

  10. Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    NASA Astrophysics Data System (ADS)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-02-01

    This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

  11. Current status of final design and R&D for ITER blanket shield blocks in Korea

    NASA Astrophysics Data System (ADS)

    Ha, M. S.; Kim, S. W.; Jung, H. C.; Hwang, H. S.; Heo, Y. G.; Kim, D. H.; Ahn, H. J.; Lee, H. G.; Jung, K. J.

    2015-07-01

    The main function of the ITER blanket shield block (SB) is to provide nuclear shielding and support the first wall (FW) panel. It needs to accommodate all the components located on the vacuum vessel (in particular the in-vessel coils, blanket manifolds and the diagnostics). The conceptual, preliminary and final design reviews have been completed in the framework of the Blanket Integrated Product Team. The Korean Domestic Agency has successfully completed not only the final design activities, including thermo-hydraulic and thermo-mechanical analyses for SBs #2, #6, #8 and #16, but also the SB full scale prototype (FSP) pre-qualification program prior to issuing of the procurement agreement. SBs #2 and #6 are located at the in-board region of the tokamak. The pressure drop was less than 0.3 MPa and fully satisfied the design criteria. The thermo-mechanical stresses were also allowable even though the peak stresses occurred at nearby radial slit end holes, and their fatigue lives were evaluated over many more than 30 000 cycles. SB #8 is one of the most difficult modules to design, since this module will endure severe thermal loading not only from nuclear heating but also from plasma heat flux at uncovered regions by the FW. In order to resolve this design issue, the neutral beam shine-through module concept was applied to the FW uncovered region and it has been successfully verified as a possible design solution. SB #16 is located at the out-board central region of the tokamak. This module is under much higher nuclear loading than other modules and is covered by an enhanced heat flux FW panel. In the early design stage, many cooling headers on the front region were inserted to mitigate peak stresses near the access hole and radial slit end hole. However, the cooling headers on the front region needed to be removed in order to reduce the risk from cover welding during manufacturing. A few cooling headers now remain after efforts through several iterations to remove

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

    NASA Astrophysics Data System (ADS)

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

    2017-03-01

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

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

    DOE PAGES

    Lanctot, Matthew J.; Snipes, J. A.; Reimerdes, H.; ...

    2016-12-12

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

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

    SciTech Connect

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

    2016-12-12

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

  15. ITER structural design criteria and their extension to advanced fusion reactor blankets.

    SciTech Connect

    Kalnin, G.; Majumdar, S.

    1999-09-03

    Application of the new low-temperature-design rules of the ITER Structural Design Criteria (ISDC) is illustrated by considering copper alloys that, according to recent data, are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures, Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. High-temperature-design rules of ISDC are applied to EVOLVE (Evaporation Of Lithium and Vapor Extraction), a blanket design concept currently being investigated under the U.S. APEX (Advanced Power Extraction) program. One version of this concept envisions the use of a series of parallel tungsten tubes (first-wall) that are cooled internally by lithium vapor, typically. at 1200 C. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method.

  16. Neutronics Experiments Using Small Partial Mockups of the ITER Test Blanket Module with a Solid Breeder

    SciTech Connect

    Sato, Satoshi; Verzilov, Yury; Nakao, Makoto; Ochiai, Kentaro; Wada, Masayuki; Nishitani, Takeo

    2005-05-15

    In order to evaluate the impacts of the incident neutron spectrum and the tungsten armor on the tritium production, integral experiments have been performed with small partial mockups relevant to the ITER test blanket module using DT neutrons at FNS of JAERI. The Monte Carlo calculation results for the integrated tritium productions agree well with the experimental data within 2 and 11 % for the mockups without the armor in the experiments without and with the neutron reflector, respectively. It is clarified that the tritium production can be very accurately predicted in the experiment without the reflector. In the mockups with the 12.6 and 25.2 mm thick tungsten armors, it is experimentally clarified that the integrated tritium productions are reduced by 3 and 6 % relative to the case without the armor, respectively.

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

    SciTech Connect

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

    1996-04-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2011-10-01

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

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

    SciTech Connect

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

    2011-01-01

    Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Delta upsilon/upsilon similar to 60% via non-resonant braking. Changes to global Delta n/n, Delta beta/beta and Delta H(98)/H(98) were similar to 3 times smaller. These effects are stronger at higher beta. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L-and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

  20. Corrosion studies of a stainless steel structure for the ITER (International Thermonuclear Experimental Reactor) aqueous lithium salt blanket concept

    SciTech Connect

    Duquette, D.J.; Wrisley, K.L.; Motyka, E.; Steiner, D.; Embrechts, M.J.

    1990-10-01

    The aqueous lithium slat blanket (ALSB) employs water, with a dissolved lithium compound, as both the coolant and tritium breeding medium. The ALSB concept is one of three blanket options currently being examined for breeding tritium in the International Thermonuclear Experimental Reactor (ITER). To provide data and recommendations for materials and chemistry selection relevant to application of the ALSB in ITER, corrosion studies have been initiated, focusing on Type 316 stainless steel in lithium hydroxide and lithium nitrate solutions. This report presents the preliminary results of these corrosion studies. The results to date, while preliminary, suggest that even at 90 C, a blanket utilizing 10% LiOH (the current lithium salt of choice for ITER ALSB applications) will not cause catastrophic failure of 316 stainless steel by either stress corrosion cracking or localized corrosion; that the general corrosion rate will not exceed about 40 {mu}m/yr and transport of material will certainly be much less than this value since most of the corrosion product will be included in the strong adherent surface film; and that, although hydrogen may be evolved due to electrolysis, the maximum amount of hydrogen is small compared to that expected to be produced by radiolysis. These observations are predicated on the assumption that the blanket will be completely deaerated, and that the corrosion potential of the alloy will be similar to that observed in the laboratory.

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

    SciTech Connect

    Smith, D.L.; Mattas, R.F.

    1997-07-01

    The objective of this task is to develop the required data base and demonstrate the performance of a liquid lithium-vanadium advanced blanket design. The task has two main activities related to vanadium structural material and liquid lithium system developments. The vanadium alloy development activity included four subtasks: (1.1) baseline mechanical properties of non irradiated base metal and weld metal joints; (1.2) compatibility with liquid lithium; (1.3) material irradiation tests; and (1.4) development of material manufacturing and joining methods. The lithium blanket technology activity included four subtasks: (2.1) electrical insulation development and testing for liquid metal systems; (2.2) MHD pressure drop and heat transfer study for self-cooled liquid metal systems; (2.3) chemistry of liquid lithium; and (2.4) design, fabrication and testing of ITER relevant size blanket mockups. A summary of the progress and results obtained during the period 1995 and 1996 in each of the subtask areas is presented in this report.

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

    SciTech Connect

    Ulrickson, M.A.; Manly, W.D.; Dombrowski, D.E.

    1995-08-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-06-01

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

  4. Semi-Technical Cryogenic Molecular Sieve Bed for the Tritium Extraction System of the Test Blanket Module for ITER

    SciTech Connect

    Beloglazov, S.; Bekris, N.; Glugla, M.; Wagner, R.

    2005-07-15

    The tritium extraction from the ITER Helium Cooled Pebble Bed (HCPB) Test Blanket Module purge gas is proposed to be performed in a two steps process: trapping water in a cryogenic Cold Trap, and adsorption of hydrogen isotopes (H{sub 2}, HT, T{sub 2}) as well as impurities (N{sub 2}, O{sub 2}) in a Cryogenic Molecular Sieve Bed (CMSB) at 77K. A CMSB in a semi-technical scale (one-sixth of the flow rate of the ITER-HCPB) was design and constructed at the Forschungszentrum Karlsruhe. The full capacity of CMSB filled with 20 kg of MS-5A was calculated based on adsorption isotherm data to be 9.4 mol of H{sub 2} at partial pressure 120 Pa. The breakthrough tests at flow rates up to 2 Nm{sup 3}h{sup -1} of He with 110 Pa of H{sub 2} conformed with good agreement the adsorption capacity of the CMSB. The mass-transfer zone was found to be relatively narrow (12.5 % of the MS Bed height) allowing to scale up the CMSB to ITER flow rates.

  5. Non-LTE line-blanketed model atmospheres of hot stars. 1: Hybrid complete linearization/accelerated lambda iteration method

    NASA Technical Reports Server (NTRS)

    Hubeny, I.; Lanz, T.

    1995-01-01

    A new munerical method for computing non-Local Thermodynamic Equilibrium (non-LTE) model stellar atmospheres is presented. The method, called the hybird complete linearization/accelerated lambda iretation (CL/ALI) method, combines advantages of both its constituents. Its rate of convergence is virtually as high as for the standard CL method, while the computer time per iteration is almost as low as for the standard ALI method. The method is formulated as the standard complete lineariation, the only difference being that the radiation intensity at selected frequency points is not explicity linearized; instead, it is treated by means of the ALI approach. The scheme offers a wide spectrum of options, ranging from the full CL to the full ALI method. We deonstrate that the method works optimally if the majority of frequency points are treated in the ALI mode, while the radiation intensity at a few (typically two to 30) frequency points is explicity linearized. We show how this method can be applied to calculate metal line-blanketed non-LTE model atmospheres, by using the idea of 'superlevels' and 'superlines' introduced originally by Anderson (1989). We calculate several illustrative models taking into accont several tens of thosands of lines of Fe III to Fe IV and show that the hybrid CL/ALI method provides a robust method for calculating non-LTE line-blanketed model atmospheres for a wide range of stellar parameters. The results for individual stellar types will be presented in subsequent papers in this series.

  6. Conceptual design description for the tritium recovery system for the US ITER (International Thermonuclear Experimental Reactor) Li sub 2 O/Be water cooled blanket

    SciTech Connect

    Finn, P.A.; Sze, D.K. . Fusion Power Program); Clemmer, R.G. )

    1990-11-01

    The tritium recovery system for the US ITER Li{sub 2}O/Be water cooled blanket processes two separate helium purge streams to recover tritium from the Li{sub 2}O zones and the Be zones of the blanket, to process the waste products, and to recirculate the helium back to the blanket. The components are selected to minimize the tritium inventory of the recovery system, and to minimize waste products. The system is robust to either an increase in the tritium release rate or to an in-leak of water in the purge system. Three major components were used to process these streams, first, 5A molecular sieves at {minus}196{degree}C separate hydrogen from the helium, second, a solid oxide electrolysis unit is used to reduce all molecular water, and third, a palladium/silver diffuser is used to ensure that only hydrogen (H{sub 2}, HT) species reach the cryogenic distillation unit. Other units are present to recover tritium from waste products but the three major components are the basis of the blanket tritium recovery system. 32 refs.

  7. Conceptual design description for the tritium recovery system for the US ITER Li sub 2 O/Be water cooled blanket

    SciTech Connect

    Finn, P.A.; Sze, D.K. ); Clemmer, R.G. )

    1990-09-01

    The tritium recovery system for the US ITER Li{sub 2}O/Be water cooled blanket processes two separate helium purge streams to recovery tritium from the Li{sub 2}O zones and the Be zones of the blanket, to process the waste products, and to recirculate the helium back to the blanket. The components are selected to minimize the tritium inventory of the recovery system, and to minimize waste products. The system is robust to either an increase in the tritium release rate or to an in-leak of water in the purge system. Three major components were used to process these streams, first, 5A molecular sieves at {minus}196{degree}C separate hydrogen from the helium, second, a solid oxide electrolysis unit is used to reduce all molecular water, and third, a palladium/silver diffuser is used to ensure that only hydrogen (H{sub 2}, HT) species reach the cryogenic distillation unit. The total tritium process inventory is 20g. The total capital cost is {approximately}$14M. Technical advantages of a solid oxide electrolysis unit and a palladium/silver diffuser are presented. 5 refs., 1 fig., 3 tabs.

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

    SciTech Connect

    Not Available

    1983-12-30

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

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

    NASA Astrophysics Data System (ADS)

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

    2009-06-01

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

  10. Modified iterative aggregation procedure for maintenance optimisation of multi-component systems with failure interaction

    NASA Astrophysics Data System (ADS)

    Zhang, Zhuoqi; Wu, Su; Lee, Seungchul; Ni, Jun

    2014-12-01

    This paper studies maintenance policies for multi-component systems which have failure interaction among their components. Component failure might accelerate deterioration processes or induce instantaneous failures of the remaining components. We formulate this maintenance problem as a Markov decision process (MDP) with an objective of minimising a total discounted maintenance cost. However, the action set and state space in MDP exponentially grow as the number of components increases. This makes traditional approaches computationally intractable. To deal with this curse of dimensionality, a modified iterative aggregation procedure (MIAP) is proposed. We mathematically prove that iterations in MIAP guarantee the convergence and the policy obtained is optimal. Numerical case studies find that failure interaction should not be ignored in a maintenance policy decision making and the proposed MIAP is faster and requires less computational memory size than that of linear programming.

  11. Optimization of HIP bonding conditions for ITER shielding blanket/first wall made from austenitic stainless steel and dispersion strengthened copper alloy

    NASA Astrophysics Data System (ADS)

    Sato, S.; Hatano, T.; Kuroda, T.; Furuya, K.; Hara, S.; Enoeda, M.; Takatsu, H.

    1998-10-01

    Optimum bonding conditions were studied on the Hot Isostatic Pressing (HIP) bonded joints of type 316L austenitic stainless steel and Dispersion Strengthened Copper alloy (DSCu) for application to the ITER shielding blanket / first wall. HIP bonded joints were fabricated at temperatures in a 980-1050°C range, and a series of mechanical tests and metallurgical observations were conducted on the joints. Also, bondability of two grades of DSCu (Glidcop Al-25 ® and Al-15 ®) with SS316L was examined in terms of mechanical properties of the HIP bonded joints. From those studies it was concluded that the HIP temperature of 1050°C was an optimal condition for obtaining higher ductility, impact values and fatigue strength. Also, SS316L/Al-15 joints showed better results in terms of ductility and impact values compared with SS316L/Al-25 joints.

  12. e-Learning Application for Machine Maintenance Process using Iterative Method in XYZ Company

    NASA Astrophysics Data System (ADS)

    Nurunisa, Suaidah; Kurniawati, Amelia; Pramuditya Soesanto, Rayinda; Yunan Kurnia Septo Hediyanto, Umar

    2016-02-01

    XYZ Company is a company based on manufacturing part for airplane, one of the machine that is categorized as key facility in the company is Millac 5H6P. As a key facility, the machines should be assured to work well and in peak condition, therefore, maintenance process is needed periodically. From the data gathering, it is known that there are lack of competency from the maintenance staff to maintain different type of machine which is not assigned by the supervisor, this indicate that knowledge which possessed by maintenance staff are uneven. The purpose of this research is to create knowledge-based e-learning application as a realization from externalization process in knowledge transfer process to maintain the machine. The application feature are adjusted for maintenance purpose using e-learning framework for maintenance process, the content of the application support multimedia for learning purpose. QFD is used in this research to understand the needs from user. The application is built using moodle with iterative method for software development cycle and UML Diagram. The result from this research is e-learning application as sharing knowledge media for maintenance staff in the company. From the test, it is known that the application make maintenance staff easy to understand the competencies.

  13. Iter

    NASA Astrophysics Data System (ADS)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

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

    SciTech Connect

    Ricapito, I.; Calderoni, P.; Poitevin, Y.; Aiello, A.; Utili, M.; Demange, D.

    2015-03-15

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

  15. An Iterative Approach for the Optimization of Pavement Maintenance Management at the Network Level

    PubMed Central

    Torres-Machí, Cristina; Chamorro, Alondra; Videla, Carlos; Yepes, Víctor

    2014-01-01

    Pavement maintenance is one of the major issues of public agencies. Insufficient investment or inefficient maintenance strategies lead to high economic expenses in the long term. Under budgetary restrictions, the optimal allocation of resources becomes a crucial aspect. Two traditional approaches (sequential and holistic) and four classes of optimization methods (selection based on ranking, mathematical optimization, near optimization, and other methods) have been applied to solve this problem. They vary in the number of alternatives considered and how the selection process is performed. Therefore, a previous understanding of the problem is mandatory to identify the most suitable approach and method for a particular network. This study aims to assist highway agencies, researchers, and practitioners on when and how to apply available methods based on a comparative analysis of the current state of the practice. Holistic approach tackles the problem considering the overall network condition, while the sequential approach is easier to implement and understand, but may lead to solutions far from optimal. Scenarios defining the suitability of these approaches are defined. Finally, an iterative approach gathering the advantages of traditional approaches is proposed and applied in a case study. The proposed approach considers the overall network condition in a simpler and more intuitive manner than the holistic approach. PMID:24741352

  16. An iterative approach for the optimization of pavement maintenance management at the network level.

    PubMed

    Torres-Machí, Cristina; Chamorro, Alondra; Videla, Carlos; Pellicer, Eugenio; Yepes, Víctor

    2014-01-01

    Pavement maintenance is one of the major issues of public agencies. Insufficient investment or inefficient maintenance strategies lead to high economic expenses in the long term. Under budgetary restrictions, the optimal allocation of resources becomes a crucial aspect. Two traditional approaches (sequential and holistic) and four classes of optimization methods (selection based on ranking, mathematical optimization, near optimization, and other methods) have been applied to solve this problem. They vary in the number of alternatives considered and how the selection process is performed. Therefore, a previous understanding of the problem is mandatory to identify the most suitable approach and method for a particular network. This study aims to assist highway agencies, researchers, and practitioners on when and how to apply available methods based on a comparative analysis of the current state of the practice. Holistic approach tackles the problem considering the overall network condition, while the sequential approach is easier to implement and understand, but may lead to solutions far from optimal. Scenarios defining the suitability of these approaches are defined. Finally, an iterative approach gathering the advantages of traditional approaches is proposed and applied in a case study. The proposed approach considers the overall network condition in a simpler and more intuitive manner than the holistic approach.

  17. Progress on DCLL Blanket Concept

    SciTech Connect

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

    2013-09-01

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

  18. Mechanical design issues associated with mounting, maintenance, and handling of an ITER divertor

    SciTech Connect

    Goranson, P.L.; Fogarty, P.J.; Jones, G.H.

    1991-01-01

    Several designs that address plasma-facing plate configurations and thermal-hydraulic design issues have been developed for the ITER divertor. Design criteria growing out of physics requirements, physical constraints, and remote handling requirements impose severe mechanical requirements on the support structure and its attachments. These pose a challenge to the mechanical design of a divertor, which must be addressed before a functional divertor is practical -- that is, one that can be remotely handled, aligned, and maintained; that functions reliably under thermal loading and disruptions; and that gives the required life in the nuclear environment predicted for ITER. This paper discusses the design criteria for the divertor mounting structure and identifies the mechanical design issues that need to be addressed. Achieving the criteria may require the development of new components and innovative configurations, specifically a new class of remote fasteners and electrically resistant material for mounts. The possible design of such components and an R D program to develop them are described, and issues specific to the high-aspect-ratio design (HARD) configuration are summarized. Analysis and experiments that will resolve these issues and concerns and lead to a final ITER design are identified. 2 refs., 2 figs.

  19. BLANKET BIOLOGICAL REVIEW FOR GENERAL MAINTENANCE ACTIVITIES WITHIN ACTIVE BURIAL GROUNDS, 200 E and 200 W Areas, ECR No. 2002-200-034

    SciTech Connect

    Sackschewsky, Michael R.

    2003-06-26

    No plant and animal species protected under the ESA, candidates for such protection, or species listed by the Washington state government were observed in the vicinity of the proposed sites. Piper's daisy may still occur in some of the burial grounds. This is a Washington State Sensitive plant species, and as such is a Level III resource under the Hanford Site Biological Resources Management Plan. Compensatory mitigation is appropriate for this species when adverse impacts cannot be avoided. The Ecological Compliance Assessment Project (ECAP) staff should consulted prior to the initiation of major work activities within areas where this species has been identified (218-E-12, 218-E-10). The stalked-pod and crouching milkvetch are relatively common throughout 200 West area, therefore even if the few individuals within the active burial grounds are disturbed, it is not likely that the overall local population will be adversely affected. The Watch List is the lowest level of listing for pl ant species of concern in the State of Washington. No adverse impacts to species or habitats of concern are expected to occur from routine maintenance within the active portions of the 218-W-4C, 218-W-4B, 218-W-3, 218-W-3A, and 218-W-5 burial grounds, as well as the portion of 218-E-12B currently used for storage of retired submarine reactor cores. The remaining portions of the 218-E-12B burial ground, the entire 218-E-10 burial ground, and the 218-W-6 burial ground currently have extensive vegetative cover and it is highly likely that migratory birds, such as meadow larks, horned larks, and curlews will nest in these areas. Therefore, it is recommended that if removal of the existing vegetation is required for burial ground operations, such removal only occur during the August through March time period (i.e. when the birds are not actively nesting). This blanket review does not apply to the portions of 218-W-4C, and 218-W-6 previously described.

  20. BLANKET BIOLOGICAL REVIEW FOR GENERAL MAINTENANCE ACTIVITIES WITHIN ACTIVE BURIAL GROUNDS, 200 E and 200 W Areas, ECR No.2001-200-048

    SciTech Connect

    Sackschewsky, Michael R.

    2002-05-08

    No plant and animal species protected under the ESA, candidates for such protection, or species listed by the Washington state government were observed in the vicinity of the proposed sites. Piper's daisy may still occur in some of the burial grounds. This is a Washington State Sensitive plant species, and as such is a Level III resource under the Hanford Site Biological Resources Management Plan. Compensatory mitigation is appropriate for this species when adverse impacts cannot be avoided. The Ecological Compliance Assessment Project (ECAP) staff should consulted prior to the initiation of major work activities within areas where this species has been identified (218-E-12, 218-E-10). The stalked-pod and crouching milkvetch are relatively common throughout 200 West area, therefore even if the few individuals within the active burial grounds are disturbed, it is not likely that the overall local population will be adversely affected. The Watch List is the lowest level of listing for plant species of concern in the State of Washington. No adverse impacts to species or habitats of concern are expected to occur from routine maintenance within the active portions of the 218-W-4C, 218-W-4B, 218-W-3, 218-W-3A, and 218-W-5 burial grounds, as well as the portion of 218-E-12B currently used for storage of retired submarine reactor cores. The remaining portions of the 218-E-12B burial ground, the entire 218-E-10 burial ground, and the 218-W-6 burial ground currently have extensive vegetative cover and it is highly likely that migratory birds, such as meadow larks, horned larks, and curlews will nest in these areas. Therefore, it is recommended that if removal of the existing vegetation is required for burial ground operations, such removal only occur during the August through March time period (i.e. when the birds are not actively nesting). This blanket review does not apply to the portions of 218-W-4C, and 218-W-6 previously described.

  1. Blanket Biological Review for General Maintenance Activities Within Active Burial Grounds, 200 East and 200 West Areas, ECR No.2003-200-035

    SciTech Connect

    Sackschewsky, Michael R.

    2003-08-25

    No plant and animal species protected under the ESA, candidates for such protection, or species listed by the Washington state government were observed in the vicinity of the proposed sites. Piper's daisy may still occur in some of the burial grounds (218-E-12, 218-E-10). This is a Washington State Sensitive plant species, and as such is a Level III resource under the Hanford Site Biological Resources Management Plan. Compensatory mitigation is appropriate for this species when adverse impacts cannot be avoided. The stalked-pod and crouching milkvetch, Watch List species, are relatively common throughout 200 West area, therefore even if the few individuals within the active burial grounds are disturbed, it is not likely that the overall local population will be adversely affected. The Watch List is the lowest level of listing for plant species of concern in the State of Washington. No adverse impacts to species or habitats of concern are expected to occur from routine maintenance within the active portions of the 218-W-4C, 218-W-4B, 218-W-3, 218-W-3A, and 218-W-5 burial grounds, as well as the portion of 218-E-12B currently used for storage of retired submarine reactor cores. The remaining portions of the 218-E-12B burial ground and the entire 218-E-10 burial ground currently have extensive vegetative cover and it is highly likely that migratory birds, such as meadowlarks, horned larks, and curlews may nest in these areas. Therefore, it is recommended that if removal of the existing vegetation is required for burial ground operations, such removal only occur during the August through March time period (i.e. when the birds are not actively nesting). If vegetation removal is required prior to August 2003 or after 1 April 2004, please contact the ECAP staff for an additional analysis to ensure compliance with the Migratory Bird Treaty Act. Workers should be advised to watch for nesting birds within the burial grounds, if any are encountered, please contact the ECAP

  2. BLANKET BIOLOGICAL REVIEW FOR GENERAL MAINTENANCE ACTIVITIES WITHIN ACTIVE BURIAL GROUNDS, 200 E and 200 W Areas, ECR #2000-200-013

    SciTech Connect

    Sackschewsky, Michael R.

    2002-04-04

    No plant and animal species protected under the ESA, candidates for such protection, or species listed by the Washington state government were observed in the vicinity of the proposed sites. Piper's daisy may still occur in some of the burial grounds. This is a Washington State Sensitive plant species, and as such is a Level III resource under the Hanford Site Biological Resources Management Plan. Compensatory mitigation is appropriate for this species when adverse impacts cannot be avoided. The Ecological Compliance Assessment Project (ECAP) staff should consulted prior to the initiation of major work activities within areas where this species has been identified (218-E-12, 218-E-10). The stalked-pod and crouching milkvetch are relatively common throughout 200 West area, therefore even if the few individuals within the active burial grounds are disturbed, it is not likely that the overall local population will be adversely affected. The Watch List is the lowest level of listing for plant species of concern in the State of Washington. No adverse impacts to species or habitats of concern are expected to occur from routine maintenance within the active portions of the 218-W-4C, 218-W-4B, 218-W-3, 218-W-3A, and 218-W-5 burial grounds, as well as the portion of 218-E-12B currently used for storage of retired submarine reactor cores. The remaining portions of the 218-E-12B burial ground, the entire 218-E-10 burial ground, and the 218-W-6 burial ground currently have extensive vegetative cover and it is highly likely that migratory birds, such as meadow larks, horned larks, and curlews will nest in these areas. Therefore, it is recommended that if removal of the existing vegetation is required for burial ground operations, such removal only occur during the August through March time period (i.e. when the birds are not actively nesting). This blanket review does not apply to the portions of 218-W-4C, and 218-W-6 previously described.

  3. A PRELIMINARY ASSESSMENT OF THE OCCUPATIONAL RADIATION EXPOSURE FROM MAINTAINING THE US ITER DCLL TBM

    SciTech Connect

    B. J. Merrill; L. C. Cadwallader; M. Dagher

    2008-09-01

    This paper details an Occupational Radiation Exposure (ORE) analysis performed for the US International Thermonuclear Experimental Reactor (ITER) Dual Coolant Lead Lithium (DCLL) Test Blanket Module (TBM). This ORE analysis was performed with the QADMOD dose code for maintenance activities anticipated for the US DCLL TBM concept and its ancillary systems. Identification of the maintenance tasks that will have to be performed and estimates of the time required to perform these tasks were developed based on either expert opinion or on industrial maintenance experience for similar technologies. This paper details the modeling activity and the calculated doses for the maintenance activities envisioned for the US DCLL TBM.

  4. Occupational Radiation Exposure Analysis of US ITER DCLL TBM

    SciTech Connect

    Merrill, Brad J; Cadwallader, Lee C; Dagher, Mohamad

    2007-08-01

    This report documents an Occupational Radiation Exposure (ORE) analysis that was performed for the US International Thermonuclear Experimental Reactor (ITER) Dual Coolant Lead Lithium (DCLL) Test Blanket Module (TBM). This analysis was performed with the QADMOD dose code for anticipated maintenance activities for this TBM concept and its ancillary systems. The QADMOD code was used to model the PbLi cooling loop of this TBM concept by specifying gamma ray source terms that simulated radioactive material within the piping, valves, heat exchanger, permeator, pump, drain tank, and cold trap of this cooling system. Estimates of the maintenance tasks that will have to be performed and the time required to perform these tasks where developed based on either expert opinion or on industrial maintenance experience for similar technologies. This report details the modeling activity and the calculated doses for the maintenance activities envisioned for the US DCLL TBM.

  5. Summary report for ITER Task -- D4: Activation calculations for the stainless steel ITER design

    SciTech Connect

    Attaya, H.

    1995-02-01

    Detailed activation analysis for ITER has been performed as a part of ITER Task D4. The calculations have been performed for the shielding blanket (SS/water) and for the breeding blanket (LiN) options. The activation code RACC-P, which has been modified under IFER Task-D-10 for pulsed operation, has been used in this analysis. The spatial distributions of the radioactive inventory, decay heat, biological hazard potential, and the contact dose were calculated for the two designs for different operation modes and targeted fluences. A one-dimensional toroidal geometrical model has been utilized to determine the neutron fluxes in the two designs. The results are normalized for an inboard and outboard neutron wall loadings of 0.91 and 1.2 MW/M{sup 2}, respectively. The point-wise distributions of the decay gamma sources have been calculated everywhere in the reactor at several times after the shutdown of the two designs and are then used in the transport code ONEDANT to calculate the biological dose everywhere in the reactor. The point-wise distributions of all the responses have also been calculated. These calculations have been performed for neutron fluences of 3.0 MWa/M{sup 2}, which corresponds to the target fluence of ITER, and 0.1 MWa/M{sup 2}, which is anticipated to correspond to the beginning of an extended maintenance period.

  6. Ejecta Blanket on Vesta

    NASA Image and Video Library

    2011-10-13

    This image from NASA Dawn spacecraft shows an ejecta blanket mantling the surface and obscuring older caters. The bright crater rim, seen in the middle right edge, is one of a group of craters which are the source of this ejecta blanket.

  7. Two-dimensional over-all neutronics analysis of the ITER device

    NASA Astrophysics Data System (ADS)

    Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

    1993-07-01

    The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

  8. First Wall, Blanket, Shield Engineering Technology Program

    SciTech Connect

    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.

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

    DOE PAGES

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; ...

    2016-09-23

    '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,more » 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.« less

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

    SciTech Connect

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

    2016-09-23

    '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.

  11. Composite flexible blanket insulation

    NASA Technical Reports Server (NTRS)

    Kourtides, Demetrius A. (Inventor); Lowe, David M. (Inventor)

    1994-01-01

    An improved composite flexible blanket insulation is presented comprising top silicon carbide having an interlock design, wherein the reflective shield is composed of single or double aluminized polyimide and wherein the polyimide film has a honeycomb pattern.

  12. Blanket Biological Review for General Maintenance Activities within Active Burial Grounds, 200 E and 200 W Areas, ECR No. 99-200-042

    SciTech Connect

    Brandt, Charles A.

    1999-04-30

    No plant and animal species protected under the Endangered Species Act, candidates for such protection, or species listed by the Washington state government were observed in the vicinity of the proposed sites. Piper's daisy is a Washington State Sensitive plant species, and as such is a Level III resource under the Hanford Site Biological Resources Management Plan. Compensatory mitigation is appropriate for this species when adverse impacts cannot be avoided. The stalked pod and crouching milkvetchs are relatively common throughout 200 West area, therefore even if the few individuals within the active burial grounds are disturbed, it is not likely that the overall local population will be adversely affected. The Watch List is the lowest level of listing for plant species of concern in the State of Washington. No adverse impacts to species or habitats of concern are expected to occur from routine maintenance within the active portions of the 218-W-4C, 218-W-4B, 218-W-3, 218-W-3A, a nd 218-W-5 burial grounds, as well as the portion of 218-E-12B currently used for storage of retired submarine reactor cores. The remaining portions of the 218-E-12B burial ground, the entire 218-E-10 burial ground, and the 218-W-6 burial ground currently have extensive vegetative cover and it is highly likely that migratory birds, such as meadow larks, horned larks, and curlews will nest in these areas. Therefore, it is recommended that if removal of the existing vegetation is required for burial ground operations, such removal only occur during the August through March time period (i.e. when the birds are not actively nesting). If vegetation removal is required prior to August 1999 or after 1 April 2000, please contact the ECAP staff for an additional analysis to ensure compliance with the Migratory Bird Treaty Act.

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

    SciTech Connect

    Giancarli, L.; Proust, E.

    1994-12-31

    The European test-blanket development programme, started in 1988, is aiming at the selection by 1995 of two DEMO-relevant blanket lines to be tested in ITER. At present, four lines of blanket are under development, two of them using solid and the other two liquid breeder materials. As far as liquid breeders are concerned, two lines of blankets have been selected within the European Union, the water-cooled lithium-lead (the eutectic Pb-17Li) blankets and the dual-coolant Pb-17Li blankets. Designs have been developed considering an agreed set of DEMO specifications, such as, for instance, a fusion power of 2,200 MW, a neutron wall-loading of 2MW/m{sup 2}, a life-time of 20,000 hours, and the use of martensitic steel as a structural material. Moreover, an experimental program has been set up in order to address the main critical issues for each line. The present paper gives an overview of both design and experimental activities within the European Union concerning these two lines of liquid breeder blankets.

  14. Gauge Measures Thicknesses Of Blankets

    NASA Technical Reports Server (NTRS)

    Hagen, George R.; Yoshino, Stanley Y.

    1991-01-01

    Tool makes highly repeatable measurements of thickness of penetrable blanket insulation. Includes commercial holder for replaceable knife blades, which holds needle instead of knife. Needle penetrates blanket to establish reference plane. Ballasted slider applies fixed preload to blanket. Technician reads thickness value on scale.

  15. Thermal insulation blanket material

    NASA Astrophysics Data System (ADS)

    Pusch, R. H.

    1982-06-01

    A study was conducted to provide a tailorable advanced blanket insulation based on a woven design having an integrally woven core structure. A highly pure quartz yarn was selected for weaving and the cells formed were filled with a microquartz felt insulation.

  16. Thermal insulation blanket material

    NASA Technical Reports Server (NTRS)

    Pusch, R. H.

    1982-01-01

    A study was conducted to provide a tailorable advanced blanket insulation based on a woven design having an integrally woven core structure. A highly pure quartz yarn was selected for weaving and the cells formed were filled with a microquartz felt insulation.

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

    PubMed

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. Copyright © 2011 Elsevier Ltd. All rights reserved.

  18. Tailorable Advanced Blanket Insulation (TABI)

    NASA Technical Reports Server (NTRS)

    Sawko, Paul M.; Goldstein, Howard E.

    1987-01-01

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

  19. Thermal insulating conformal blanket

    NASA Technical Reports Server (NTRS)

    Barney, Andrea (Inventor); Whittington, Charles A (Inventor); Eilertson, Bryan (Inventor); Siminski, Zenon (Inventor)

    2003-01-01

    The conformal thermal insulating blanket may have generally rigid batting material covered by an outer insulating layer formed of a high temperature resistant woven ceramic material and an inner insulating layer formed of a woven ceramic fiber material. The batting and insulating layers may be fastened together by sewing or stitching using an outer mold layer thread fabricated of a high temperature resistant material and an inner mold layer thread of a ceramic fiber material. The batting may be formed to a composite structure that may have a firmness factor sufficient to inhibit a pillowing effect after the stitching to not more than 0.03 inch. The outer insulating layer and an upper portion of the batting adjacent the outer insulating layer may be impregnated with a ceramic coating material.

  20. Blanket comparison and selection study. Volume II

    SciTech Connect

    Not Available

    1983-10-01

    This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies. (MOW)

  1. Sonically-Welded Thermal Blankets

    NASA Technical Reports Server (NTRS)

    Bettini, R.; Citrin, F.

    1985-01-01

    Insulating layers joined by seams. Sonic welding saves time and reduces weight. Layers of blanket built up from alternating layers of polyester sheet and polyester net. Paper pattern showing lines along which layers are joined together placed over top layer.

  2. Methodology for accident analyses of fusion breeder blankets

    SciTech Connect

    Dobromir Panayotov; Andrew Grief; Brad J. Merrill; Julian T. Murgatroyd; Paul Humrickhouse; Yves Poitevin; Simon Owen; Markus Iseli

    2015-06-01

    'Fusion for Energy' (F4E) develops designs and implements the European Test Blanket Systems (TBS) in ITER - Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB). Safety demonstration is an essential element for the integration of TBS in ITER and accident analyses are one of its critical segments. A systematic approach to the accident analyses had been acquired under the F4E contract on TBS safety analyses. F4E technical requirements and AMEC and INL efforts resulted in the development of a comprehensive methodology for fusion breeding blanket accident analyses. It addresses the specificity of the breeding blankets design, materials and phenomena and at the same time is consistent with the one already applied to ITER accident analyses. Methodology consists of several phases. At first the reference scenarios are selected on the base of FMEA studies. In the second place elaboration of the accident analyses specifications we use phenomena identification and ranking tables to identify the requirements to be met by the code(s) and TBS models. Thus the limitations of the codes are identified and possible solutions to be built into the models are proposed. These include among others the loose coupling of different codes or code versions in order to simulate multi-fluid flows and phenomena. The code selection and issue of the accident analyses specifications conclude this second step. Furthermore the breeding blanket and ancillary systems models are built on. In this work challenges met and solutions used in the development of both MELCOR and RELAP5 codes models of HCLL and HCPB TBSs will be shared. To continue the developed models are qualified by comparison with finite elements analyses, by code to code comparison and sensitivity studies. Finally, the qualified models are used for the execution of the accident analyses of specific scenario. When possible the methodology phases will be illustrated in the paper by limited number of tables and figures

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  5. Current Trends of Blanket Research and Deveopment in Japan 2.Roles and Requirements of the Blanket System of Fusion Power Reactors

    NASA Astrophysics Data System (ADS)

    Asaoka, Yoshiyuki; Mohri, Kensuke; Hashizume, Hidetoshi; Tanaka, Satoru; Ueda, Yoshio

    Roles and requirements of the blanket system of the fusion power reactors are discussed from viewpoints of economics, fuel supply, generation system, maintenance, radioactive waste, and interaction with the plasma core. As the blanket system influences the cost of the fusion energy, the blanket system must be designed to minimize the fusion energy cost. Tritium breeding performance of the blanket is indispensable role to show the advantage of fusion energy on energy security. Material development for high temperature use under high neutron flux is one of the key issues of the generation system because the thermal efficiency depends on the coolant temperature of the blanket. Innovative maintenance technologies such as dividable superconducting coil system are very effective to make the fusion power reactor attractive. From viewpoints of natural resources and waste management, materials used in the fusion reactors should be recycled. Material selection is also of a large importance on the cost of radioactive waste disposal. Finally, it must be paid a careful attention that the design of the blanket system is inseparable from the achievement of a high performance plasma core.

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

    NASA Astrophysics Data System (ADS)

    Yu, Guanying; Liu, Xufeng; Liu, Songlin

    2016-10-01

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

  7. Neutron cameras for ITER

    SciTech Connect

    Johnson, L.C.; Barnes, C.W.; Batistoni, P.

    1998-12-31

    Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from {sup 16}N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with {sup 16}N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins.

  8. Foreign Blanket Orders: Precedent and practice

    ERIC Educational Resources Information Center

    Wedgeworth, Robert

    1970-01-01

    This investigation based upon a survey of 26 American and West European booksellers attemts to provide some basic information about foreign blanket orders. The characteristics of foreign blanket orders as well as the attitudes of the booksellers are summarized. (Author)

  9. 47 CFR 22.353 - Blanketing interference.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... are not required to resolve blanketing interference to mobile receivers or non-RF devices or... Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) COMMON CARRIER SERVICES PUBLIC MOBILE SERVICES... Public Mobile Services stations are responsible for resolving cases of blanketing interference...

  10. Ceramic Breeder Blanket for ARIES-CS

    SciTech Connect

    Raffray, A.R.; Malang, S.; El-Guebaly, L.; Wang, X.

    2005-05-15

    This paper describes the conceptual design of a ceramic breeder blanket considered as one of the candidate blankets in the first phase of the ARIES-CS study. The blanket is coupled to a Brayton power cycle to avoid the safety concern associated with the possibility of Be/steam reaction in case of accident.

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

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

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

  12. The ITER project construction status

    NASA Astrophysics Data System (ADS)

    Motojima, O.

    2015-10-01

    The pace of the ITER project in St Paul-lez-Durance, France is accelerating rapidly into its peak construction phase. With the completion of the B2 slab in August 2014, which will support about 400 000 metric tons of the tokamak complex structures and components, the construction is advancing on a daily basis. Magnet, vacuum vessel, cryostat, thermal shield, first wall and divertor structures are under construction or in prototype phase in the ITER member states of China, Europe, India, Japan, Korea, Russia, and the United States. Each of these member states has its own domestic agency (DA) to manage their procurements of components for ITER. Plant systems engineering is being transformed to fully integrate the tokamak and its auxiliary systems in preparation for the assembly and operations phase. CODAC, diagnostics, and the three main heating and current drive systems are also progressing, including the construction of the neutral beam test facility building in Padua, Italy. The conceptual design of the Chinese test blanket module system for ITER has been completed and those of the EU are well under way. Significant progress has been made addressing several outstanding physics issues including disruption load characterization, prediction, avoidance, and mitigation, first wall and divertor shaping, edge pedestal and SOL plasma stability, fuelling and plasma behaviour during confinement transients and W impurity transport. Further development of the ITER Research Plan has included a definition of the required plant configuration for 1st plasma and subsequent phases of ITER operation as well as the major plasma commissioning activities and the needs of the accompanying R&D program to ITER construction by the ITER parties.

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

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

  14. U.S. Contributions to ITER

    SciTech Connect

    Ned R. Sauthoff

    2005-05-13

    The United States participates in the ITER project and program to enable the study of the science and technology of burning plasmas, a key programmatic element missing from the world fusion program. The 2003 U.S. decision to enter the ITER negotiations followed an extensive series of community and governmental reviews of the benefits, readiness, and approaches to the study of burning plasmas. This paper describes both the technical and the organizational preparations and plans for U.S. participation in the ITER construction activity: in-kind contributions, staff contributions, and cash contributions as well as supporting physics and technology research. Near-term technical activities focus on the completion of R&D and design and mitigation of risks in the areas of the central solenoid magnet, shield/blanket, diagnostics, ion cyclotron system, electron cyclotron system, pellet fueling system, vacuum system, tritium processing system, and conventional systems. Outside the project, the U .S. is engaged in preparations for the test blanket module program. Organizational activities focus on preparations of the project management arrangements to maximize the overall success of the ITER Project; elements include refinement of U.S. directions on the international arrangements, the establishment of the U.S. Domestic Agency, progress along the path of the U.S. Department of Energy's Project Management Order, and overall preparations for commencement of the fabrication of major items of equipment and for provision of staff and cash as specified in the upcoming ITER agreement.

  15. Packed fluidized bed blanket for fusion reactor

    DOEpatents

    Chi, John W. H.

    1984-01-01

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

  16. Toughened Thermal Blanket for MMOD Protection

    NASA Technical Reports Server (NTRS)

    Christiansen, Eric L.; Lear, Dana M.

    2014-01-01

    Thermal blankets are used extensively on spacecraft to provide passive thermal control of spacecraft hardware from thermal extremes encountered in space. Toughened thermal blankets have been developed that greatly improve protection from hypervelocity micrometeoroid and orbital debris (MMOD) impacts. These blankets can be outfitted if so desired with a reliable means to determine the location, depth and extent of MMOD impact damage by incorporating an impact sensitive piezoelectric film. Improved MMOD protection of thermal blankets was obtained by adding selective materials at various locations within the thermal blanket. As given in Figure 1, three types of materials were added to the thermal blanket to enhance its MMOD performance: (1) disrupter layers, near the outside of the blanket to improve breakup of the projectile, (2) standoff layers, in the middle of the blanket to provide an area or gap that the broken-up projectile can expand, and (3) stopper layers, near the back of the blanket where the projectile debris is captured and stopped. The best suited materials for these different layers vary. Density and thickness is important for the disrupter layer (higher densities generally result in better projectile breakup), whereas a highstrength to weight ratio is useful for the stopper layer, to improve the slowing and capture of debris particles.

  17. Energetic ions in ITER plasmas

    SciTech Connect

    Pinches, S. D.; Chapman, I. T.; Sharapov, S. E.; Lauber, Ph. W.; Oliver, H. J. C.; Shinohara, K.; Tani, K.

    2015-02-15

    This paper discusses the behaviour and consequences of the expected populations of energetic ions in ITER plasmas. It begins with a careful analytic and numerical consideration of the stability of Alfvén Eigenmodes in the ITER 15 MA baseline scenario. The stability threshold is determined by balancing the energetic ion drive against the dominant damping mechanisms and it is found that only in the outer half of the plasma (r/a>0.5) can the fast ions overcome the thermal ion Landau damping. This is in spite of the reduced numbers of alpha-particles and beam ions in this region but means that any Alfvén Eigenmode-induced redistribution is not expected to influence the fusion burn process. The influence of energetic ions upon the main global MHD phenomena expected in ITER's primary operating scenarios, including sawteeth, neoclassical tearing modes and Resistive Wall Modes, is also reviewed. Fast ion losses due to the non-axisymmetric fields arising from the finite number of toroidal field coils, the inclusion of ferromagnetic inserts, the presence of test blanket modules containing ferromagnetic material, and the fields created by the Edge Localised Mode (ELM) control coils in ITER are discussed. The greatest losses and associated heat loads onto the plasma facing components arise due to the use of the ELM control coils and come from neutral beam ions that are ionised in the plasma edge.

  18. Tritium module for ITER/Tiber system code

    SciTech Connect

    Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

    1988-01-01

    A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs.

  19. Psychosomatic effects of blanketing in nursing care.

    PubMed

    Tokunaga, Namiji; Fukai, Kiyoko

    2017-04-01

    Blanketing, which involves covering patients with a blanket in order to maintain their body temperature and privacy and help alleviate the stress that is associated with medical procedures, is a fundamental skill in nursing care. However, the actual therapeutic effects of blanketing have not been adequately investigated. Therefore, the aim of the present study was to clarify the psychosomatic effects of blanketing in nursing care by observing autonomic nervous activity. This single-cohort, intra-participant, comparative cross-over study included 30 healthy female adolescents. Two of the 30 participants were excluded from the analysis. After splitting the participants into blanketing and non-blanketing groups, the stress and anxiety that patients undergo during nursing care was simulated and the responses were analyzed by using psychological and physiological indicators. The parametric data were analyzed by using a paired t-test or one-way ANOVA and the non-parametric data were analyzed by using the Wilcoxon signed-rank test. The significance level was <5%. No significant difference was seen between the two groups in vital signs, salivary alpha-amylase activity, respiratory movement or skin surface temperature. However, the postintervention scores for "Tension-Anxiety" in the Profile of Mood States - Brief Form were significantly lower for the blanketed than the non-blanketed arm, "sense of ease" and "feeling protected" were significantly higher for the blanketed arm, and in the middle intervention segment, the parasympathetic nervous activity and sympathetic nervous activity were significantly higher and lower, respectively, for the blanketed arm. These results suggest that blanketing is effective in relieving psychological stress and promoting a sense of ease among patients who are receiving nursing care. © 2016 Japan Academy of Nursing Science.

  20. RELAP5 Model of the First Wall/Blanket Primary Heat Transfer System

    SciTech Connect

    Popov, Emilian L; Yoder Jr, Graydon L; Kim, Seokho H

    2010-06-01

    ITER inductive power operation is modeled and simulated using a system level computer code to evaluate the behavior of the Primary Heat Transfer System (PHTS) and predict parameter operational ranges. The control algorithm strategy and derivation are summarized in this report as well. A major feature of ITER is pulsed operation. The plasma does not burn continuously, but the power is pulsed with large periods of zero power between pulses. This feature requires active temperature control to maintain a constant blanket inlet temperature and requires accommodation of coolant thermal expansion during the pulse. In view of the transient nature of the power (plasma) operation state a transient system thermal-hydraulics code was selected: RELAP5. The code has a well-documented history for nuclear reactor transient analyses, it has been benchmarked against numerous experiments, and a large user database of commonly accepted modeling practices exists. The process of heat deposition and transfer in the blanket modules is multi-dimensional and cannot be accurately captured by a one-dimensional code such as RELAP5. To resolve this, a separate CFD calculation of blanket thermal power evolution was performed using the 3-D SC/Tetra thermofluid code. A 1D-3D co-simulation more realistically models FW/blanket internal time-dependent thermal inertia while eliminating uncertainties in the time constant assumed in a 1-D system code. Blanket water outlet temperature and heat release histories for any given ITER pulse operation scenario are calculated. These results provide the basis for developing time dependent power forcing functions which are used as input in the RELAP5 calculations.

  1. Neutronics analyses of tritium breeding blanket performance in a spherical torus based volumetric neutron source

    SciTech Connect

    Cerbone, R. J.; Cheng, E.T.; Peng, Yueng Kay Martin

    1998-01-01

    A spherical torus based volumetric neutron source (ST-VNS) concept has been developed in recent studies as a possible intermediate step to develop the necessary technology for reactor components of future fusion power plants. Such a VNS would complement ITER in testing, developing, and qualifying nuclear technology components. A recently developed design concept for a spherical torus based VNS permits the development capability to increase fusion power and wall loading. Results of neutronics calculations for such a ST-VNS with neutron wall loading ranging from 0.5 to 5 MW/m(2) have been competed. In this paper, we report the tritium breeding and neutronics performance of several power blanket compositions and configurations. These include, a helium cooled natural lithium in vanadium alloy structural material blanket; a helium cooled enriched lithium-lead, in a vanadium-alloy structure blanket; and a heterogeneous configured blanket consisting of a dual cooled blanket consisting of enriched lithium-lead enclosed in silicon carbide with ferritic steel for the structural material.

  2. GCFR radial blanket and shield experiment

    SciTech Connect

    Muckenthaler, F.J.; Hull, J.L.; Manning, J.J.

    1980-12-01

    This report presents integral neutron flux, energy spectra, and gamma-ray heating measurements made for the Radial Blanket and Shield Experiment at the ORNL Tower Shielding Facility as part of a continuing Gas Cooled Fast Breeder Reactor program. The experimental configurations were divided into four basic segments: a spectrum modifier inserted into the Tower Shielding Reactor II beam; blanket slabs consisting of either ThO/sub 2/ or UO/sub 2/ placed directly behind the spectrum modifier; an inner radial shield behind the blankets; and an outer radial shield to complete the mockup. The segments were added in sequence, with selected measurements made within and beyond each segment. The integral experiment was performed to provide verification of calculational methods and nuclear data used in designing a radial shield for the GCFR and determining the effectiveness of the design. The ThO/sub 2/ blanket measurements were needed to bracket the uncertainties in the nuclear cross sections for calculating both the neutron transmission through the blanket and the gamma-ray heating rates within the blanket. Measurements with a UO/sub 2/ blanket were included both as a reference for the ThO/sub 2/ analysis, neutron transmission through UO/sub 2/ having been successfully calculated in previous experiments, and to provide comparison information for other breeder reactor designs.

  3. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    SciTech Connect

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

    2002-04-01

    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.

  4. The ITER in-vessel system

    SciTech Connect

    Lousteau, D.C.

    1994-09-01

    The overall programmatic objective, as defined in the ITER Engineering Design Activities (EDA) Agreement, is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER EDA Phase, due to last until July 1998, will encompass the design of the device and its auxiliary systems and facilities, including the preparation of engineering drawings. The EDA also incorporates validating research and development (R&D) work, including the development and testing of key components. The purpose of this paper is to review the status of the design, as it has been developed so far, emphasizing the design and integration of those components contained within the vacuum vessel of the ITER device. The components included in the in-vessel systems are divertor and first wall; blanket and shield; plasma heating, fueling, and vacuum pumping equipment; and remote handling equipment.

  5. Rotation and neoclassical ripple transport in ITER

    DOE PAGES

    Paul, Elizabeth Joy; Landreman, Matt; Poli, Francesca M.; ...

    2017-07-13

    Neoclassical transport in the presence of non-axisymmetric magnetic fields causes a toroidal torque known as neoclassical toroidal viscosity (NTV). The toroidal symmetry of ITER will be broken by the finite number of toroidal field coils and by test blanket modules (TBMs). The addition of ferritic inserts (FIs) will decrease the magnitude of the toroidal field ripple. 3D magnetic equilibria in the presence of toroidal field ripple and ferromagnetic structures are calculated for an ITER steady-state scenario using the Variational Moments Equilibrium Code (VMEC). Furthermore, neoclassical transport quantities in the presence of these error fields are calculated using the Stellarator Fokker-Planckmore » Iterative Neoclassical Conservative Solver (SFINCS).« less

  6. Status of EC solid breeder blanket designs and R&D for DEMO fusion reactors

    SciTech Connect

    Dalle Donne, M.; Anziedi, L.A.; Kwast, H.

    1994-12-31

    In the framework of the European Community Fusion Technology Program four blanket concepts for a DEMO reactor are being investigated. DEMO is the next step after ITER. It should ensure tritium self-sufficiency and operate at coolant temperatures high enough to have a reasonable plant efficiency. Further requirements have been specified for the four concepts, namely an average neutron wall load of 2.2 MW/m{sup 2}, a blanket lifetime of 20000 hours and the capability of the blanket segment to withstand the forces caused by a rapid distribution of the plasma current (20 MA to zero in 20 ms), so that after the disruption the segment can still allow a comparison of the various options, in view of reducing this number to two in 1995 and to design and develop modules and articles representative of the chosen blankets to be tested in ITER. The present paper deals with two solid breeder concepts. They have many features in common: both use high pressure helium as coolant and helium to purge the tritium from the breeder material, martensitic steel as structural material and beryllium as neutron multiplier. The configuration of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder material is LiAlO{sub 2} or LiZrO{sub 3} in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li{sub 4}SiO{sub 4} and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium.

  7. Method of fabricating a multilayer insulation blanket

    DOEpatents

    Gonczy, J.D.; Niemann, R.C.; Boroski, W.N.

    1993-07-06

    An improved multilayer insulation blanket for insulating cryogenic structures operating at very low temperatures is disclosed. An apparatus and method for fabricating the improved blanket are also disclosed. In the improved blanket, each successive layer of insulating material is greater in length and width than the preceding layer so as to accommodate thermal contraction of the layers closest to the cryogenic structure. The fabricating apparatus has a rotatable cylindrical mandrel having an outer surface of fixed radius that is substantially arcuate, preferably convex, in cross-section. The method of fabricating the improved blanket comprises (a) winding a continuous sheet of thermally reflective material around the circumference of the mandrel to form multiple layers, (b) binding the layers along two lines substantially parallel to the edges of the circumference of the mandrel, (c) cutting the layers along a line parallel to the axle of the mandrel, and (d) removing the bound layers from the mandrel.

  8. Method of fabricating a multilayer insulation blanket

    DOEpatents

    Gonczy, John D.; Niemann, Ralph C.; Boroski, William N.

    1993-01-01

    An improved multilayer insulation blanket for insulating cryogenic structures operating at very low temperatures is disclosed. An apparatus and method for fabricating the improved blanket are also disclosed. In the improved blanket, each successive layer of insulating material is greater in length and width than the preceding layer so as to accommodate thermal contraction of the layers closest to the cryogenic structure. The fabricating apparatus has a rotatable cylindrical mandrel having an outer surface of fixed radius that is substantially arcuate, preferably convex, in cross-section. The method of fabricating the improved blanket comprises (a) winding a continuous sheet of thermally reflective material around the circumference of the mandrel to form multiple layers, (b) binding the layers along two lines substantially parallel to the edges of the circumference of the mandrel, (c) cutting the layers along a line parallel to the axle of the mandrel, and (d) removing the bound layers from the mandrel.

  9. Multilayer insulation blanket, fabricating apparatus and method

    DOEpatents

    Gonczy, J.D.; Niemann, R.C.; Boroski, W.N.

    1992-09-01

    An improved multilayer insulation blanket for insulating cryogenic structures operating at very low temperatures is disclosed. An apparatus and method for fabricating the improved blanket are also disclosed. In the improved blanket, each successive layer of insulating material is greater in length and width than the preceding layer so as to accommodate thermal contraction of the layers closest to the cryogenic structure. The fabricating apparatus has a rotatable cylindrical mandrel having an outer surface of fixed radius that is substantially arcuate, preferably convex, in cross-section. The method of fabricating the improved blanket comprises (a) winding a continuous sheet of thermally reflective material around the circumference of the mandrel to form multiple layers, (b) binding the layers along two lines substantially parallel to the edges of the circumference of the mandrel, (c) cutting the layers along a line parallel to the axle of the mandrel, and (d) removing the bound layers from the mandrel. 7 figs.

  10. Multilayer insulation blanket, fabricating apparatus and method

    DOEpatents

    Gonczy, John D.; Niemann, Ralph C.; Boroski, William N.

    1992-01-01

    An improved multilayer insulation blanket for insulating cryogenic structures operating at very low temperatures is disclosed. An apparatus and method for fabricating the improved blanket are also disclosed. In the improved blanket, each successive layer of insulating material is greater in length and width than the preceding layer so as to accommodate thermal contraction of the layers closest to the cryogenic structure. The fabricating apparatus has a rotatable cylindrical mandrel having an outer surface of fixed radius that is substantially arcuate, preferably convex, in cross-section. The method of fabricating the improved blanket comprises (a) winding a continuous sheet of thermally reflective material around the circumference of the mandrel to form multiple layers, (b) binding the layers along two lines substantially parallel to the edges of the circumference of the mandrel, (c) cutting the layers along a line parallel to the axle of the mandrel, and (d) removing the bound layers from the mandrel.

  11. APT target-blanket fabrication development

    SciTech Connect

    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.

  12. Multifractal Framework Based on Blanket Method

    PubMed Central

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

    2014-01-01

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

  13. 49 CFR 366.5 - Blanket designations.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... FMCSA a list of process agents for each State and DC (blanket agent), motor carriers, brokers and... Transportation Other Regulations Relating to Transportation (Continued) FEDERAL MOTOR CARRIER SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION FEDERAL MOTOR CARRIER SAFETY REGULATIONS DESIGNATION OF PROCESS...

  14. Silver Teflon blanket: LDEF tray C-08

    NASA Technical Reports Server (NTRS)

    Crutcher, E. Russ; Nishimura, L. S.; Warner, K. J.; Wascher, W. W.

    1992-01-01

    A study of the Teflon blanket surface at the edge of tray C-08 illustrates the complexity of the microenvironments on the Long Duration Exposure Facility (LDEF). The distribution of particulate contaminants varied dramatically over a distance of half a centimeter (quarter of an inch) near the edge of the blanket. The geometry and optical effects of the atomic oxygen erosion varied significantly over the few centimeters where the blanket folded over the edge of the tray resulting in a variety of orientations to the atomic oxygen flux. A very complex region of combined mechanical and atomic oxygen damage occurred where the blanket contacted the edge of the tray. A brown film deposit apparently fixed by ultraviolet light traveling by reflection through the Teflon film was conspicuous beyond the tray contract zone. Chemical and structural analysis of the surface of the brown film and beyond toward the protected edge of the blanket indicated some penetration of energetic atomic oxygen at least five millimeters past the blanket-tray contact interface.

  15. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    NASA Astrophysics Data System (ADS)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

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

    SciTech Connect

    Dumortier, Pierre; Louche, Fabrice; Messiaen, André; Vervier, Michel

    2014-02-12

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

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

    SciTech Connect

    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

    1996-12-31

    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{sup 3} to 10{sup 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.

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

    SciTech Connect

    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, Z.Y.

    1996-12-31

    The Vanadium/Lithium system has been the recent focus of ANL`s Blanket Technology Program, and for the last several years, ANL`s Liquid Metal Blanket activities have been carried out in direct support of the ITER (International Thermonuclear Experimental Reactor) breeding blanket task area. A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the development of insulator coatings. Design calculations, Hua and Gohar, show that an electrically insulating layer is necessary to maintain an acceptably low magnetohydrodynamic (MHD) pressure drop in the current ITER design. Consequently, the decision was made to convert Argonne`s Liquid Metal EXperiment (ALEX) from a 200{degree}C NaK facility to a 350{degree}C lithium facility. The upgraded facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups at Hartmann numbers (M) and interaction parameters (N) in the range of 10{sup 3} to 10{sup 5} in lithium at 350{degree}C. Following completion of the upgrade work, a short performance test was conducted, followed by two longer, multiple-hour, MHD tests, all at 230{degree}C. The modified ALEX facility performed up to expectations in the testing. MHD pressure drop and test section voltage distributions were collected at Hartmann numbers of 1000. 4 refs., 2 figs.

  20. Space or rescue blanket--a bluff?

    PubMed

    Renström, B

    1992-10-01

    A product which has been heavily promoted for a number of years is an aluminized plastic foil designed to reflect 80% of body heat. However, a large body of experience points to the following: 1. A minimal amount of heat is radiated by a clothed body. This can easily be demonstrated with a thermocamera. 2. The plastic foils is impermeable to both water and water vapour. Thus condensation is formed on the blanket's interior. This virtually eliminates the reflective properties of the aluminum foil, often after only 15-20 minutes. 3. The blanket usually works best when the wind velocity is less than 9 m/s, otherwise the wind tears it into pieces. The greater the wind velocity and amount of clothing worn, the less the effect of the blanket. 4. Water accumulation on the clothing inside the blanket rapidly conducts heat from the skin. The blanket and the aluminum then take over and conduct heat 9,300 times faster than air. 5. It is supposed to reflect radar waves. Theoretically yes, but practical tests made in lifeboats at sea have showed that it gives no help. 6. The plastic film is inflammable.

  1. Thermally distinct ejecta blankets from Martian craters

    NASA Astrophysics Data System (ADS)

    Betts, B. H.; Murray, B. C.

    1993-06-01

    A study of Martian ejecta blankets is carried out using the high-resolution thermal IR/visible data from the Termoskan instrument aboard Phobos '88 mission. It is found that approximately 100 craters within the Termoskan data have an ejecta blanket distinct in the thermal infrared (EDITH). These features are examined by (1) a systematic examination of all Termoskan data using high-resolution image processing; (2) a study of the systematics of the data by compiling and analyzing a data base consisting of geographic, geologic, and mormphologic parameters for a significant fraction of the EDITH and nearby non-EDITH; and (3) qualitative and quantitative analyses of localized regions of interest. It is noted that thermally distinct ejecta blankets are excellent locations for future landers and remote sensing because of relatively dust-free surface exposures of material excavated from depth.

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

    SciTech Connect

    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.

  3. Insulation Blankets for High-Temperature Use

    NASA Technical Reports Server (NTRS)

    Goldstein, H.; Leiser, D.; Sawko, P. M.; Larson, H. K.; Estrella, C.; Smith, M.; Pitoniak, F. J.

    1986-01-01

    Insulating blanket resists temperatures up to 1,500 degrees F (815 degrees C). Useful where high-temperature resistance, flexibility, and ease of installation are important - for example, insulation for odd-shaped furnaces and high-temperature ducts, curtains for furnace openings and fire control, and conveyor belts in hot processes. Blanket is quilted composite consisting of two face sheets: outer one of silica, inner one of silica or other glass cloth with center filling of pure silica glass felt sewn together with silica glass threads.

  4. Lightweight IMM PV Flexible Blanket Assembly

    NASA Technical Reports Server (NTRS)

    Spence, Brian

    2015-01-01

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

  5. Some new ideas for Tandem Mirror blankets

    SciTech Connect

    Neef, W.S. Jr.

    1981-10-12

    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.

  6. Plasma vertical stabilisation in ITER

    NASA Astrophysics Data System (ADS)

    Gribov, Y.; Kavin, A.; Lukash, V.; Khayrutdinov, R.; Huijsmans, G. T. A.; Loarte, A.; Snipes, J. A.; Zabeo, L.

    2015-07-01

    This paper describes the progress in analysis of the ITER plasma vertical stabilisation (VS) system since its design review in 2007-2008. Two indices characterising plasma VS were studied. These are (1) the maximum value of plasma vertical displacement due to free drift that can be stopped by the VS system and (2) the maximum root mean square value of low frequency noise in the dZ/dt measurement signal used in the VS feedback loop. The first VS index was calculated using the PET code for 15 MA plasmas with the nominal position and shape. The second VS index was studied with the DINA code in the most demanding simulations for plasma magnetic control of 15 MA scenarios with the fastest plasma current ramp-up and early X-point formation, the fastest plasma current ramp-down in a divertor configuration, and an H to L mode transition at the current flattop. The studies performed demonstrate that the VS in-vessel coils, adopted recently in the baseline design, significantly increase the range of plasma controllability in comparison with the stabilising systems VS1 and VS2, providing operating margins sufficient to achieve ITER's goals specified in the project requirements. Additionally two sets of the DINA code simulations were performed with the goal of assessment of the capability of the PF system with the VS in-vessel coils: (i) to control the position of runaway electrons generated during disruptions in 15 MA scenarios and (ii) to trigger ELMs in H-mode plasmas of 7.5 MA/2.65 T scenarios planned for the early phase of ITER operation. It was also shown that ferromagnetic structures of the vacuum vessel (ferromagnetic inserts) and test blanket modules insignificantly affect the plasma VS.

  7. Flute stabilization by a cold line-tied blanket

    SciTech Connect

    Segal, D.; Wickham, M.; Rynn, N.

    1982-09-01

    The curvature-driven flute instability in an axisymmetric mirror was stabilized by an annular line-tied plasma blanket. A significant temperature difference was maintained between core and blanket. Theoretical calculations support the experimental observations.

  8. 48 CFR 1313.303 - Blanket Purchase Agreements (BPAs).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  9. 48 CFR 13.303 - Blanket purchase agreements (BPAs).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  10. 48 CFR 613.303 - Blanket purchase agreements (BPAs).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  11. Lithium Ceramic Blankets for Russian Fusion Reactors and Influence of Breeding Operation Mode on Parameters of Reactor Tritium Systems

    SciTech Connect

    Kapyshev, Victor K.; Chernetsov, Mikhail Yu.; Zhevotov, Sergej I.; Kersnovskij, Alexandr Yu.; Kolbasov, Boris N.; Kovalenko, Victor G.; Paltusov, Nikolaj P.; Sernyaev, Georgeij A.; Sterebkov, Juri S.; Zyryanov, Alexej P.

    2005-07-15

    Russian controlled fusion program supposes development of a DEMO reactor design and participation in ITER Project. A solid breeder blanket of DEMO contains a ceramic lithium orthosilicate breeder and a beryllium multiplier. Test modules of the blanket are developed within the scope of ITER activities. Experimental models of module tritium breeding zones (TBZ), materials and fabrication technology of the TBZ, tritium reactor systems to analyse and process gas released from lithium ceramics are being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ have been designed, manufactured and tested in IVV-2M nuclear reactor. Initial results of the in-pile experiments and outcome of lithium ceramics irradiation in a water-graphite nuclear reactor are considered to be a data base for development of the test modules and initial requirements for DEMO tritium system design. Influence of the tritium release parameters and hydrogen concentration in a purge gas on parameters of reactor system are discussed.

  12. 75 FR 51482 - Woven Electric Blankets From China

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-20

    ... COMMISSION Woven Electric Blankets From China Determination On the basis of the record \\1\\ developed in the... United States is materially injured by reason of imports from China of woven electric blankets, provided... woven electric blankets from China were being sold at LTFV within the meaning of section 733(b) of the...

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

    SciTech Connect

    Not Available

    1993-08-04

    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.

  14. Aerogel Blanket Insulation Materials for Cryogenic Applications

    NASA Technical Reports Server (NTRS)

    Coffman, B. E.; Fesmire, J. E.; White, S.; Gould, G.; Augustynowicz, S.

    2009-01-01

    Aerogel blanket materials for use in thermal insulation systems are now commercially available and implemented by industry. Prototype aerogel blanket materials were presented at the Cryogenic Engineering Conference in 1997 and by 2004 had progressed to full commercial production by Aspen Aerogels. Today, this new technology material is providing superior energy efficiencies and enabling new design approaches for more cost effective cryogenic systems. Aerogel processing technology and methods are continuing to improve, offering a tailor-able array of product formulations for many different thermal and environmental requirements. Many different varieties and combinations of aerogel blankets have been characterized using insulation test cryostats at the Cryogenics Test Laboratory of NASA Kennedy Space Center. Detailed thermal conductivity data for a select group of materials are presented for engineering use. Heat transfer evaluations for the entire vacuum pressure range, including ambient conditions, are given. Examples of current cryogenic applications of aerogel blanket insulation are also given. KEYWORDS: Cryogenic tanks, thermal insulation, composite materials, aerogel, thermal conductivity, liquid nitrogen boil-off

  15. Drugs: blanket ban or harm reduction?

    PubMed

    Gross, Michael

    2015-06-29

    As the commercial success of electronic cigarettes offers the opportunity to study harm reduction by replacement therapy, the UK government steers in the opposite direction and plans to expand the 'war on drugs' with a blanket ban on all psychoactive substances. Michael Gross reports.

  16. The climatic impact of supervolcanic ash blankets

    NASA Astrophysics Data System (ADS)

    Jones, Morgan T.; Sparks, R. Stephen J.; Valdes, Paul J.

    2007-11-01

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

  17. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    .... Where a manufacturer of tobacco products operates more than one factory in the same region he may, in... provisions of § 40.134, for any or all of the factories in the same region. The total amount of any blanket... factory covered by the bond. (72 Stat. 1421; 26 U.S.C. 5711) ...

  18. 18 CFR 157.203 - Blanket certification.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 18 Conservation of Power and Water Resources 1 2013-04-01 2013-04-01 false Blanket certification. 157.203 Section 157.203 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY COMMISSION... AND NECESSITY AND FOR ORDERS PERMITTING AND APPROVING ABANDONMENT UNDER SECTION 7 OF THE NATURAL GAS...

  19. 18 CFR 157.203 - Blanket certification.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 18 Conservation of Power and Water Resources 1 2012-04-01 2012-04-01 false Blanket certification. 157.203 Section 157.203 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY COMMISSION... AND NECESSITY AND FOR ORDERS PERMITTING AND APPROVING ABANDONMENT UNDER SECTION 7 OF THE NATURAL GAS...

  20. The climatic impact of supervolcanic ash blankets

    NASA Astrophysics Data System (ADS)

    Jones, M. T.; Sparks, S. J.; Valdes, P. J.

    2006-12-01

    Supervolcanoes are capable of ejecting 1000's of cubic kilometres of magmatic material in a single eruption, far surpassing anything recorded in human history. It has been postulated that these eruptions have acted as catalysts for long-term climate change and are responsible for bottlenecks in human and animal populations. Tephra deposits from a super-eruption are capable of covering an area the size of USA (~10,000,000 sq. km) with ash, destroying vegetation and considerably raising the surface albedo. Ecological responses to smaller eruptions show that recovery of flora takes over 15 years, while previous studies of ash blankets demonstrate sustained surface residence times. This suggests that a supervolcanic ash blanket would instigate a decadal climate response that would dominate in the aftermath of the effects of aerosols in the stratosphere. We use a coupled atmosphere-ocean General Circulation Model (GCM) to simulate the effect of an ash blanket from Yellowstone volcano, USA, and show that it causes major disruptions to the climate, particularly to oscillatory systems such as the El Niño Southern Oscillation (ENSO). The regional disturbance instigates a global response, with significant variations in surface temperatures, pressures and precipitation patterns. The ocean remains largely unaffected, though a marked increase in sea ice is seen in the North Atlantic. While the response to a supervolcanic ash blanket is predicted to be severe, the isolated effects of the disturbance are not significant enough to instigate long-term climate change at present day boundary conditions.

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

    SciTech Connect

    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.

  2. Localized Fast-Ion Induced Heat Loads in Test Blanket Module Mockup Experiments on DIII-D

    NASA Astrophysics Data System (ADS)

    Kramer, G. J.; Budny, R. V.; Ellis, R. A.; Nazikian, R.; McLean, A. G.; Brooks, N. H.; Schaffer, M. J.; van Zeeland, M. A.; Heidbrink, W. W.; Kurki-Suonio, T.; Koskela, T.; Shinohara, K.; Snipes, J. A.; Spong, D. A.

    2012-10-01

    Localized hot spots can be created in ITER on the Test Blanket Modules (TBMs) because the ferritic steel of the TBMs distorts the local magnetic field near the modules and alters fast ion confinement. Predicting the TBM heat load levels is important for assessing their effects on the ITER first wall. Experiments in DIII-D were carried out with a mock-up of the ITER TBM ferromagnetic error field to provide data for validation of fast-ion orbit following codes. The front surface temperature of the protective TBM tiles was imaged directly with a calibrated infrared camera and heat loads were extracted. The detailed spot sizes and measured heat loads are compared with results from heat load calculations performed with a suite of orbit following codes. The codes reproduce the hot spots well, thereby validating the codes and giving confidence in predictions for fast-ion heat loads in ITER.

  3. Nuclear systems and testing programs for ITER. Progress report for FY 1998

    SciTech Connect

    1998-12-31

    The effort during this performance period focused on a number of TBWG activities (including test module design and analysis) that were identified and agreed upon (in the presence of the ITER Director and Deputy Director) at TBWG-4. These include: (a) DEMO test module design and performance analysis under pulsed operation; (b) Test program operation plan; (c) Test port design and analysis; (d) Decay heat calculations and safety analysis; (e) Further discussion among the parties to define collaboratory on R and D for the test program as well as possible collaboration on the construction and operation of test articles; (f) Remote handling and ancillary equipment; (g) Criteria for qualifying a blanket module or submodule for actual insertion and testing in ITER; (h) Definition of test module instrumentation and verification of capability to perform in the ITER fusion environment (magnetic field, radiation, heating, etc.); and (i) Analysis to show that the results to be obtained from the test modules as designed can be extrapolated to DEMO and reactor conditions (e.g., higher wall loads and the need to demonstrate tritium self-sufficiency). The main achievements during this performance period include: (1) updating and finalizing the US DDDs for the ITER Blanket Program to form part of the ITER Final Design Report (FDR). Specific revisions were in response to the minimal lithium volume test blanket design requirements and safety impact and (2) evaluating the feasibility of the US test program, including instrumentation and the benefits of the ITER test program. Details of this assessment, including solid breeder and liquid breeder blanket test plans, are documented in UCLA-IFNT-13 (attached). In addition, dose mapping calculations were performed for the ITER Building, including equipment and layout of coolant pipes/heat exchangers. A report on ITER Building dose calculations was sent to UD ITER management and to the Garching Task Coordinator in April, 1998. The report

  4. A Precambrian proximal ejecta blanket from Scotland

    NASA Astrophysics Data System (ADS)

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

    2008-04-01

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

  5. Recent designs for advanced fusion reactor blankets

    SciTech Connect

    Sze, D.K.

    1994-11-01

    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{sub 2}ZrO{sub 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{sub 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-{sup 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.

  6. Detection of Breeding Blankets Using Antineutrinos

    NASA Astrophysics Data System (ADS)

    Cogswell, Bernadette; Huber, Patrick

    2016-03-01

    The Plutonium Management and Disposition Agreement between the United States and Russia makes arrangements for the disposal of 34 metric tons of excess weapon-grade plutonium. Under this agreement Russia plans to dispose of its excess stocks by processing the plutonium into fuel for fast breeder reactors. To meet the disposition requirements this fuel would be burned while the fast reactors are run as burners, i.e., without a natural uranium blanket that can be used to breed plutonium surrounding the core. This talk discusses the potential application of antineutrino monitoring to the verification of the presence or absence of a breeding blanket. It is found that a 36 kg antineutrino detector, exploiting coherent elastic neutrino-nucleus scattering and made of silicon, could determine the presence of a breeding blanket at a liquid sodium cooled fast reactor at the 95% confidence level within 90 days. Such a detector would be a novel non-intrusive verification tool and could present a first application of coherent elastic neutrino-nucleus scattering to a real-world challenge.

  7. Ceramic pebble bed development for fusion blankets

    SciTech Connect

    Gierszewski, P.; Kawamura, H.; Donne, M.D.

    1994-12-31

    Research on lithium ceramic breeders has been intensive since the late 1970`s. The bulk material properties of several candidate lithium ceramics are generally available, although there is still much work to be done on properties under irradiation and on overall behavior in blanket modules. Based on these results, lithium ceramic breeders have been selected in many fusion design studies. These lithium ceramics are incorporated into blankets typically as monolithic pellets of packed pebble beds. There is substantial industrial experience with pebble beds made from other ceramics, notably in chemical processes as catalyst supports and grinding media, and in advanced fission reactor fuels. In fusion blankets, the pebble bed form offers several attractive features, including simpler assembly into complex geometry, uniform pore network, and low sensitivity to cracking or irradiation damage. Ceramic breeder pebbles have been a focus for several research groups. In general, the database is similar to that of monolithic pellets for the materials studied: basic production and material property data are available, but the irradiation and engineering database remains sparse.

  8. Design of the APT Target/Blanket

    NASA Astrophysics Data System (ADS)

    Cappiello, M. W.

    1998-04-01

    The Accelerator Production of Tritium Target/Blanket system is composed of a separated tungsten spallation target surrounded by a lead moderator, as well as attendant heat removal systems. The system is housed in a building located at the end of a 1.3 km long linear accelerator, which can produce a 100 mA proton beam up to 1700 MeV (170MW). The beam is expanded by a rastering system to a 0.19m x 190.m shape before passing through an Inconel window and impacting the heavy-water cooled tungsten target. Neutrons produced in the tungsten by the spallation process are further multiplied and moderated in a surrounding light-water cooled lead blanket. Neutron capture in tubes of Helium-3 gas inserted in the blanket produce tritium which is removed on a continual basis in an adjacent Tritium Separation Facility (TSF). The APT T/B is a robust design based on existing technology. Where possible, proven materials and component designs are used. To accommodate uncertainties in predicted lifetimes, the design is modularized to allow for a straightforward replacement of spent components. The thermal hydraulic design is well within allowable limits and due to the low temperature and pressure systems, offers additional safety and reliability benefits. The safety by design process has incorporated passive design features, redundancy, and defense in depth to provide adequate protection of both the worker and the public.

  9. US ITER Moving Forward

    ScienceCinema

    US ITER / ORNL

    2016-07-12

    US ITER Project Manager Ned Sauthoff, joined by Wayne Reiersen, Team Leader Magnet Systems, and Jan Berry, Team Leader Tokamak Cooling System, discuss the U.S.'s role in the ITER international collaboration.

  10. Iter and Ornl

    NASA Astrophysics Data System (ADS)

    Uckan, N. A.; Milora, S. L.

    2004-11-01

    ITER (means ``the way''), a tokamak burning plasma experiment, is the next step device toward making fusion energy a reality. The programmatic objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. ITER began in 1985 as collaboration between the Russian Federation (former Soviet Union), the USA, European Union, and Japan. ITER conceptual and engineering design activities led to a detailed design in 2001. The USA opted out of the project between 1999-2003, but rejoined in 2004 for site selection and construction negotiations. China and Korea joined the project in 2003. Negotiations are continuing and a decision on the site for ITER construction [France versus Japan] is pending. The ITER international undertaking is an unprecedented scale and the six ITER parties represent 40% of the world population. By 2018, ITER will produce a fusion power of 500 million Watts for time periods up to an hour with one-tenth of the power needed to sustain it. Steady state operation is also possible at lower power levels with higher fraction of circulated power. The ITER parties invested about $1 billion into the research and development (R) and related fusion experiments to establish the ITER's feasibility. ORNL has been a key player in the ITER project and contributed to its physics and engineering design and related R since its inception. Recently, the U.S. DOE selected the PPPL/ORNL partnership to lead the U.S. project office for ITER.

  11. Cost and performance projections for SPS photovoltaic blankets

    NASA Technical Reports Server (NTRS)

    Scott-Monck, J. A.

    1981-01-01

    An estimate, based on optimistic projections of current technology, is given for the specific power of photovoltaic blankets which might be achieved if the SPS concept was to be implemented. A simultaneous consideration of cost and technical requirements is used to identify key blanket technologies which must be developed for this reference system. The terrestrial photovoltaic experience coupled with new technology is used to develop cost estimates for the blanket, assuming an annual demand of 5 GW and a manufacturing industry dedicated to blanket production. The results indicate that blanket specific power goals may be exceeded, but there is little prospect that the cost goals can be met. This argues for a reconsideration of the photovoltaic option based on more expensive but higher performance blankets.

  12. Numerical modelling of dynamic sludge blanket behaviour in secondary clarifiers.

    PubMed

    Armbruster, M; Krebs, P; Rodi, W

    2001-01-01

    New developments in numerical modelling of turbulent and density-affected flow in secondary clarifiers are reported. The sludge blanket is included in the computation domain which allows us to account for sedimentation and resuspension of sludge as well as the growth and diminution of the sludge blanket and at the same time respecting mass conservation. It is shown how strongly the prediction of the sludge-blanket height depends on the approaches to describe the settling behaviour of the sludge and the rheological properties within the sludge blanket. Further, an example of dynamic simulation is presented and discussed. This demonstrates how the sludge blanket behaves during load variation and that instabilities may occur at the interface of sludge blanket and supernatant, potentially resulting in sludge wash-off during transient phases, which is not only during load increase but also during load decrease.

  13. Neutronic optimization of solid breeder blankets for STARFIRE design

    SciTech Connect

    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.

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

    SciTech Connect

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

    1999-07-01

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

  15. Nuclear modules of ITER tokamak systems code

    SciTech Connect

    Gohar, Y.; Baker, C.; Brooks, J.; Finn, P.; Hassanein, A.; Willms, S.; Barr, W.; Bushigin, A.; Kalyanam, K.M.; Haines, J.

    1987-10-01

    Nuclear modules were developed to model various reactor components in the ITER systems code. Several design options and cost algorithms are included for each component. The first wall, blanket and shield modules calculate the beryllium zone thickness, the disruptions results, the nuclear responses in different components including the toroidal field coils. Tungsten shield/water coolant/steel structure and steel shield/water coolant are the shield options for the inboard and outboard sections of the reactor. Lithium nitrate dissolved in the water coolant with a variable beryllium zone thickness in the outboard section of the reactor provides the tritium breeding capability. The reactor vault module defines the thickness of the reactor wall and the roof based on the dose equivalent during operation including skyshine contribution. The impurity control module provides the design parameters for the divertor including plate design, heat load, erosion rate, tritium permeation through the plate material to the coolant, plasma contamination by sputtered impurities, and plate lifetime. Several materials: Be, C, V, Mo, and W can be used for the divertor plate to cover a range of plasma edge temperatures. The tritium module calculates tritium and deuterium flow rates for the reactor plant. The tritium inventory in the fuelers, neutral beams, vacuum pumps, impurity control, first wall, and blanket is calculated. Tritium requirements are provided for different operating conditions. The nuclear models are summarized in this paper including the different design options and key analyses of each module. 39 refs., 3 tabs.

  16. Iterated fractional Tikhonov regularization

    NASA Astrophysics Data System (ADS)

    Bianchi, Davide; Buccini, Alessandro; Donatelli, Marco; Serra-Capizzano, Stefano

    2015-05-01

    Fractional Tikhonov regularization methods have been recently proposed to reduce the oversmoothing property of the Tikhonov regularization in standard form, in order to preserve the details of the approximated solution. Their regularization and convergence properties have been previously investigated showing that they are of optimal order. This paper provides saturation and converse results on their convergence rates. Using the same iterative refinement strategy of iterated Tikhonov regularization, new iterated fractional Tikhonov regularization methods are introduced. We show that these iterated methods are of optimal order and overcome the previous saturation results. Furthermore, nonstationary iterated fractional Tikhonov regularization methods are investigated, establishing their convergence rate under general conditions on the iteration parameters. Numerical results confirm the effectiveness of the proposed regularization iterations.

  17. Status of US ITER Diagnostics

    NASA Astrophysics Data System (ADS)

    Stratton, B.; Delgado-Aparicio, L.; Hill, K.; Johnson, D.; Pablant, N.; Barnsley, R.; Bertschinger, G.; de Bock, M. F. M.; Reichle, R.; Udintsev, V. S.; Watts, C.; Austin, M.; Phillips, P.; Beiersdorfer, P.; Biewer, T. M.; Hanson, G.; Klepper, C. C.; Carlstrom, T.; van Zeeland, M. A.; Brower, D.; Doyle, E.; Peebles, A.; Ellis, R.; Levinton, F.; Yuh, H.

    2013-10-01

    The US is providing 7 diagnostics to ITER: the Upper Visible/IR cameras, the Low Field Side Reflectometer, the Motional Stark Effect diagnostic, the Electron Cyclotron Emission diagnostic, the Toroidal Interferometer/Polarimeter, the Core Imaging X-Ray Spectrometer, and the Diagnostic Residual Gas Analyzer. The front-end components of these systems must operate with high reliability in conditions of long pulse operation, high neutron and gamma fluxes, very high neutron fluence, significant neutron heating (up to 7 MW/m3) , large radiant and charge exchange heat flux (0.35 MW/m2) , and high electromagnetic loads. Opportunities for repair and maintenance of these components will be limited. These conditions lead to significant challenges for the design of the diagnostics. Space constraints, provision of adequate radiation shielding, and development of repair and maintenance strategies are challenges for diagnostic integration into the port plugs that also affect diagnostic design. The current status of design of the US ITER diagnostics is presented and R&D needs are identified. Supported by DOE contracts DE-AC02-09CH11466 (PPPL) and DE-AC05-00OR22725 (UT-Battelle, LLC).

  18. High temperature - low mass solar blanket

    NASA Technical Reports Server (NTRS)

    Mesch, H. G.

    1979-01-01

    Interconnect materials and designs for use with ultrathin silicon solar cells are discussed, as well as the results of an investigation of the applicability of parallel-gap resistance welding for interconnecting these cells. Data relating contact pull strength and cell electrical degradation to variations in welding parameters such as time, voltage and pressure are presented. Methods for bonding ultrathin cells to flexible substances and for bonding thin (75 micrometers) covers to these cells are described. Also, factors influencing fabrication yield and approaches for increasing yield are discussed. The results of vacuum thermal cycling and thermal soak tests on prototype ultrathin cell test coupons and one solar module blanket are presented.

  19. Recent designs for advanced fusion reactor blankets

    SciTech Connect

    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.

  20. Isotope production in fast reactor blankets

    SciTech Connect

    Zvonarev, A.V.; Koloskov, B.V.; Kochetkov, L.A.

    1993-12-31

    At the BR-10 research reactor radioactive isotopes are produced that are required for the production of radiopharmaceuticals and phosphor 32 used for the synthesis of biochemical labelled compounds. A procedure has been developed of uranium target irradiation followed by radiochemical processing aimed at isotope isolation of molybdenum 99, xenon 133 and iodine 131,132,and 133 isotopes mixture. Irradiation is carried out in a special channel of the radial blanket. The production of cobalt 60 at the BN-600 reactor and facilities are also described.

  1. An overview of the US DCLL ITER-TBM program

    SciTech Connect

    Wong, C. P. C.; Abdou, M.; Dagher, M.; Katoh, Y.; Kurtz, R. J.; Malang, S.; Marriott, E. P.; Merrill, B. J.; Messadek, K.; Morley, N. B.; Sawan, M. E.; Sharafat, S.; Smolentsev, S.; Sze, D. K.; Willms, S.; Ying, A.; Youssef, M. Z.

    2010-12-01

    Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li (~650–700 °C) from the RAF/M structure, which has a corrosion temperature limit of ~480 °C. The RAF/M material must also operate at temperatures above 350 °C but less than 550 °C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. Finally, this paper is a summary report on the progress and results of recent DCLL TBM development activities.

  2. An Overview of the US DCLL ITER-TBM Program

    SciTech Connect

    C.P.C. Wong; M. Abdou; M. Dagher; Y. Katoh; B. J. Merrill

    2010-12-01

    Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li (650–700 °C) from the RAF/M structure, which has a corrosion temperature limit of 480 °C. The RAF/M material must also operate at temperatures above 350 °C but less than 550 °C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.

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

    SciTech Connect

    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.

  4. Recent activities related to the development of the plasma facing components for the ITER and fusion DEMO plant

    NASA Astrophysics Data System (ADS)

    Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji; Hirose, Takanori; Mori, Seiji; Enoeda, Mikio

    2009-12-01

    In the ITER procurement, the Japanese Domestic Agency (JADA) will procure the divertor outer vertical target (OVT). The qualification of the manufacture of the components has been started by validating JADA's technical capability. JADA has developed vertical target qualification prototypes that cover most of the critical technical issues in the series production. The prototypes have been high heat flux tested under effective coordination by the ITER organization (IO) and showed sufficient durability. JADA has successfully obtained the certification to conduct the procurement of the OVT by the IO. Development of a breeding blanket is one of the most important issues to realize the DEMO. Test blanket module (TBM) testing is a key milestone toward the DEMO. In JAEA, R&D on the water-cooled blanket has been performed. As a result, a full-scale TBM first wall mock-up has successfully been developed. This mock-up showed sound thermal performance in preliminary testing.

  5. Factors affecting the measurement accuracy of ITER neutron activation system

    NASA Astrophysics Data System (ADS)

    Cheon, M. S.; Ahn, Y. H.; Pak, S.; Seon, C.; Krasilnikov, V.; Bertalot, L.

    2017-08-01

    One of the main purposes of the ITER2 neutron activation system (NAS) is to evaluate the total neutron production rate from all over the plasma. The measurement accuracy depends on the position and profile of the plasma and the material in front of the irradiation end. It is required to minimize the amount of material and its density variation across the field of view between the plasma and the irradiation end. Due to the radiation and thermal environment of the ITER in-vessel, however, the measurement from ITER NAS cannot avoid the strong influence from in-vessel materials such as the diagnostic first wall, blanket modules, and divertor cassettes, those are located near the irradiation ends. In order to improve the reliability of the measurement in such environment, special cutouts in the diagnostic first wall are introduced near the irradiation end structures located in the port plugs. The effect of the materials and the position and profile of the neutron source in the plasma are evaluated for these irradiation locations, as well as the ones under the divertor cassettes and between blanket modules, by the neutron transport calculation. Calculation results show that simultaneous measurements at upper port and divertor location can provide highly accurate results even without a position or profile correction from other diagnostics.

  6. Development of blanket box structure fabrication technology

    SciTech Connect

    Mohri, K.; Sata, S.; Kawaguchi, I.

    1994-12-31

    Fabrication studies have been performed for first wall and blanket box structure in the Fusion Experimental Reactor designed in Japan. The first wall must have internal cooling channels to remove volumetric heat loading by neutron wall load and surface heat loading from the plasma. The blanket which is higher than 10 m and 1 m wide withstands enormous electromagnetic load (about 10 MN/m). And a fabrication accuracy is required in the order of 10 mm from the machine configuration and remote assembling standpoints. To make cooling channels inside the first wall and to reduce the deformation during fabrication, the authors adopted advance techniques Hot Isostatic Pressing method (HIP) and Electron Beam Welding (EBW) respectively. Evaluation studies for the bondability of the HIP bonding joint have been performed. To evaluate the bondability, the mechanical properties such as tensile strength, impact value, low cycle fatigue strength and creep strength of the bonded part were investigated using HIP bonded test specimens. And the detectability of ultrasonic detection tests were also studied on them.

  7. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 18 Conservation of Power and Water Resources 1 2011-04-01 2011-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket marketing...

  8. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket marketing...

  9. 75 FR 11557 - Woven Electric Blankets From China

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-11

    ... COMMISSION Woven Electric Blankets From China AGENCY: United States International Trade Commission. ACTION... of woven electric blankets, provided for in subheading 6301.10.00 of the Harmonized Tariff Schedule... has defined the subject merchandise as finished, semi- finished, and unassembled woven electric...

  10. Security Blanket or Mother: Which Benefits Linus during Pediatric Examinations?

    ERIC Educational Resources Information Center

    Ybarra, Gabriel; Passman, Richard H.; Eisenberg, Carl S. L.

    This study compared the degree to which young children were placated during a standard medical evaluation by the presence of their mother, blanket, mother plus blanket, or no supportive agent. Participating were 64 three-year-olds who underwent 4 routine medical procedures. Children were rated by their mothers as attached or nonattached to…

  11. US blanket technology programs. [Directory of current research

    SciTech Connect

    Nygren, R.E.

    1985-01-01

    Experimental research in US programs related to blanket technology is described through brief summaries of the objectives, facilities, recent experimental results and principal investigators for the Blanket Technology Program, TRIO-1 Experiment, TSTA, Fusion Hybrid Program and selected activities in the Fusion Materials and Fusion Safety Programs in neutronics research.

  12. Conceptual design of a self-cooled Flibe blanket

    SciTech Connect

    Sze, D.K.; Jung, J.; Cheng, E.T.; Piet, S.; Klein, A.

    1986-06-01

    A self-cooled Flibe blanket concept has been developed. The problems associated with tritium breeding, tritium containment, and corrosion have been investigated and potential solutions developed. A highly efficient and compact blanket and power conversion system has been incorporated. The resulting system is low pressure with high thermal efficiency and is inherently safe.

  13. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 18 Conservation of Power and Water Resources 1 2013-04-01 2013-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket...

  14. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 18 Conservation of Power and Water Resources 1 2014-04-01 2014-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket...

  15. 48 CFR 213.303 - Blanket purchase agreements (BPAs).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  16. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 18 Conservation of Power and Water Resources 1 2012-04-01 2012-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket...

  17. MIT LMFBR blanket research project. Final summary report

    SciTech Connect

    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.

  18. Fullerene Type Multilayer Insulation Blanket on a Spherical Cold Surface

    NASA Astrophysics Data System (ADS)

    Ohmori, T.; Shinozaki, T.; Kaneko, H.

    2010-04-01

    Fullerene type multilayer insulation blanket is proposed for the insulation around a spherical cold surface, and has been applied to a 10 liter spherical tank of liquid nitrogen. As fullerene has 32 faces with 90 edges, 32 polygons of stacked insulation sheets must be connected each other to fabricate such a MLI blanket. The MLI blanket has 90 slots at connecting part between the polygons where thermal radiation heat transfer should be reduced. The size of the polygon made by the insulation films must be chosen so as to avoid excess compressive pressure between stacked insulation sheets. The design principle and the method to fabricate the blanket were studied, and the experimental results obtained by the fullerene type MLI blanket applied to the 10 liter spherical tank were measured.

  19. ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

    SciTech Connect

    Berry, Jan; Ferrada, Juan J; Curd, Warren; Dell Orco, Dr. Giovanni; Barabash, Vladimir; Kim, Seokho H

    2011-01-01

    During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support

  20. Aerosol Blanket Likely Thinned During 1990s

    NASA Technical Reports Server (NTRS)

    2007-01-01

    Each day, a blanket of tiny particles drifting through the Earth's atmosphere filters out some of the sunlight headed for the planet's surface. These aerosols, including dust, smoke, and human-produced pollution, can reflect incoming light or absorb it, directly affecting the Earth's energy balance and climate. Aerosols also influence the climate indirectly, by affecting the brightness and amount of clouds. Research by NASA scientists on global aerosol patterns since the 1990s indicate the global aerosol blanket has likely thinned, allowing more sunlight to reach the Earth's surface over the past decade. The thinning of the blanket is shown by this trio of images based on satellite observations of aerosol optical thickness, a measurement that scientists use to describe how much the aerosols filter the incoming sunlight. Higher optical thickness (orange and red) means more sunlight blocking. The globes show average aerosol optical thickness for 1988-1991 (top), 2002-2005 (middle), and the change between the two time periods (bottom). Overall, the 1988-1991 image appears redder, a sign that aerosols were blocking more incoming sunlight; the 2002-2005 image has more light yellow areas. In the bottom image, small pockets of red (increased aerosol optical thickness), mostly near land masses in the Northern Hemisphere, are far outnumbered by blue areas (decreased aerosol optical thickness). Because they block incoming sunlight from reaching Earth's surface, aerosols may counterbalance greenhouse gas warming. The decline in the dimming power of aerosols over the past decade may have made the greenhouse warming trend more evident in the past decade than in previous decades. The scientists describe their results as a 'likely' trend because the National Oceanic and Atmospheric Administration satellite sensors they used in their analysis were not specifically designed to observe aerosols, and may contain some errors. However, specific, major aerosol events, such as large

  1. Iteration, Not Induction

    ERIC Educational Resources Information Center

    Dobbs, David E.

    2009-01-01

    The main purpose of this note is to present and justify proof via iteration as an intuitive, creative and empowering method that is often available and preferable as an alternative to proofs via either mathematical induction or the well-ordering principle. The method of iteration depends only on the fact that any strictly decreasing sequence of…

  2. Iteration, Not Induction

    ERIC Educational Resources Information Center

    Dobbs, David E.

    2009-01-01

    The main purpose of this note is to present and justify proof via iteration as an intuitive, creative and empowering method that is often available and preferable as an alternative to proofs via either mathematical induction or the well-ordering principle. The method of iteration depends only on the fact that any strictly decreasing sequence of…

  3. Review of accidental safety studies for the European HCPB test blanket system

    NASA Astrophysics Data System (ADS)

    Boccaccini, L. V.; Ciattaglia, S.; Meyder, R.; Jin, X.

    2007-07-01

    This paper presents a review of safety studies for accidental sequences in the European solid breeder test blanket module (TBM) system. These studies are the starting point for the Preliminary Safety Analysis Report of ITER, under preparation to get the construction permit first and then later the operation licence. In general the reduced inventory of activation products and tritium associated with the TBM system makes the impact of this test system almost negligible on the overall safety risk of ITER. Nevertheless, the possibility of jeopardizing the ITER safety concept has been analysed in connection to the consequences of specific accident sequences, e.g. the pressurization of the vacuum vessel due to the He coolant blow-down, the hydrogen production from the Be-steam reaction, the possible interconnection between the port cell and the vacuum vessel causing air ingress and the necessity to assure heat removal in the short and long periods. In the frame of this assessment, three LOCA sequences have been selected as representative of accidents judged to cover all scenarios envisaged in Cat II to IV events involving the TBM, namely, in-vessel LOCA, ex-vessel LOCA and in-box LOCA.

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

    NASA Technical Reports Server (NTRS)

    Scialdone, John J.

    1990-01-01

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

  5. Spacecraft thermal blanket cleaning - Vacuum baking or gaseous flow purging

    NASA Technical Reports Server (NTRS)

    Scialdone, John J.

    1992-01-01

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

  6. Spacecraft thermal blanket cleaning - Vacuum baking or gaseous flow purging

    NASA Technical Reports Server (NTRS)

    Scialdone, John J.

    1992-01-01

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

  7. Fracture strengths of HIPed DS-Cu/SS joints for ITER shielding blanket/first wall

    NASA Astrophysics Data System (ADS)

    Hatano, T.; Kanari, M.; Sato, S.; Gotoh, M.; Furuya, K.; Kuroda, T.; Saito, M.; Enoeda, M.; Takatsu, H.

    1998-10-01

    Fracture toughness and crack propagation tests were performed to investigate the effect of HIP temperature and fracture behavior of HIPed DS-Cu/SS joints. Test specimens of DS-Cu/SS HIPed joints were manufactured by bonding flat plates of DS-Cu and SS under HIP temperatures of 980°C, 1030°C and 1050°C. JQ of the joint at HIP temperature of 1050°C was larger than the other two joints. For the crack propagation test, two types of test specimens were prepared. One had a notch along the HIPed interface and the other in DS-Cu and normal to the interface. The crack in the former specimen propagated along the interface. On the other hand, the crack in the latter specimen propagated in the DS-Cu perpendicular to the loading direction, stopped at the interface, and then exfoliated along the HIPed interface. In the fracture tests, the crack was observed propagating in DS-Cu side at approximately 5-10 μm away from the interface.

  8. Hubble Space Telescope Thermal Blanket Repair Design and Implementation

    NASA Technical Reports Server (NTRS)

    Ousley, Wes; Skladany, Joseph; Dell, Lawrence

    2000-01-01

    Substantial damage to the outer layer of Hubble Space Telescope (HST) thermal blankets was observed during the February 1997 servicing mission. After six years in LEO, many areas of the aluminized Teflon(R) outer blanket layer had significant cracks, and some material was peeled away to expose inner layers to solar flux. After the mission, the failure mechanism was determined, and repair materials and priorities were selected for follow-on missions. This paper focuses on the thermal, mechanical, and EVA design requirements for the blanket repair, the creative solutions developed for these unique problems, hardware development, and testing.

  9. ITER ECH launcher options for start-up assist, bulk heating, and EC current drive experiments

    SciTech Connect

    Bigelow, T.S.

    1994-03-01

    Electron Cyclotron Heating (ECH) is proposed for providing plasma start-up, bulk heating, current drive, and other applications on the International Tokamak Experimental Reactor (ITER) project. The requirements for ECH power launching systems for ITER have been investigated, and several possible configurations that have been devised are described in this report. The proposed launcher designs use oversized circular corrugated waveguides that make small penetrations through the blanket modules and radiate into the plasma. The criteria used for the design calls for minimum blanket penetration area, maximum reliability, and optimum launched beam quality. The effects of the harsh plasma edge environment on the launcher are discussed. Power generation systems, windows, and other components of the ECH systems are also investigated. The designs presented are believed to be capable of operating reliably and are relatively easy to maintain remotely.

  10. The ITER design

    NASA Astrophysics Data System (ADS)

    Aymar, R.; Barabaschi, P.; Shimomura, Y.

    2002-05-01

    In 1998, after six years of joint work originally foreseen under the ITER engineering design activities (EDA) agreement, a design for ITER had been developed fulfilling all objectives and the cost target adopted by the ITER parties in 1992 at the start of the EDA. While accepting this design, the ITER parties recognized the possibility that they might be unable, for financial reasons, to proceed to the construction of the then foreseen device. The focus of effort in the ITER EDA since 1998 has been the development of a new design to meet revised technical objectives and a cost reduction target of about 50% of the previously accepted cost estimate. The rationale for the choice of parameters of the design has been based largely on system analysis drawing on the design solutions already developed and using the latest physics results and outputs from technology R&D projects. In so doing the joint central team and home teams converge towards a new design which will allow the exploration of a range of burning plasma conditions. The new ITER design, whilst having reduced technical objectives from its predecessor, will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy. Background, design features, performance, safety features, and R&D and future perspectives of the ITER design are discussed.

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

    NASA Astrophysics Data System (ADS)

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

    2013-11-01

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

  12. Full orbit calculation for lost alpha particle measurement on ITER

    SciTech Connect

    Funaki, D.; Isobe, M.; Nishiura, M.; Sato, Y.; Okamoto, A.; Kobuchi, T.; Kitajima, S.; Sasao, M.

    2008-10-15

    An orbit following calculation code with full gyromotion under the ITER magnetic field configuration has been developed to investigate escaping alpha particle orbits in ITER and to determine the geometrical arrangement for alpha particle detection. The code contained the full geometrical information of the first wall panels. It was carefully investigated whether an alpha particle escaping from the plasma through the last closed flux surface does not touch or intersect the first wall boundary before reaching the detection point. Candidates of blanket module modification have been studied to achieve effective measurement geometry for escaping alpha particle detection. The calculations showed that direct orbit loss and banana diffusion can be detected with a probe head recessed from the first wall surface.

  13. Wall conditioning on ITER

    NASA Astrophysics Data System (ADS)

    Shimada, Michiya; Pitts, Richard A.

    2011-08-01

    Like all tokamaks, ITER will require wall conditioning systems and strategies for successful operation from the point of view of plasma-facing surface preparation. Unlike today's devices however, ITER will have to manage large quantities of tritium fuel, imposing on wall conditioning a major responsibility for tritium inventory control. It will also feature the largest plasma-facing beryllium surface ever used in a tokamak and its high duty cycle and long pulse are expected to lead to the rapid formation of deposited layers in which tritium can accumulate. This paper summarises the currently planned ITER wall conditioning systems and describes the strategy for their use throughout exploitation of the device.

  14. RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

    SciTech Connect

    Ferrada, Juan J; Reiersen, Wayne T

    2011-01-01

    U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment

  15. Testing needs and experiments for solid breeder blankets

    NASA Astrophysics Data System (ADS)

    Gierszewski, P.; Puigh, R.

    1986-11-01

    Most of the critical issues for solid breeder blankets are related to the tritium and thermomechanical behavior of the solid breeder and multiplier. A major contributor to this uncertainty is the lack of definition of a preferred material, microstructure and form. Material selection and development is particularly cost-effective at present since much of the important blanket behavior is driven by local conditions rather than the overall blanket design, and since the choice of material affects the nature of the later and larger experiments. This stage of testing is also the most appropriate time to consider basic material choices such as sphere-pac or sintered pellet, and the incorporation of multiplier material into the blanket. The major testing needs and experiments are characterized here.

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

    NASA Astrophysics Data System (ADS)

    Sawko, Paul M.; Goldstein, Howard E.

    1992-04-01

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

  17. Blanket options for high-efficiency fusion power

    SciTech Connect

    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/sup 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/sub 2/ interior (cooled by argon) utilizing Li/sub 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/sup 0/C leading to an overall efficiency estimate of 55 to 60% for this reference case.

  18. Fusion blanket for high-efficiency power cycles

    SciTech Connect

    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/sup 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/sub 2/ interior (cooled by Ar) utilizing Li/sub 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/sup 0/C leading to an overall efficiency estimate of 55 to 60% for this reference case.

  19. Young Stars in Their Baby Blanket of Dust

    NASA Image and Video Library

    2008-02-11

    Newborn stars peek out from beneath their natal blanket of dust in this dynamic image of the Rho Ophiuchi dark cloud from NASA Spitzer Space Telescope. This nebula is about 407 light years away from Earth.

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

    NASA Technical Reports Server (NTRS)

    Sawko, Paul M.; Goldstein, Howard E.

    1992-01-01

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

  1. Cassini/Titan-4 Acoustic Blanket Development and Testing

    NASA Technical Reports Server (NTRS)

    Hughes, William O.; McNelis, Anne M.

    1996-01-01

    NASA Lewis Research Center recently led a multi-organizational effort to develop and test verify new acoustic blankets. These blankets support NASA's goal in reducing the Titan-4 payload fairing internal acoustic environment to allowable levels for the Cassini spacecraft. To accomplish this goal a two phase acoustic test program was utilized. Phase One consisted of testing numerous blanket designs in a flat panel configuration. Phase Two consisted of testing the most promising designs out of Phase One in a full scale cylindrical payload fairing. This paper will summarize this highly successful test program by providing the rationale and results for each test phase, the impacts of this testing on the Cassini mission, as well as providing some general information on blanket designs.

  2. Perl Modules for Constructing Iterators

    NASA Technical Reports Server (NTRS)

    Tilmes, Curt

    2009-01-01

    The Iterator Perl Module provides a general-purpose framework for constructing iterator objects within Perl, and a standard API for interacting with those objects. Iterators are an object-oriented design pattern where a description of a series of values is used in a constructor. Subsequent queries can request values in that series. These Perl modules build on the standard Iterator framework and provide iterators for some other types of values. Iterator::DateTime constructs iterators from DateTime objects or Date::Parse descriptions and ICal/RFC 2445 style re-currence descriptions. It supports a variety of input parameters, including a start to the sequence, an end to the sequence, an Ical/RFC 2445 recurrence describing the frequency of the values in the series, and a format description that can refine the presentation manner of the DateTime. Iterator::String constructs iterators from string representations. This module is useful in contexts where the API consists of supplying a string and getting back an iterator where the specific iteration desired is opaque to the caller. It is of particular value to the Iterator::Hash module which provides nested iterations. Iterator::Hash constructs iterators from Perl hashes that can include multiple iterators. The constructed iterators will return all the permutations of the iterations of the hash by nested iteration of embedded iterators. A hash simply includes a set of keys mapped to values. It is a very common data structure used throughout Perl programming. The Iterator:: Hash module allows a hash to include strings defining iterators (parsed and dispatched with Iterator::String) that are used to construct an overall series of hash values.

  3. ITER Cryoplant Infrastructures

    NASA Astrophysics Data System (ADS)

    Fauve, E.; Monneret, E.; Voigt, T.; Vincent, G.; Forgeas, A.; Simon, M.

    2017-02-01

    The ITER Tokamak requires an average 75 kW of refrigeration power at 4.5 K and 600 kW of refrigeration Power at 80 K to maintain the nominal operation condition of the ITER thermal shields, superconducting magnets and cryopumps. This is produced by the ITER Cryoplant, a complex cluster of refrigeration systems including in particular three identical Liquid Helium Plants and two identical Liquid Nitrogen Plants. Beyond the equipment directly part of the Cryoplant, colossal infrastructures are required. These infrastructures account for a large part of the Cryoplants lay-out, budget and engineering efforts. It is ITER Organization responsibility to ensure that all infrastructures are adequately sized and designed to interface with the Cryoplant. This proceeding presents the overall architecture of the cryoplant. It provides order of magnitude related to the cryoplant building and utilities: electricity, cooling water, heating, ventilation and air conditioning (HVAC).

  4. Diagnostics for ITER

    SciTech Connect

    Donne, A. J. H.; Hellermann, M. G. von; Barnsley, R.

    2008-10-22

    After an introduction into the specific challenges in the field of diagnostics for ITER (specifically high level of nuclear radiation, long pulses, high fluxes of particles to plasma facing components, need for reliability and robustness), an overview will be given of the spectroscopic diagnostics foreseen for ITER. The paper will describe both active neutral-beam based diagnostics as well as passive spectroscopic diagnostics operating in the visible, ultra-violet and x-ray spectral regions.

  5. Applications of the Aqueous Self-Cooled Blanket concept

    SciTech Connect

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

    1986-11-01

    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.

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

    SciTech Connect

    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.

  7. Thin Thermal-Insulation Blankets for Very High Temperatures

    NASA Technical Reports Server (NTRS)

    Choi, Michael K.

    2003-01-01

    Thermal-insulation blankets of a proposed type would be exceptionally thin and would endure temperatures up to 2,100 C. These blankets were originally intended to protect components of the NASA Solar Probe spacecraft against radiant heating at its planned closest approach to the Sun (a distance of 4 solar radii). These blankets could also be used on Earth to provide thermal protection in special applications (especially in vacuum chambers) for which conventional thermal-insulation blankets would be too thick or would not perform adequately. A blanket according to the proposal (see figure) would be made of molybdenum, titanium nitride, and carbon- carbon composite mesh, which melt at temperatures of 2,610, 2,930, and 2,130 C, respectively. The emittance of molybdenum is 0.24, while that of titanium nitride is 0.03. Carbon-carbon composite mesh is a thermal insulator. Typically, the blanket would include 0.25-mil (.0.00635-mm)-thick hot-side and cold-side cover layers of molybdenum. Titanium nitride would be vapor-deposited on both surfaces of each cover layer. Between the cover layers there would be 10 inner layers of 0.15-mil (.0.0038-mm)-thick molybdenum with vapor-deposited titanium nitride on both sides of each layer. The thickness of each titanium nitride coat would be about 1,000 A. The cover and inner layers would be interspersed with 0.25-mil (0.00635-mm)-thick layers of carbon-carbon composite mesh. The blanket would have total thickness of 4.75 mils (approximately equal to 0.121 mm) and an areal mass density of 0.7 kilograms per square meter. One could, of course, increase the thermal- insulation capability of the blanket by increasing number of inner layers (thereby unavoidably increasing the total thickness and mass density).

  8. Performance assessment of the antenna setup for the ITER plasma position reflectometry in-vessel systems.

    PubMed

    Varela, P; Belo, J H; Quental, P B

    2016-11-01

    The design of the in-vessel antennas for the ITER plasma position reflectometry diagnostic is very challenging due to the need to cope both with the space restrictions inside the vacuum vessel and with the high mechanical and thermal loads during ITER operation. Here, we present the work carried out to assess and optimise the design of the antenna. We show that the blanket modules surrounding the antenna strongly modify its characteristics and need to be considered from the early phases of the design. We also show that it is possible to optimise the antenna performance, within the design restrictions.

  9. Performance assessment of the antenna setup for the ITER plasma position reflectometry in-vessel systems

    NASA Astrophysics Data System (ADS)

    Varela, P.; Belo, J. H.; Quental, P. B.

    2016-11-01

    The design of the in-vessel antennas for the ITER plasma position reflectometry diagnostic is very challenging due to the need to cope both with the space restrictions inside the vacuum vessel and with the high mechanical and thermal loads during ITER operation. Here, we present the work carried out to assess and optimise the design of the antenna. We show that the blanket modules surrounding the antenna strongly modify its characteristics and need to be considered from the early phases of the design. We also show that it is possible to optimise the antenna performance, within the design restrictions.

  10. ITER Cryoplant Status and Economics of the LHe plants

    NASA Astrophysics Data System (ADS)

    Monneret, E.; Chalifour, M.; Bonneton, M.; Fauve, E.; Voigt, T.; Badgujar, S.; Chang, H.-S.; Vincent, G.

    The ITER cryoplant is composed of helium and nitrogen refrigerators and generator combined with 80 K helium loop plants and external purification systems. Storage and recovery of the helium inventory is provided in warm and cold (80 K and 4.5 K) helium tanks.The conceptual design of the ITER cryoplant has been completed, the technical requirements defined for industrial procurement and contracts signed with industry. Each contract covers the design, manufacturing, installation and commissioning. Design is under finalization and manufacturing has started. First deliveries are scheduled by end of 2015.The various cryoplant systems are designed based on recognized codes and international standards to meet the availability, the reliability and the time between maintenance imposed by the long-term uninterrupted operation of the ITER Tokamak. In addition, ITER has to consider the constraint of a nuclear installation.ITER Organization (IO) is responsible for the liquid helium (LHe) Plants contract signed end of 2012 with industry. It is composed of three LHe Plants, working in parallel and able to provide a total average cooling capacity of 75 kW at 4.5 K. Based on concept designed developed with industries and the procurement phase, ITER has accumulated data to broaden the scaling laws for costing such systems.After describing the status of ITER cryoplant part of the cryogenic system, we shall present the economics of the ITER LHe Plants based on key design requirements, choice and challenges of this ITER Organization procurement.

  11. Flow characteristics of the Cascade granular blanket

    SciTech Connect

    Pitts, J.H.; Walton, O.R.

    1985-07-01

    Analysis of a single granule on a rotating cone shows that for the 35/sup 0/ half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer.

  12. Flow characteristics of the Cascade granular blanket

    SciTech Connect

    Pitts, J.H.; Walton, O.R.

    1985-04-15

    Analysis of a single granule on a rotating cone shows that for the 35/sup 0/ half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer.

  13. Blanket of Snow Covers Salt Lake City

    NASA Technical Reports Server (NTRS)

    2002-01-01

    On December 23, 2001, less than two months before the start of the 2002 Winter Olympics, snow blankets Salt Lake City and the surrounding area. The Great Salt Lake, on the left hand side of the image above, often contributes to the region's snowfall through the 'lake-effect.' As cold air passes over a large body of water it both warms and absorbs moisture. The warm air then rises (like a hot air balloon) and cools again. As it cools, the water vapor condenses out, resulting in snowfall. Just to the east (right) of the Great Salt Lake the mountains of the Wasatch Range lift air from the lake even higher, enhancing the lake-effect, resulting in an average snowfall of 64 inches a year in Salt Lake City and 140 inches in Park City, which is located at the foot of the Wasatch Front. For more information about the lake-effect, read Lake-Effect Snowfalls. Image courtesy Jacques Descloitres, MODIS Land Rapid Response Team at NASA GSFC

  14. Blanket of Snow Covers Salt Lake City

    NASA Technical Reports Server (NTRS)

    2002-01-01

    On December 23, 2001, less than two months before the start of the 2002 Winter Olympics, snow blankets Salt Lake City and the surrounding area. The Great Salt Lake, on the left hand side of the image above, often contributes to the region's snowfall through the 'lake-effect.' As cold air passes over a large body of water it both warms and absorbs moisture. The warm air then rises (like a hot air balloon) and cools again. As it cools, the water vapor condenses out, resulting in snowfall. Just to the east (right) of the Great Salt Lake the mountains of the Wasatch Range lift air from the lake even higher, enhancing the lake-effect, resulting in an average snowfall of 64 inches a year in Salt Lake City and 140 inches in Park City, which is located at the foot of the Wasatch Front. For more information about the lake-effect, read Lake-Effect Snowfalls. Image courtesy Jacques Descloitres, MODIS Land Rapid Response Team at NASA GSFC

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

    SciTech Connect

    Hamm, L.L.

    1998-10-07

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

  16. Mars and the Amazing Technicolor Ejecta Blanket

    NASA Image and Video Library

    2017-07-12

    This image from NASA's Mars Reconnaissance Orbiter shows the exposed bedrock of an ejecta blanket of an unnamed crater in the Mare Serpentis region of Mars. Ejecta, when exposed, are truly an eye-opening feature, as they reveal the sometimes exotic subsurface, and materials created by impacts (close-up view). This ejecta shares similarities to others found elsewhere on Mars, which are of particular scientific interest for the extent of exposure and diverse colors. (For example, the Hargraves Crater ejecta, in the Nili Fossae trough region, was once considered as a candidate landing site for the next NASA Mars rover 2020.) The colors observed in this picture represent different rocks and minerals, now exposed on the surface. Blue in HiRISE infrared color images generally depicts iron-rich minerals, like olivine and pyroxene. Lighter colors, such as yellow, indicate the presence of altered rocks. The possible sources of the ejecta is most likely from two unnamed craters. How do we determine which crater deposited the ejecta? A full-scale image shows numerous linear features that are observed trending in an east-west direction. These linear features indicate the flow direction of the ejecta from its unnamed host crater. Therefore, if we follow them, we find that they emanate from the bottom of the two unnamed craters. If the ejecta had originated from the top crater, then we would expect the linear features at the location of our picture to trend northwest to southeast. https://photojournal.jpl.nasa.gov/catalog/PIA21782

  17. Robust iterative methods

    SciTech Connect

    Saadd, Y.

    1994-12-31

    In spite of the tremendous progress achieved in recent years in the general area of iterative solution techniques, there are still a few obstacles to the acceptance of iterative methods in a number of applications. These applications give rise to very indefinite or highly ill-conditioned non Hermitian matrices. Trying to solve these systems with the simple-minded standard preconditioned Krylov subspace methods can be a frustrating experience. With the mathematical and physical models becoming more sophisticated, the typical linear systems which we encounter today are far more difficult to solve than those of just a few years ago. This trend is likely to accentuate. This workshop will discuss (1) these applications and the types of problems that they give rise to; and (2) recent progress in solving these problems with iterative methods. The workshop will end with a hopefully stimulating panel discussion with the speakers.

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

    NASA Astrophysics Data System (ADS)

    Kramer, G. J.; Budny, R.; Nazikian, R.; Heidbrink, W. W.; Kurki-Suonio, T.; Salmi, A.; Schaffer, M. J.; van Zeeland, M. A.; Shinohara, K.; Snipes, J. A.; Spong, D.

    2010-11-01

    The fast beam-ion confinement in the presence of a scaled mock-up of two Test Blanket Modules (TBM) for ITER was studied in DIII-D. The TBM on DIII-D has four vertically arranged protective carbon tiles with thermocouples placed at the back of each tile. Temperature increases of up to 200^oC were measured for the two tiles closest to the midplane when the TBM fields were present. These measurements agree qualitatively with results from the full orbit-following beam-ion code, SPIRAL, that predict beam-ion losses to be localized on the central two carbon tiles when the TBM fields present. Within the experimental uncertainties no significant change in the fast-ion population was found in the core of these plasmas which is consistent with SPIRAL analysis. These experiments indicate that the TBM fields do not affect the fast-ion confinement in a harmful way which is good news for ITER.

  19. Iterated multidimensional wave conversion

    SciTech Connect

    Brizard, A. J.; Tracy, E. R.; Johnston, D.; Kaufman, A. N.; Richardson, A. S.; Zobin, N.

    2011-12-23

    Mode conversion can occur repeatedly in a two-dimensional cavity (e.g., the poloidal cross section of an axisymmetric tokamak). We report on two novel concepts that allow for a complete and global visualization of the ray evolution under iterated conversions. First, iterated conversion is discussed in terms of ray-induced maps from the two-dimensional conversion surface to itself (which can be visualized in terms of three-dimensional rooms). Second, the two-dimensional conversion surface is shown to possess a symplectic structure derived from Dirac constraints associated with the two dispersion surfaces of the interacting waves.

  20. Rescheduling with iterative repair

    NASA Technical Reports Server (NTRS)

    Zweben, Monte; Davis, Eugene; Daun, Brian; Deale, Michael

    1992-01-01

    This paper presents a new approach to rescheduling called constraint-based iterative repair. This approach gives our system the ability to satisfy domain constraints, address optimization concerns, minimize perturbation to the original schedule, produce modified schedules, quickly, and exhibits 'anytime' behavior. The system begins with an initial, flawed schedule and then iteratively repairs constraint violations until a conflict-free schedule is produced. In an empirical demonstration, we vary the importance of minimizing perturbation and report how fast the system is able to resolve conflicts in a given time bound. We also show the anytime characteristics of the system. These experiments were performed within the domain of Space Shuttle ground processing.

  1. Rescheduling with iterative repair

    NASA Technical Reports Server (NTRS)

    Zweben, Monte; Davis, Eugene; Daun, Brian; Deale, Michael

    1992-01-01

    This paper presents a new approach to rescheduling called constraint-based iterative repair. This approach gives our system the ability to satisfy domain constraints, address optimization concerns, minimize perturbation to the original schedule, and produce modified schedules quickly. The system begins with an initial, flawed schedule and then iteratively repairs constraint violations until a conflict-free schedule is produced. In an empirical demonstration, we vary the importance of minimizing perturbation and report how fast the system is able to resolve conflicts in a given time bound. These experiments were performed within the domain of Space Shuttle ground processing.

  2. Thin blanket designs for the Elongated Tokamak Commercial Reactor

    SciTech Connect

    Wong, C.P.C.; Bourque, R.F.; Cheng, E.T.; Creedon, R.L.; Schultz, K.R.

    1986-11-01

    The Elongated Tokamak (ET)* is an innovative concept that uses as highly elongated plasma (plasma height-to-width ratio of 6-10) to allow high plasma current and high toroidal betas. ET has the potential for the development of small-size, high-power density, low-cost fusion reactors using normal conducting coils. The elongated plasma shape is achieved by use of a continuous stack of PF coils parallel to the plasma surface on both inbound and outbound sides. To achieve plasma stability, these coil stacks must be located no further than one plasma minor radius from the plasma edge, greatly restricting the space available for blankets. In order to assess the potential of a small reactor, the authors evaluated and designed blankets 30 to 40 cm thick. Three different thin blanket designs were found to be acceptable: FLiBe self-cooled, helium-cooled lithium, and helium-cooled 17Li83Pb blanket designs. A lithium-cooled integrated blanket-coil design (BLITZ-coil) was also found to be suitable for the ET commercial reactors.

  3. ITER Fusion Energy

    ScienceCinema

    Dr. Norbert Holtkamp

    2016-07-12

    ITER (in Latin “the way”) is designed to demonstrate the scientific and technological feasibility of fusion energy. Fusion is the process by which two light atomic nuclei combine to form a heavier over one and thus release energy. In the fusion process two isotopes of hydrogen – deuterium and tritium – fuse together to form a helium atom and a neutron. Thus fusion could provide large scale energy production without greenhouse effects; essentially limitless fuel would be available all over the world. The principal goals of ITER are to generate 500 megawatts of fusion power for periods of 300 to 500 seconds with a fusion power multiplication factor, Q, of at least 10. Q ? 10 (input power 50 MW / output power 500 MW). The ITER Organization was officially established in Cadarache, France, on 24 October 2007. The seven members engaged in the project – China, the European Union, India, Japan, Korea, Russia and the United States – represent more than half the world’s population. The costs for ITER are shared by the seven members. The cost for the construction will be approximately 5.5 billion Euros, a similar amount is foreseen for the twenty-year phase of operation and the subsequent decommissioning.

  4. An Iterative Angle Trisection

    ERIC Educational Resources Information Center

    Muench, Donald L.

    2007-01-01

    The problem of angle trisection continues to fascinate people even though it has long been known that it can't be done with straightedge and compass alone. However, for practical purposes, a good iterative procedure can get you as close as you want. In this note, we present such a procedure. Using only straightedge and compass, our procedure…

  5. Iterative software kernels

    SciTech Connect

    Duff, I.

    1994-12-31

    This workshop focuses on kernels for iterative software packages. Specifically, the three speakers discuss various aspects of sparse BLAS kernels. Their topics are: `Current status of user lever sparse BLAS`; Current status of the sparse BLAS toolkit`; and `Adding matrix-matrix and matrix-matrix-matrix multiply to the sparse BLAS toolkit`.

  6. Direct Lit Electrolysis In A Metallic Lithium Fusion Blanket

    SciTech Connect

    Colon-Mercado, H.; Babineau, D.; Elvington, M.; Garcia-Diaz, B.; Teprovich, J.; Vaquer, A.

    2015-10-13

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

  7. Direct LiT Electrolysis in a Metallic Fusion Blanket

    SciTech Connect

    Olson, Luke

    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.

  8. Preventative Maintenance.

    ERIC Educational Resources Information Center

    Migliorino, James

    Boards of education must be convinced that spending money up front for preventive maintenance will, in the long run, save districts' tax dollars. A good program of preventive maintenance can minimize disruption of service; reduce repair costs, energy consumption, and overtime; improve labor productivity and system equipment reliability; handle…

  9. Preventative Maintenance.

    ERIC Educational Resources Information Center

    Migliorino, James

    Boards of education must be convinced that spending money up front for preventive maintenance will, in the long run, save districts' tax dollars. A good program of preventive maintenance can minimize disruption of service; reduce repair costs, energy consumption, and overtime; improve labor productivity and system equipment reliability; handle…

  10. Maintenance Downtime

    Atmospheric Science Data Center

    2013-07-10

    ... will be unavailable March 5, 2013 8:00 am to 5:00 pm due to database maintenance. Date(s):  Tuesday, March 5, 2013 ... will be unavailable March 5, 2013 8:00 am to 5:00 pm due to database maintenance. ...

  11. Maintenance Budgeting.

    ERIC Educational Resources Information Center

    Smith, J. McCree

    Three methods for the preparation of maintenance budgets are discussed--(1) a traditional method, inconclusive and obsolete, based on gross square footage, (2) the formula approach method based on building classification (wood-frame, masonry-wood, masonry-concrete) with maintenance cost factors for each type plus custodial service rates by type of…

  12. Software Maintenance.

    ERIC Educational Resources Information Center

    Cannon, Glenn; Jobe, Holly

    Proper cleaning and storage of audiovisual aids is outlined in this brief guide. Materials and equipment needed for first line maintenance are listed, as well as maintenance procedures for records, audio and video tape, film, filmstrips, slides, realia, models, prints, graphics, maps, and overhead transparencies. A 15-item quiz on software…

  13. Experimental impacts into Teflon targets and LDEF thermal blankets

    NASA Technical Reports Server (NTRS)

    Hoerz, F.; Cintala, M. J.; Zolensky, M. E.; Bernhard, R. P.; See, T. H.

    1994-01-01

    The Long Duration Exposure Facility (LDEF) exposed approximately 20 sq m of identical thermal protective blankets, predominantly on the Ultra-Heavy Cosmic Ray Experiment (UHCRE). Approximately 700 penetration holes greater than 300 micron in diameter were individually documented, while thousands of smaller penetrations and craters occurred in these blankets. As a result of their 5.7 year exposure and because they pointed into a variety of different directions relative to the orbital motion of the nonspinning LDEF platform, these blankets can reveal important dynamic aspects of the hypervelocity particle environment in near-earth orbit. The blankets were composed of an outer teflon layer (approximately 125 micron thick), followed by a vapor-deposited rear mirror of silver (less than 1000 A thick) that was backed with an organic binder and a thermal protective paint (approximately 50 to 75 micron thick), resulting in a cumulative thickness (T) of approximately 175 to 200 microns for the entire blanket. Many penetrations resulted in highly variable delaminations of the teflon/metal or metal/organic binder interfaces that manifest themselves as 'dark' halos or rings, because of subsequent oxidation of the exposed silver mirror. The variety of these dark albedo features is bewildering, ranging from totally absent, to broad halos, to sharp single or multiple rings. Over the past year experiments were conducted over a wide range of velocities (i.e., 1 to 7 km/s) to address velocity dependent aspects of cratering and penetrations of teflon targets. In addition, experiments were performed with real LDEF thermal blankets to duplicate the LDEF delaminations and to investigate a possible relationship of initial impact conditions on the wide variety of dark halo and ring features.

  14. Study of Automated Module Fabrication for Lightweight Solar Blanket Utilization

    NASA Technical Reports Server (NTRS)

    Gibson, C. E.

    1979-01-01

    Cost-effective automated techniques for accomplishing the titled purpose; based on existing in-house capability are described. As a measure of the considered automation, the production of a 50 kilowatt solar array blanket, exclusive of support and deployment structure, within an eight-month fabrication period was used. Solar cells considered for this blanket were 2 x 4 x .02 cm wrap-around cells, 2 x 2 x .005 cm and 3 x 3 x .005 cm standard bar contact thin cells, all welded contacts. Existing fabrication processes are described, the rationale for each process is discussed, and the capability for further automation is discussed.

  15. 77 FR 12281 - Williston Basin Interstate Pipeline Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-29

    ... Energy Regulatory Commission Williston Basin Interstate Pipeline Company; Notice of Request Under Blanket Authorization Take notice that on February 15, 2012, Williston Basin Interstate Pipeline Company (Williston Basin), 1250 West Century Avenue, Bismarck, North Dakota 58503, pursuant to its blanket...

  16. Spray-on technique simplifies fabrication of complex thermal insulation blanket

    NASA Technical Reports Server (NTRS)

    Bond, W. E. G.; Raymond, R.

    1966-01-01

    Spray-on process constructs molds used in forming sections of thermal insulation blankets. The process simplifies the fabrication of blankets by eliminating much of the equipment formerly required and decreasing the time involved.

  17. Task toward a Realization of Commercial Tokamak Fusion Plants in 2050 -The Role of ITER and the Succeeding Developments- 4.Technology and Material Research in Fusion Power Plant Development

    NASA Astrophysics Data System (ADS)

    Akiba, Masato; Matsui, Hideki; Takatsu, Hideyuki; Konishi, Satoshi

    Technical issues regarding the fusion power plant that are required to be developed in the period of ITER construction and operation, both with ITER and with other facilities that complement ITER are described in this section. Three major fields are considered to be important in fusion technology. Section 4.1 summarizes blanket study, and ITER Test Blanket Module (TBM) development that focuses its effort on the first generation power blanket to be installed in DEMO. ITER will be equipped with 6 TBMs which are developed under each party's fusion program. In Japan, the solid breeder using water as a coolant is the primary candidate, and He-cooled pebble bed is the alternative. Other liquid options such as LiPb, Li or molten salt are developed by other parties' initiatives. The Test Blanket Working Group (TBWG) is coordinating these efforts. Japanese universities are investigating advanced concepts and fundamental crosscutting technologies. Section 4.2 introduces material development and particularly, the international irradiation facility, IFMIF. Reduced activation ferritic/martensitic steels are identified as promising candidates for the structural material of the first generation fusion blanket, while and vanadium alloy and SiC/SiC composite are pursued as advanced options. The IFMIF is currently planning the next phase of joint activity, EVEDA (Engineering Validation and Engineering Design Activity) that encompasses construction. Material studies together with the ITER TBM will provide essential technical information for development of the fusion power plant. Other technical issues to be addressed regarding the first generation fusion power plant are summarized in section 4.3. Development of components for ITER made remarkable progress for the major essential technology also necessary for future fusion plants, however many still need further improvements toward power plant. Such areas includes; the divertor, plasma heating/current drive, magnets, tritium, and

  18. Progress on ITER Diagnostic Integration

    NASA Astrophysics Data System (ADS)

    Johnson, David; Feder, Russ; Klabacha, Jonathan; Loesser, Doug; Messineo, Mike; Stratton, Brentley; Wood, Rick; Zhai, Yuhu; Andrew, Phillip; Barnsley, Robin; Bertschinger, Guenter; Debock, Maarten; Reichle, Roger; Udintsev, Victor; Vayakis, George; Watts, Christopher; Walsh, Michael

    2013-10-01

    On ITER, front-end components must operate reliably in a hostile environment. Many will be housed in massive port plugs, which also shield the machine from radiation. Multiple diagnostics reside in a single plug, presenting new challenges for developers. Front-end components must tolerate thermally-induced stresses, disruption-induced mechanical loads, stray ECH radiation, displacement damage, and degradation due to plasma-induced coatings. The impact of failures is amplified due to the difficulty in performing robotic maintenance on these large structures. Motivated by needs to minimize disruption loads on the plugs, standardize the handling of shield modules, and decouple the parallel efforts of the many parties, the packaging strategy for diagnostics has recently focused on the use of 3 vertical shield modules inserted from the plasma side into each equatorial plug structure. At the front of each is a detachable first wall element with customized apertures. Progress on US equatorial and upper plugs will be used as examples, including the layout of components in the interspace and port cell regions. Supported by PPPL under contract DE-AC02-09CH11466 and UT-Battelle, LLC under contract DE-AC05-00OR22725 with the U.S. DOE.

  19. The Iterate Manual

    DTIC Science & Technology

    1990-10-01

    is probably a bad idea. A better versica would use a temporary: (defmacro sum-of-squares (expr) (let ((temp ( gensym ))) ’(lot (,temp ,expr)) (sum...val ( gensym )) (tempi ( gensym )) (temp2 ( gensym )) (winner (or var iterate::*result-var*))) ’(progn (with ,max-val - nil) (with ,winner = nil) (cond ((null...the elements of a vector (disregards fill-pointer)" (let ((vect ( gensym )) (end ( gensym )) (index ( gensym ))) ’(progn (with ,vect - v) (with ,end = (array

  20. Iterative initial condition reconstruction

    NASA Astrophysics Data System (ADS)

    Schmittfull, Marcel; Baldauf, Tobias; Zaldarriaga, Matias

    2017-07-01

    Motivated by recent developments in perturbative calculations of the nonlinear evolution of large-scale structure, we present an iterative algorithm to reconstruct the initial conditions in a given volume starting from the dark matter distribution in real space. In our algorithm, objects are first moved back iteratively along estimated potential gradients, with a progressively reduced smoothing scale, until a nearly uniform catalog is obtained. The linear initial density is then estimated as the divergence of the cumulative displacement, with an optional second-order correction. This algorithm should undo nonlinear effects up to one-loop order, including the higher-order infrared resummation piece. We test the method using dark matter simulations in real space. At redshift z =0 , we find that after eight iterations the reconstructed density is more than 95% correlated with the initial density at k ≤0.35 h Mpc-1 . The reconstruction also reduces the power in the difference between reconstructed and initial fields by more than 2 orders of magnitude at k ≤0.2 h Mpc-1 , and it extends the range of scales where the full broadband shape of the power spectrum matches linear theory by a factor of 2-3. As a specific application, we consider measurements of the baryonic acoustic oscillation (BAO) scale that can be improved by reducing the degradation effects of large-scale flows. In our idealized dark matter simulations, the method improves the BAO signal-to-noise ratio by a factor of 2.7 at z =0 and by a factor of 2.5 at z =0.6 , improving standard BAO reconstruction by 70% at z =0 and 30% at z =0.6 , and matching the optimal BAO signal and signal-to-noise ratio of the linear density in the same volume. For BAO, the iterative nature of the reconstruction is the most important aspect.

  1. 75 FR 50991 - Antidumping Duty Order: Certain Woven Electric Blankets From the People's Republic of China

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-18

    ... accommodate either the electric blanket's wiring or a subassembly containing the electric blanket's wiring (e... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF COMMERCE International Trade Administration Antidumping Duty Order: Certain Woven Electric Blankets From the People's...

  2. Neutron activation for ITER

    SciTech Connect

    Barnes, C.W.; Loughlin, M.J.; Nishitani, Takeo

    1996-04-29

    There are three primary goals for the Neutron Activation system for ITER: maintain a robust relative measure of fusion power with stability and high dynamic range (7 orders of magnitude); allow an absolute calibration of fusion power (energy); and provide a flexible and reliable system for materials testing. The nature of the activation technique is such that stability and high dynamic range can be intrinsic properties of the system. It has also been the technique that demonstrated (on JET and TFTR) the highest accuracy neutron measurements in DT operation. Since the gamma-ray detectors are not located on the tokamak and are therefore amenable to accurate characterization, and if material foils are placed very close to the ITER plasma with minimum scattering or attenuation, high overall accuracy in the fusion energy production (7--10%) should be achievable on ITER. In the paper, a conceptual design is presented. A system is shown to be capable of meeting these three goals, also detailed design issues remain to be solved.

  3. Calculation of Neoclassical Ripple Transport in ITER using SFINCS

    NASA Astrophysics Data System (ADS)

    Paul, Elizabeth; Landreman, Matt; Spong, Donald; Poli, Francesca

    2016-10-01

    Neoclassical interactions with non-axisymmetric magnetic fields cause a damping force known as neoclassical toroidal viscosity (NTV). The toroidal symmetry of ITER will be broken by the finite number of toroidal field coils and the presence of perturbing ferromagnetic structures such as test blanket modules (TBM) and ferritic inserts (FI). We compute 3D equilibrium magnetic fields for an ITER steady-state scenario using VMEC and calculate neoclassical transport quantities in the presence of these error fields using the Stellarator Fokker-Planck Iterative Neoclassical Conservative Solver (SFINCS) code. In the presence of both the FI and TBM, the net effect is a decrease in toroidal damping. The magnitude of NTV torque density at large radii (r / a >= 0.7) is comparable to the NBI torque density at small radii (r / a <= 0.4), but is opposite in direction. This could indicate the possibility of generating sheared flows. The magnetic field ripple does not significantly affect the neoclassical tokamak relationship between radial electric field and parallel flow velocity, and at r / a >= 0.7 the ripple drives additional collisional heat flux comparable to the axisymmetric neoclassical flux. Supported by DEFG0293ER54197 and DEFC0208ER54964.

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

    SciTech Connect

    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.

  5. Nitrogen sparging and blanketing of water storage tanks

    SciTech Connect

    Jonas, O.

    2000-04-01

    In many industrial processes, including most utility and industrial steam systems, good deaerated makeup and condensate water is stored in open-to-air storage tanks where it is contaminated by oxygen, carbon dioxide (CO{sub 2}), and dirt before it is used. This contamination can be prevented by nitrogen sparging and blanketing of storage tanks.

  6. Unified first wall - blanket structure for plasma device applications

    DOEpatents

    Gruen, D.M.

    A plasma device is described for use in controlling nuclear reactions within the plasma including a first wall and blanket formed in a one-piece structure composed of a solid solution containing copper and lithium and melting above about 500/sup 0/C.

  7. Summary of the target-blanket breakout group

    SciTech Connect

    Capiello, M.; Bell, C.; 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.

  8. Measuring High Temperatures In Ceramic-Fiber Blankets

    NASA Technical Reports Server (NTRS)

    Kourtides, Demetrius A.

    1996-01-01

    Thermocouple assemblies devised specifically for measuring temperatures at fixed locations within insulating blankets made of such ceramic fibers as alumina, silicon carbide, and/or aluminoborosilicate. Thermocouples measure temperatures from 100 to 3,200 degrees F in oxidizing atmospheres. Wires enclosed in alumina sheath for protection against hot oxidation and mechanical damage.

  9. 48 CFR 313.303 - Blanket purchase agreements.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Blanket purchase agreements. 313.303 Section 313.303 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES CONTRACTING METHODS AND CONTRACT TYPES SIMPLIFIED ACQUISITION PROCEDURES Simplified Acquisition Methods...

  10. Warm Ocean Temperatures Blanket the Far-Western Pacific

    NASA Image and Video Library

    2001-03-28

    Data taken during a 10-day collection cycle ending March 9, 2001, show that above-normal sea-surface heights and warmer ocean temp. red and white areas still blanket the far-western tropical Pacific and much of the north and south mid-Pacific.

  11. Technical issues for beryllium use in fusion blanket applications

    SciTech Connect

    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.

  12. Unified first wall-blanket structure for plasma device applications

    DOEpatents

    Gruen, Dieter M.

    1987-01-01

    A plasma device for use in controlling nuclear reactions within the plasma including a first wall and blanket formed in a one-piece structure composed of a solid solution containing copper and lithium and melting above about 500.degree. C.

  13. 18 CFR 284.303 - OCS blanket certificates.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... open-access, nondiscriminatory transportation service pursuant to a blanket transportation certificate... COMMISSION, DEPARTMENT OF ENERGY OTHER REGULATIONS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES CERTAIN SALES AND TRANSPORTATION OF NATURAL GAS UNDER THE NATURAL GAS POLICY ACT OF 1978...

  14. 18 CFR 152.1 - Exemption applications and blanket certificates.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 18 Conservation of Power and Water Resources 1 2014-04-01 2014-04-01 false Exemption applications and blanket certificates. 152.1 Section 152.1 Conservation of Power and Water Resources FEDERAL ENERGY... certificates of public convenience and necessity are issued pursuant to section 7(c) of the Natural Gas Act to...

  15. 18 CFR 152.1 - Exemption applications and blanket certificates.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 18 Conservation of Power and Water Resources 1 2011-04-01 2011-04-01 false Exemption applications and blanket certificates. 152.1 Section 152.1 Conservation of Power and Water Resources FEDERAL ENERGY... certificates of public convenience and necessity are issued pursuant to section 7(c) of the Natural Gas Act to...

  16. 18 CFR 152.1 - Exemption applications and blanket certificates.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... REGULATORY COMMISSION, DEPARTMENT OF ENERGY REGULATIONS UNDER NATURAL GAS ACT APPLICATION FOR EXEMPTION FROM THE PROVISIONS OF THE NATURAL GAS ACT PURSUANT TO SECTION 1(C) THEREOF AND ISSUANCE OF BLANKET...) Application for exemption from the provisions of the Natural Gas Act and the rules and regulations of...

  17. SEAL Studies of Variant Blanket Concepts and Materials

    NASA Astrophysics Data System (ADS)

    Cook, I.; Taylor, N. P.; Forty, C. B. A.; Han, W. E.

    1997-09-01

    Within the European SEAL ( Safety and Environmental Assessment of fusion power, Long-term) program, safety and environmental assessments have been performed which extend the results of the earlier SEAFP (Safety and Environmental Assessment of Fusion Power) program to a wider range of blanket designs and material choices. The four blanket designs analysed were those which had been developed within the Blanket program of the European Fusion Programme. All four are based on martensitic steel as structural material, and otherwise may be summarized as: water-cooled lithium-lead; dual-cooled lithium-lead; helium-cooled lithium silicate (BOT geometry); helium-cooled lithium aluminate (or zirconate) (BIT geometry). The results reveal that all the blankets show the favorable S&E characteristics of fusion, though there are interesting and significant differences between them. The key results are described. Assessments have also been performed of a wider range of materials than was considered in SEAFP. These were: an alternative vanadium alloy, an alternative low-activation martensitic steel, titanium-aluminum intermetallic, and SiC composite. Assessed impurities were included in the compositions, and these had very important effects upon some of the results. Key results impacting upon accident characteristics, recycling, and waste management are described.

  18. ITER in-vessel system design and performance

    NASA Astrophysics Data System (ADS)

    Parker, R. R.

    2000-03-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events.

  19. Design Analysis and Manufacturing Studies for ITER In-Vessel Coils

    SciTech Connect

    Kalish, M.; Heitzenroeder, P.; Neumeyer, C.; Titus, P.; Zhai, Y.; Zatz, I.; Messineo, M.; Gomez, M.; Hause, C.; Daly, E.; Martin, A.; Wu, Y.; Jin, J.; Long, F.; Song, Y.; Wang, Z.; Yun, Zan; Hsiao, J.; Pillsbury, J. R.; Bohm, T.; Sawan, M.; Jiang, NFN

    2014-07-01

    ITER is incorporating two types of In Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required so that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are in close proximity to the plasma, mounted just behind the Blanket Shield Modules. This location results in a radiation and temperature environment that is severe necessitating new solutions for material selection as well as challenging analysis and design solutions. Fitting the coil systems in between the blanket shield modules and the vacuum vessel leads to difficult integration with diagnostic cabling and cooling water manifolds.

  20. DESIGN OF THE ITER IN-VESSEL COILS

    SciTech Connect

    Neumeyer, C; Bryant, L; Chrzanowski, J; Feder, R; Gomez, M; Heitzenroeder, P; Kalish, M; Lipski, A; Mardenfeld, M; Simmons, R; Titus, P; Zatz, I; Daly, E; Martin, A; Nakahira, M; Pillsbury, R; Feng, J; Bohm, T; Sawan, M; Stone, H; Griffiths, I; Schaffer, M

    2010-11-27

    The ITER project is considering the inclusion of two sets of in-vessel coils, one to mitigate the effect of Edge Localized Modes (ELMs) and another to provide vertical stabilization (VS). The in-vessel location (behind the blanket shield modules, mounted to the vacuum vessel inner wall) presents special challenges in terms of nuclear radiation (~3000 MGy) and temperature (100oC vessel during operations, 200oC during bakeout). Mineral insulated conductors are well suited to this environment but are not commercially available in the large cross section required. An R&D program is underway to demonstrate the production of mineral insulated (MgO or Spinel) hollow copper conductor with stainless steel jacketing needed for these coils. A preliminary design based on this conductor technology has been developed and is presented herein.

  1. Runaway electrons and ITER

    NASA Astrophysics Data System (ADS)

    Boozer, Allen H.

    2017-05-01

    The potential for damage, the magnitude of the extrapolation, and the importance of the atypical—incidents that occur once in a thousand shots—make theory and simulation essential for ensuring that relativistic runaway electrons will not prevent ITER from achieving its mission. Most of the theoretical literature on electron runaway assumes magnetic surfaces exist. ITER planning for the avoidance of halo and runaway currents is focused on massive-gas or shattered-pellet injection of impurities. In simulations of experiments, such injections lead to a rapid large-scale magnetic-surface breakup. Surface breakup, which is a magnetic reconnection, can occur on a quasi-ideal Alfvénic time scale when the resistance is sufficiently small. Nevertheless, the removal of the bulk of the poloidal flux, as in halo-current mitigation, is on a resistive time scale. The acceleration of electrons to relativistic energies requires the confinement of some tubes of magnetic flux within the plasma and a resistive time scale. The interpretation of experiments on existing tokamaks and their extrapolation to ITER should carefully distinguish confined versus unconfined magnetic field lines and quasi-ideal versus resistive evolution. The separation of quasi-ideal from resistive evolution is extremely challenging numerically, but is greatly simplified by constraints of Maxwell’s equations, and in particular those associated with magnetic helicity. The physics of electron runaway along confined magnetic field lines is clarified by relations among the poloidal flux change required for an e-fold in the number of electrons, the energy distribution of the relativistic electrons, and the number of relativistic electron strikes that can be expected in a single disruption event.

  2. Iterative Magnetometer Calibration

    NASA Technical Reports Server (NTRS)

    Sedlak, Joseph

    2006-01-01

    This paper presents an iterative method for three-axis magnetometer (TAM) calibration that makes use of three existing utilities recently incorporated into the attitude ground support system used at NASA's Goddard Space Flight Center. The method combines attitude-independent and attitude-dependent calibration algorithms with a new spinning spacecraft Kalman filter to solve for biases, scale factors, nonorthogonal corrections to the alignment, and the orthogonal sensor alignment. The method is particularly well-suited to spin-stabilized spacecraft, but may also be useful for three-axis stabilized missions given sufficient data to provide observability.

  3. Searching with iterated maps

    PubMed Central

    Elser, V.; Rankenburg, I.; Thibault, P.

    2007-01-01

    In many problems that require extensive searching, the solution can be described as satisfying two competing constraints, where satisfying each independently does not pose a challenge. As an alternative to tree-based and stochastic searching, for these problems we propose using an iterated map built from the projections to the two constraint sets. Algorithms of this kind have been the method of choice in a large variety of signal-processing applications; we show here that the scope of these algorithms is surprisingly broad, with applications as diverse as protein folding and Sudoku. PMID:17202267

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

    SciTech Connect

    Kondo, K.; Fischer, U.; Klix, A.; Pereslavtsev, P.; Serikov, A.; Villari, R.

    2014-06-15

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

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

    NASA Astrophysics Data System (ADS)

    Kondo, K.; Fischer, U.; Klix, A.; Pereslavtsev, P.; Serikov, A.; Villari, R.

    2014-06-01

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

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

    SciTech Connect

    Kuroda, T.; Furuya, K.; Sato, S.

    1995-12-31

    A modular blanket/first wall has been proposed for a fusion experimental reactor, e.g., International Thermonuclear Experimental Reactor (ITER), with support ribs connecting to a strong back plate. For the connection method, a welding approach has been investigated. Welding and cutting tests of the support ribs have been performed with three types of test specimens; flat plate (200 mm x 400 mm), partial model (700 mm x 200 mm), and full-box model (600 mm x 1000 mm x 430 mm). The support ribs were made of type 316L austenitic stainless steel with the thickness of 50 mm in all these tests. The welding method applied to these tests was narrow gap TIG, and water jet for cutting. Through these tests, engineering data including optimum welding conditions, welding distortion, and welding/cutting speeds have been obtained. Transverse shrinkage was about 10 mm for the welding of 50 mm thick rib. However, the difference in distortion at the first wall surface was within 1--2 mm. Therefore, the blanket/first wall module can be installed with quite a high accuracy by taking into account the module moving to the back plate during the welding.

  7. ITER helium ash accumulation

    SciTech Connect

    Hogan, J.T.; Hillis, D.L.; Galambos, J.; Uckan, N.A. ); Dippel, K.H.; Finken, K.H. . Inst. fuer Plasmaphysik); Hulse, R.A.; Budny, R.V. . Plasma Physics Lab.)

    1990-01-01

    Many studies have shown the importance of the ratio {upsilon}{sub He}/{upsilon}{sub E} in determining the level of He ash accumulation in future reactor systems. Results of the first tokamak He removal experiments have been analysed, and a first estimate of the ratio {upsilon}{sub He}/{upsilon}{sub E} to be expected for future reactor systems has been made. The experiments were carried out for neutral beam heated plasmas in the TEXTOR tokamak, at KFA/Julich. Helium was injected both as a short puff and continuously, and subsequently extracted with the Advanced Limiter Test-II pump limiter. The rate at which the He density decays has been determined with absolutely calibrated charge exchange spectroscopy, and compared with theoretical models, using the Multiple Impurity Species Transport (MIST) code. An analysis of energy confinement has been made with PPPL TRANSP code, to distinguish beam from thermal confinement, especially for low density cases. The ALT-II pump limiter system is found to exhaust the He with maximum exhaust efficiency (8 pumps) of {approximately}8%. We find 1<{upsilon}{sub He}/{upsilon}{sub E}<3.3 for the database of cases analysed to date. Analysis with the ITER TETRA systems code shows that these values would be adequate to achieve the required He concentration with the present ITER divertor He extraction system.

  8. Runaway electrons and ITER

    NASA Astrophysics Data System (ADS)

    Boozer, Allen

    2016-10-01

    ITER planning for avoiding runaway damage depends on magnetic surface breakup in fast relaxations. These arise in thermal quenches and in the spreading of impurities from massive gas injection or shattered pellets. Surface breakup would prevent a runaway to relativistic energies were it not for non-intercepting flux tubes, which contain magnetic field lines that do not intercept the walls. Such tubes persist near the magnetic axis and in the cores of islands but must dissipate before any confining surfaces re-form. Otherwise, a highly dangerous situation arises. Electrons that were trapped and accelerated in these flux tubes can fill a large volume of stochastic field lines and serve as a seed for the transfer of the full plasma current to runaways. If the outer confining surfaces are punctured, as by a drift into the wall, then the full runaway inventory will be lost in a short pulse along a narrow flux tube. Although not part of ITER planning, currents induced in the walls by the fast magnetic relaxation could be used to passively prevent outer surfaces re-forming. If magnetic surface breakup can be avoided during impurity injection, the plasma current could be terminated in tens of milliseconds by plasma cooling with no danger of runaway. Support by DoE Office of Fusion Energy Science Grant De-FG02-03ER54696.

  9. Microtearing instability in ITER*

    NASA Astrophysics Data System (ADS)

    Wong, King-Lap; Mikkelsen, David; Budny, Robert; Breslau, Joshua

    2010-11-01

    Microtearing modes are found to be unstable in some regions of a simulated ITER H-mode plasma [1] with the GS2 code [2]. Modes with kρs>1 are in the interior (r/a˜0.65-0.85) while longer wavelength modes are in the pedestal region. This instability may keep the pedestal within the peeling-ballooning stability boundary [3]. Microtearing modes can produce stochastic magnetic field similar to RMP coils; they may have similar effects on ELMs by increasing the pedestal width. The possibility of using this technique for ELM mitigation in ITER is explored. We propose to use a deuterium gas jet to control the microtearing instability and the Chirikov parameter at the edge. Preliminary evaluation of its effectiveness will be presented and the limitations of the GS2 code will be discussed based on our understanding from NSTX [4]. *This work is supported by USDoE contract DE-AC02-09CH11466. [4pt] [1] R. V. Budny, Nucl. Fusion (2009)[0pt] [2] W. Dorland et al., Phys. Rev. Lett. (2000).[0pt] [3] P. B. Snyder et al.,Nucl. Fusion (2009).[0pt] [4] K. L. Wong et al., Phys. Rev. Lett. (2007).

  10. APT target/blanket design and thermal hydraulics

    SciTech Connect

    Cappiello, M.; Pitcher, E.; Pasamehmetoglu, K.

    1999-04-01

    The Accelerator Production of Tritium (APT) Target/Blanket (T/B) system is comprised of an assembly of tritium producing modules supported by control, heat removal, shielding and retargeting systems. The T/B assembly produces tritium using a high-energy proton beam, a tungsten/lead spallation neutron source and {sup 3}He gas as the tritium producing feedstock. For the nominal production mode, protons are accelerated to an energy of 1030 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons are expanded using a raster/expansion system to illuminate a 0.19m by 1.9m beam spot on the front face of a centrally located tungsten neutron source. A surrounding lead blanket produces additional neutrons from scattered high-energy particles. The tungsten neutron source consists of nested, Inconel-718 clad tungsten cylinders assembled in horizontal Inconel-718 tubes. Each tube contains up to 6 cylinders with annular flow channel gaps of 0.102 cm. These horizontal tubes are manifolded into larger diameter vertical inlet and outlet pipes, which provide coolant. The horizontal and vertical tubes make up a structure similar to that of rungs on a ladder. The entire tungsten neutron source consists of 11 such ladders separated into two modules, one containing five ladders and the other six. Ladders are separated by a 0.3 m void region to increase nucleon leakage. The peak thermal-hydraulic conditions in the tungsten neutron source occur in the second ladder from the front. Because tungsten neutron source design has a significant number of parallel flow channels, the limiting thermal-hydraulic parameter is the onset of significant void (OSV) rather than critical heat flux (CHF). A blanket region surrounds the tungsten neutron source. The lateral blanket region is approximately 120 cm thick and 400 cm high. Blanket material consists of lead, {sup 3}He gas, aluminum, and light-water coolant. The blanket region is subdivided into rows based on the local power

  11. WHC Maintenance

    NASA Image and Video Library

    2009-04-28

    ISS019-E-011464 (28 April 2009) --- Japan Aerospace Exploration Agency (JAXA) astronaut Koichi Wakata, Expedition 19/20 flight engineer, performs in-flight maintenance on the waste and hygiene compartment located in the Destiny laboratory of the International Space Station.

  12. WHC Maintenance

    NASA Image and Video Library

    2009-04-28

    ISS019-E-011471 (28 April 2009) --- Japan Aerospace Exploration Agency (JAXA) astronaut Koichi Wakata, Expedition 19/20 flight engineer, performs in-flight maintenance on the waste and hygiene compartment located in the Destiny laboratory of the International Space Station.

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

    NASA Astrophysics Data System (ADS)

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

    2011-10-01

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

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

    SciTech Connect

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

    2011-06-03

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

  15. Neutronic design of the APT Target/Blanket

    SciTech Connect

    Pitcher, E.J.; Russell, G.J.; Kidman, R.B.; Ferguson, P.D.

    1997-12-01

    The primary function of the Accelerator Production of Tritium Target/Blanket assembly is the safe and efficient production of tritium. The T/B accepts a 1.7-GeV, 100-mA proton beam and produces neutrons via the spallation process. These neutrons then react with {sup 3}He to produce tritium. Neutronic optimization of the T/B is achieved by efficiently using the proton beam to produce neutrons and then, once produced, assuring that they are captured mostly by {sup 3}He. This optimization must occur within the constraints imposed by engineering considerations such as heat flux limits, structural integrity, fabricability, and safe and reliable operation. The target/blanket achieves these goals with a neutron production rate that is 75% of that achievable with an ideal target, and a neutronic efficiency of 84%, leading to an overall tritium production rate that is 63% of the theoretical maximum.

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

    SciTech Connect

    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.

  17. Thermal creep of granular breeder materials in fusion blankets

    NASA Astrophysics Data System (ADS)

    Bühler, L.; Reimann, J.

    2002-12-01

    Continuum models for granular materials in fusion blankets are efficient tools for modeling of the nonlinear elastic behavior of pebble beds, granular particle flow caused by shear, volume compaction and hardening. The present paper describes how the material models used in finite element analyses can be extended in order to account additionally for thermally activated creep. The derived material model gives results which are in reasonable accordance with experimental data for pebble beds.

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

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

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

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

    SciTech Connect

    Not Available

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  20. Heat Loads Due to Small Penetrations in Multilayer Insulation Blankets

    NASA Technical Reports Server (NTRS)

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

    2017-01-01

    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 each 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 Fouriers 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.

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

    SciTech Connect

    Not Available

    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.

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

    SciTech Connect

    Not Available

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  3. Efficacy of the blizzard blanket or blizzard blanket plus thermal angel in preventing hypothermia in a hemorrhagic shock victim (Sus scrofa) under operational conditions.

    PubMed

    Bridges, Elizabeth; Schmelz, Joseph; Evers, Karen

    2007-01-01

    The prevention of hypothermia in military casualties under field conditions is challenging. The efficacy of a baffled reflective Blanket (Blizzard Blanket), a portable intravenous fluid warmer (Thermal Angel), and wool Blankets (control) in preventing hypothermia was tested under military field conditions in a swine hemorrhagic shock model. Fifteen pigs were bled at 10 degrees C. After 45 minutes, Hextend was administered (groups 1 and 3, at 10 degrees C; group 2, via Thermal Angel); groups 2 and 3 were encircled with a Blizzard Blanket. After 120 minutes, the pigs were moved to 21 degrees C to simulate a field hospital; group 1 was covered with Blankets. Blood was administered (groups 1 and 3, at 4 degrees C; group 2, via Thermal Angel) with 180 minutes of monitoring. The core temperature was <35 degrees C in five of five control pigs, four of five Blizzard-only pigs, and one of five Thermal Angel plus Blizzard Blanket pigs. The Blizzard Blanket limited but did not prevent hypothermia. The Thermal Angel plus Blizzard Blanket combination prevented hypothermia. The Thermal Angel is useful for bolus administration when electricity is limited; its military field use is constrained by battery weight and battery life.

  4. Iterated crowdsourcing dilemma game

    PubMed Central

    Oishi, Koji; Cebrian, Manuel; Abeliuk, Andres; Masuda, Naoki

    2014-01-01

    The Internet has enabled the emergence of collective problem solving, also known as crowdsourcing, as a viable option for solving complex tasks. However, the openness of crowdsourcing presents a challenge because solutions obtained by it can be sabotaged, stolen, and manipulated at a low cost for the attacker. We extend a previously proposed crowdsourcing dilemma game to an iterated game to address this question. We enumerate pure evolutionarily stable strategies within the class of so-called reactive strategies, i.e., those depending on the last action of the opponent. Among the 4096 possible reactive strategies, we find 16 strategies each of which is stable in some parameter regions. Repeated encounters of the players can improve social welfare when the damage inflicted by an attack and the cost of attack are both small. Under the current framework, repeated interactions do not really ameliorate the crowdsourcing dilemma in a majority of the parameter space. PMID:24526244

  5. Iterated crowdsourcing dilemma game

    NASA Astrophysics Data System (ADS)

    Oishi, Koji; Cebrian, Manuel; Abeliuk, Andres; Masuda, Naoki

    2014-02-01

    The Internet has enabled the emergence of collective problem solving, also known as crowdsourcing, as a viable option for solving complex tasks. However, the openness of crowdsourcing presents a challenge because solutions obtained by it can be sabotaged, stolen, and manipulated at a low cost for the attacker. We extend a previously proposed crowdsourcing dilemma game to an iterated game to address this question. We enumerate pure evolutionarily stable strategies within the class of so-called reactive strategies, i.e., those depending on the last action of the opponent. Among the 4096 possible reactive strategies, we find 16 strategies each of which is stable in some parameter regions. Repeated encounters of the players can improve social welfare when the damage inflicted by an attack and the cost of attack are both small. Under the current framework, repeated interactions do not really ameliorate the crowdsourcing dilemma in a majority of the parameter space.

  6. Structural materials for ITER in-vessel component design

    NASA Astrophysics Data System (ADS)

    Kalinin, G.; Gauster, W.; Matera, R.; Tavassoli, A.-A. F.; Rowcliffe, A.; Fabritsiev, S.; Kawamura, H.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m 2 in the basic performance phase (BPP)) within a temperature range from 20 to 300°C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350°C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. Estimates of radiation damage at the locations for re-welding show that the dose will not exceed 0.05 dpa (with He generation of 1 appm) for the manifold and 0.01 dpa (with He generation 0.1 appm) for the back plate for the BPP of ITER operation. Existing experimental data show that these levels will not result in property changes for SS; however, neutron irradiation and He generation promote crack formation in the heat affected zone during welding. Cu based alloys, DS-Cu (Glidcop A125) and PHCu CuCrZr bronze) are proposed as a structural materials for high heat flux

  7. NOEL: a no-leak fusion blanket concept

    SciTech Connect

    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/sub 2/, LiPb, Pb) and fused salt choices for material A.

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

    NASA Technical Reports Server (NTRS)

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

    SciTech Connect

    Logan, B.G.

    1992-03-06

    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.

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

    NASA Technical Reports Server (NTRS)

    Hillard, G. Barry

    1993-01-01

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

  11. Structural analysis of the ITER vacuum vessel from disruption loading with halo asymmetry

    SciTech Connect

    Riemer, B.W.; Sayer, R.O.

    1996-07-01

    Static structural analyses of the ITER vacuum vessel were performed with toroidally asymmetric disruption loads. Asymmetric halo current conditions were assumed to modify symmetric disruption loads which resulted in net lateral loading on the vacuum vessel torus. Structural analyses with the asymmetric loading indicated significantly higher vessel stress and blanket support forces than with symmetric disruption loads. A recent change in the vessel support design which provided toroidal constraints at each mid port was found to be effective in reducing torus lateral movement and vessel stress.

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

    NASA Astrophysics Data System (ADS)

    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.

  13. Periodontal maintenance.

    PubMed

    Tan, A E S

    2009-09-01

    The main goal of periodontal therapy is to establish an oral environment compatible with periodontal health by the physical disruption of the plaque biofilm and adjunctive chemical means if required. Implicit in this objective is the ongoing requirement of detection and interception of new and recurrent disease, which continues at selected intervals for the life of the dentition after the initial ("active") phase of periodontal treatment. This concept of ongoing periodontal maintenance therapy has been embraced as the mandatory requirement for favourable periodontal outcomes based on institutional clinical trials and in practice-based studies in various parts of the world. This review examines the ramifications of periodontal maintenance therapy based upon a multi-level assessment of logistic issues and risk factors at three levels: (1) The patient level - treatment time; patient attendance compliance; and homecare measures, antiseptics/antibiotics and smoking. (2) The level of the individual tooth - tooth loss; and evaluation of success versus survival. (3) The level of each tooth surface ("site") - probing depth, loss of attachment and bleeding on probing; and changes in clinical attachment levels. In spite of the diversity of studies conducted, there is agreement on the efficacy of periodontal maintenance therapy when compared with studies on untreated populations and in treated cases that were not maintained.

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

    NASA Astrophysics Data System (ADS)

    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.

  15. 32 CFR Appendix C to Part 806b - DoD ‘Blanket Routine Uses’

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 32 National Defense 6 2013-07-01 2013-07-01 false DoD âBlanket Routine Usesâ C Appendix C to Part... PRIVACY ACT PROGRAM Pt. 806b, App. C Appendix C to Part 806b—DoD ‘Blanket Routine Uses’ Certain DoD “blanket routine uses” have been established that are applicable to every record system maintained by the...

  16. 32 CFR 701.112 - “Blanket routine uses.”

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 5 2011-07-01 2011-07-01 false âBlanket routine uses.â 701.112 Section 701.112... THE NAVY DOCUMENTS AFFECTING THE PUBLIC DON Privacy Program § 701.112 “Blanket routine uses.” In the interest of simplicity, economy, and to avoid redundancy, DOD has established “DOD Blanket Routine Uses...

  17. 32 CFR 701.112 - “Blanket routine uses.”

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 5 2012-07-01 2012-07-01 false âBlanket routine uses.â 701.112 Section 701.112... THE NAVY DOCUMENTS AFFECTING THE PUBLIC DON Privacy Program § 701.112 “Blanket routine uses.” In the interest of simplicity, economy, and to avoid redundancy, DOD has established “DOD Blanket Routine Uses...

  18. 32 CFR 701.112 - “Blanket routine uses.”

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 5 2010-07-01 2010-07-01 false âBlanket routine uses.â 701.112 Section 701.112... THE NAVY DOCUMENTS AFFECTING THE PUBLIC DON Privacy Program § 701.112 “Blanket routine uses.” In the interest of simplicity, economy, and to avoid redundancy, DOD has established “DOD Blanket Routine Uses...

  19. 32 CFR Appendix C to Part 806b - DoD ‘Blanket Routine Uses’

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 6 2011-07-01 2011-07-01 false DoD âBlanket Routine Usesâ C Appendix C to Part... PRIVACY ACT PROGRAM Pt. 806b, App. C Appendix C to Part 806b—DoD ‘Blanket Routine Uses’ Certain DoD “blanket routine uses” have been established that are applicable to every record system maintained by the...

  20. 32 CFR 701.112 - “Blanket routine uses.”

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 5 2014-07-01 2014-07-01 false âBlanket routine uses.â 701.112 Section 701.112... THE NAVY DOCUMENTS AFFECTING THE PUBLIC DON Privacy Program § 701.112 “Blanket routine uses.” In the interest of simplicity, economy, and to avoid redundancy, DOD has established “DOD Blanket Routine Uses...

  1. 32 CFR 701.112 - “Blanket routine uses.”

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 32 National Defense 5 2013-07-01 2013-07-01 false âBlanket routine uses.â 701.112 Section 701.112... THE NAVY DOCUMENTS AFFECTING THE PUBLIC DON Privacy Program § 701.112 “Blanket routine uses.” In the interest of simplicity, economy, and to avoid redundancy, DOD has established “DOD Blanket Routine Uses...

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

    PubMed

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

    2000-03-01

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

  3. Surge current and electron swarm tunnel tests of thermal blanket and ground strap materials

    NASA Technical Reports Server (NTRS)

    Hoffmaster, D. K.; Inouye, G. T.; Sellen, J. M., Jr.

    1977-01-01

    The results are described of a series of current conduction tests with a thermal control blanket to which grounding straps have been attached. The material and the ground strap attachment procedure are described. The current conduction tests consisted of a surge current examination of the ground strap and a dilute flow, energetic electron deposition and transport through the bulk of the insulating film of this thermal blanket material. Both of these test procedures were used previously with thermal control blanket materials.

  4. Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

    SciTech Connect

    Herman, H.; Baker, C.C.; Maroni, V.A.

    1981-10-01

    The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods.

  5. Mode conversion in ITER

    NASA Astrophysics Data System (ADS)

    Jaeger, E. F.; Berry, L. A.; Myra, J. R.

    2006-10-01

    Fast magnetosonic waves in the ion cyclotron range of frequencies (ICRF) can convert to much shorter wavelength modes such as ion Bernstein waves (IBW) and ion cyclotron waves (ICW) [1]. These modes are potentially useful for plasma control through the generation of localized currents and sheared flows. As part of the SciDAC Center for Simulation of Wave-Plasma Interactions project, the AORSA global-wave solver [2] has been ported to the new, dual-core Cray XT-3 (Jaguar) at ORNL where it demonstrates excellent scaling with the number of processors. Preliminary calculations using 4096 processors have allowed the first full-wave simulations of mode conversion in ITER. Mode conversion from the fast wave to the ICW is observed in mixtures of deuterium, tritium and helium3 at 53 MHz. The resulting flow velocity and electric field shear will be calculated. [1] F.W. Perkins, Nucl. Fusion 17, 1197 (1977). [2] E.F. Jaeger, L.A. Berry, J.R. Myra, et al., Phys. Rev. Lett. 90, 195001-1 (2003).

  6. Iterative marker excision system.

    PubMed

    Myronovskyi, Maksym; Rosenkränzer, Birgit; Luzhetskyy, Andriy

    2014-05-01

    The deletions of large genomic DNA fragments and consecutive gene knockouts are prerequisites for the generation of organisms with improved properties. One of the key issues in this context is the removal of antibiotic resistance markers from engineered organisms without leaving an active recombinase recognition site. Here, we report the establishment of an iterative marker excision system (IMES) that solves this problem. Based on the phiC31 integrase and its mutant att sites, IMES can be used for highly effective deletion of DNA fragments between inversely oriented B-CC and P-GG sites. The B-CC and P-GG sites are derived from attB and attP by substitution of the central core TT dinucleotide with CC and GG, respectively. An unnatural RR site that resides in the chromosome following deletion is the joining product of the right shoulders of B-CC and P-GG. We show that the RR sites do not recombine with each other as well as the RR site recombines with B-CC. The recombination efficiencies between RR and P-GG or RR and LL are only 0.1 % and 1 %, respectively. Thus, IMES can be used for multistep genomic engineering without risking unwanted DNA recombination. The fabrication of multi-purpose antibiotic cassettes and examples of the utilisation of IMES are described.

  7. ECE Diagnostics for ITER

    NASA Astrophysics Data System (ADS)

    Ellis, Richard; Austin, Max; Beno, Joseph; Rowan, William; Phillips, Perry; Hubbard, Amanda; Pandya, Hitesh; Feder, Russel

    2013-10-01

    ECE on ITER will be used to measure electron temperature profiles and non thermal features of the distribution. The diagnostic has two systems, one radial, and the other viewing at a small oblique angle. Radiation will be conducted to the diagnostic area with large smooth wall waveguide. Emission will be measured with a multichannel Michelson interferometer and two microwave radiometers which cover the fundamental and second harmonic ECE (X and O mode). In-situ calibration employs a hot calibration source which has been designed, constructed, and tested. We report extensive wideband transmission measurements made on the DIII-D Michelson corrugated waveguide system. We have now completed design of the beam splitter box which separates X and O modes for both views. The box inputs are now located flush up against the vacuum windows on the port plug. We have then redesigned the Gaussian beam optics of the system to reduce the size of the calibration sources by 20% to allow a better fit with other diagnostics in the port plug. We will present the details of the entire new design.

  8. Helium-cooled, FLiBe-breeder, beryllium-multiplier blanket for MINIMARS

    SciTech Connect

    Moir, R.W.; Lee, J.D.

    1986-06-01

    We adapted the helium-cooled, FLiBe-breeder blanket to the commercial tandem-mirror fusion-reactor design, MINIMARS. Vanadium was used to achieve high performance from the high-energy-release neutron-capture reactions and from the high-temperature operation permitted by the refractory property of the material, which increases the conversion efficiency and decreases the helium-pumping power. Although this blanket had the highest performance among the MINIMARS blankets designs, measured by Mn/sub th/ (blanket energy multiplication times thermal conversion efficiency), it had a cost of electricity (COE) 18% higher than the University of Wisconsin (UW) blanket design (42.5 vs 35.9 mills/kW.h). This increased cost was due to using higher-cost blanket materials (beryllium and vanadium) and a thicker blanket, which resulted in higher-cost central-cell magnets and the need for more blanket materials. Apparently, the high efficiency does not substantially affect the COE. Therefore, in the future, we recommend lowering the helium temperature so that ferritic steel can be used. This will result in a lower-cost blanket, which may compensate for the lower performance resulting from lower efficiency.

  9. Intense diagnostic neutral beam development for ITER

    SciTech Connect

    Rej, D.J.; Henins, I.; Fonck, R.J.; Kim, Y.J.

    1992-05-01

    For the next-generation, burning tokamak plasmas such as ITER, diagnostic neutral beams and beam spectroscopy will continue to be used to determine a variety of plasma parameters such as ion temperature, rotation, fluctuations, impurity content, current density profile, and confined alpha particle density and energy distribution. Present-day low-current, long-pulse beam technology will be unable to provide the required signal intensities because of higher beam attenuation and background bremsstrahlung radiation in these larger, higher-density plasmas. To address this problem, we are developing a short-pulse, intense diagnostic neutral beam. Protons or deuterons are accelerated using magnetic-insulated ion-diode technology, and neutralized in a transient gas cell. A prototype 25-kA, 100-kV, 1-{mu}s accelerator is under construction at Los Alamos. Initial experiments will focus on ITER-related issues of beam energy distribution, current density, pulse length, divergence, propagation, impurity content, reproducibility, and maintenance.

  10. Intense diagnostic neutral beam development for ITER

    SciTech Connect

    Rej, D.J.; Henins, I. ); Fonck, R.J.; Kim, Y.J. . Dept. of Nuclear Engineering and Engineering Physics)

    1992-01-01

    For the next-generation, burning tokamak plasmas such as ITER, diagnostic neutral beams and beam spectroscopy will continue to be used to determine a variety of plasma parameters such as ion temperature, rotation, fluctuations, impurity content, current density profile, and confined alpha particle density and energy distribution. Present-day low-current, long-pulse beam technology will be unable to provide the required signal intensities because of higher beam attenuation and background bremsstrahlung radiation in these larger, higher-density plasmas. To address this problem, we are developing a short-pulse, intense diagnostic neutral beam. Protons or deuterons are accelerated using magnetic-insulated ion-diode technology, and neutralized in a transient gas cell. A prototype 25-kA, 100-kV, 1-{mu}s accelerator is under construction at Los Alamos. Initial experiments will focus on ITER-related issues of beam energy distribution, current density, pulse length, divergence, propagation, impurity content, reproducibility, and maintenance.

  11. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    NASA Astrophysics Data System (ADS)

    Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2009-04-01

    The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

  12. Development of in-vessel components of the microfission chamber for ITER1

    PubMed Central

    Ishikawa, M.; Kondoh, T.; Ookawa, K.; Fujita, K.; Yamauchi, M.; Hayakawa, A.; Nishitani, T.; Kusama, Y.

    2010-01-01

    Microfission chambers (MFCs) will measure the total neutron source strength in ITER. The MFCs will be installed behind blanket modules in the vacuum vessel (VV). Triaxial mineral insulated (MI) cables will carry signals from the MFCs. The joint connecting triaxial MI cables in the VV must be considered because the MFCs and the MI cables will be installed separately at different times. Vacuum tight triaxial connector of the MI cable has been designed and a prototype has been constructed. Performance tests indicate that the connector can be applied to the ITER environment. A small bending-radius test of the MI cable indicates no observed damage at a curvature radius of 100 mm. PMID:21033834

  13. In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

    SciTech Connect

    L. C. Cadwallader

    2013-01-01

    The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

  14. The first fusion reactor: ITER

    NASA Astrophysics Data System (ADS)

    Campbell, D. J.

    2016-11-01

    Established by the signature of the ITER Agreement in November 2006 and currently under construction at St Paul-lez-Durance in southern France, the ITER project [1,2] involves the European Union (including Switzerland), China, India, Japan, the Russian Federation, South Korea and the United States. ITER (`the way' in Latin) is a critical step in the development of fusion energy. Its role is to provide an integrated demonstration of the physics and technology required for a fusion power plant based on magnetic confinement.

  15. Impact of prescribed burning on blanket peat hydrology

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

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

  16. Methadone maintenance.

    PubMed

    Russell-Taylor, W J

    1975-07-01

    Management of heroin addiction with oral methadone continues to create controversy. Since the first Dole-Nyswander experiments were reported almost a decade ago, methadone maintenance programs have been started in every major city in North America at an estimated public cost in the hundreds of millions of tax dollars. By the end of 1972, the United States alone spending at an annual rate in excess of $250 million, and an expenditure rate in excess of one billion dollars was occurring by the end of 1973. Despite the FDA investigational New Drug status of methadone when used for this purpose, objective data have still not been generated to substantiate the worth of Narcotic Substitution Therapy. Iatrogenic consequences are on the increase, and scepticism is growing among most pharmacologists. A critical evaluation of the entire methadone maintenance approach is being advocated by many and being attempted by some. Data currently being generated at the Philadelphia General Hospital suggest that many patients may have been better managed without methadone.

  17. Recovery of tritium from a liquid lithium blanket

    SciTech Connect

    Talbot, J.B.

    1981-01-01

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

  18. ORFEUS-SPAS MAIN TELESCOPE THERMAL BLANKETS ARE INSTALLED

    NASA Technical Reports Server (NTRS)

    1996-01-01

    German technicians in the Multi-Payload Processing Facility at KSC are installing thermal blankets on the main telescope of the Orbiting and Retrievable Far and Extreme Ultraviolet Spectrograph-Shuttle Pallet Satellite (ORFEUS- SPAS) II, mounted atop the Astronomy Shuttle Pallet Satellite (ASTRO-SPAS) platform. The main telescope houses two spectrographs, the Extreme Ultraviolet Spectrograph (EUV) and the Far Ultraviolet Spectrograph (FUV), which will gather data about the life cycle of stars during the flight of ORFEUS-SPAS II on Space Shuttle Mission STS-80 this fall. A third spectrograph, the Interstellar Medium Profile Spectrometer (IMAPS), will be installed on the side of the ASTRO-SPAS platform.

  19. Assessment of anti-fouling strategies for membrane coupled with upflow anaerobic sludge blanket (MUASB) process.

    PubMed

    Tran, Thao Minh; Ye, Yun; Chen, Vicki; Stuetz, Richard; Le-Clech, Pierre

    2013-01-01

    In this novel process, domestic wastewater was filtered by a hollow-fibre membrane coupled with an upflow anaerobic sludge blanket (MUASB) bioreactor. To improve the process sustainability and decrease energy costs, the membranes were operated under low fluxes with little, or no, shear. The efficiency of anti-fouling strategies, including relaxation, backwashing and supply of low aeration and stir rates, was assessed through detailed characterization of the fouling layers. Results indicated that backwashing was more efficient than relaxation, even when the systems were operated under the same flux productivity. In terms of shear supply, stir provided a better fouling limitation strategy compared to aeration, at similar shear stress values. Physical and chemical cleaning methods were applied to recover three fouling fractions (i.e. cake, residual and irreversible) for better characterization of the fouling layers. Under the sustainable operating conditions used in this study, most of the fouling was easily reversible by simple rinsing. In addition, permanent and irreversible fouling, resulting in the need for frequent chemical cleanings and potential membrane degradation, is limited once small shear stresses are applied. These outcomes are expected to form the basis for the future assessment of trade-off between operation, maintenance and replacement costs of membrane filtration processes used in wastewater treatment.

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

    PubMed

    Totani, Nagao; Inoue, Ryota; Yawata, Miho

    2016-06-01

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

  1. Performance of silvered Teflon (trademark) thermal control blankets on spacecraft

    NASA Technical Reports Server (NTRS)

    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.

  2. Forced-air patient warming blankets disrupt unidirectional airflow.

    PubMed

    Legg, A J; Hamer, A J

    2013-03-01

    We have recently shown that waste heat from forced-air warming blankets can increase the temperature and concentration of airborne particles over the surgical site. The mechanism for the increased concentration of particles and their site of origin remained unclear. We therefore attempted to visualise the airflow in theatre over a simulated total knee replacement using neutral-buoyancy helium bubbles. Particles were created using a Rocket PS23 smoke machine positioned below the operating table, a potential area of contamination. The same theatre set-up, warming devices and controls were used as in our previous study. This demonstrated that waste heat from the poorly insulated forced-air warming blanket increased the air temperature on the surgical side of the drape by > 5°C. This created convection currents that rose against the downward unidirectional airflow, causing turbulence over the patient. The convection currents increased the particle concentration 1000-fold (2 174 000 particles/m(3) for forced-air warming vs 1000 particles/m(3) for radiant warming and 2000 particles/m(3) for the control) by drawing potentially contaminated particles from below the operating table into the surgical site. Cite this article: Bone Joint J 2013;95-B:407-10.

  3. Axially staggered seed-blanket reactor fuel module construction

    DOEpatents

    Cowell, Gary K.; DiGuiseppe, Carl P.

    1985-01-01

    A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

  4. 78 FR 30911 - Texas Eastern Transmission, LP; Prior Notice Activity Under Blanket Certificate

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-23

    ... Energy Regulatory Commission Texas Eastern Transmission, LP; Prior Notice Activity Under Blanket Certificate On May 8, 2013, Texas Eastern Transmission, LP (Texas Eastern), filed a prior notice request... Act, and Texas Eastern's blanket certificate issued in Docket No. CP82-535-000. Texas Eastern seeks...

  5. 77 FR 38622 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-28

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on June 4, 2012, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700...) regulations under the Natural Gas Act as amended and Southern Star's blanket certificate issued in Docket No...

  6. Heat-pipe liquid-pool-blanket concept for the Tandem Mirror Reactor

    SciTech Connect

    Hoffman, M.A.; Werner, R.W.; Johnson, G.L.

    1981-10-01

    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.

  7. 77 FR 31004 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-24

    ...] Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on May 9, 2012, Southern Natural Gas Company (Southern), 569 Brookwood Village, Suite 501, Birmingham, Alabama 35209, filed... Commission's regulations under the Natural Gas Act (NGA), and Southern's blanket certificate issued in Docket...

  8. 75 FR 60095 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-29

    ... LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY..., by Sempra LNG Marketing, LLC (Sempra), requesting blanket authorization to export up to a total of..., Louisiana. Sempra is engaged in the business of purchasing and marketing supplies of LNG. Sempra is a...

  9. 77 FR 52713 - PetroLogistics Natural Gas Storage, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-30

    ... Energy Regulatory Commission PetroLogistics Natural Gas Storage, LLC; Notice of Request Under Blanket Authorization Take notice that on August 17, 2012, PetroLogistics Natural Gas Storage, LLC (PetroLogistics... Iberville Parish, Louisiana, under PetroLogistics' blanket certificate issued in Docket No. CP07-427-000, et...

  10. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 2 2012-07-01 2012-07-01 false DeCA Blanket Routine Uses C Appendix C to Part... Blanket Routine Uses (a) Routine Use—Law Enforcement. If a system of records maintained by a DoD Component... issued pursuant thereto, the relevant records in the system of records may be referred, as a routine use...

  11. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 2 2014-07-01 2014-07-01 false DeCA Blanket Routine Uses C Appendix C to Part... Blanket Routine Uses (a) Routine Use—Law Enforcement. If a system of records maintained by a DoD Component... issued pursuant thereto, the relevant records in the system of records may be referred, as a routine use...

  12. 7 CFR Exhibit B to Subpart A of... - Memorandum of Understanding and Blanket Commodity Lien Waiver

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 14 2010-01-01 2009-01-01 true Memorandum of Understanding and Blanket Commodity Lien...—Memorandum of Understanding and Blanket Commodity Lien Waiver The Farmers Home Administration or its... without unnecessarily impairing or undermining the respective security interests of FmHA or its successor...

  13. 78 FR 13663 - Equitrans, L.P. Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-28

    ... Energy Regulatory Commission Equitrans, L.P. Notice of Request Under Blanket Authorization Take notice that on February 12, 2013, Equitrans, L.P. (Equitrans), pursuant to the blanket certificate... open to public inspection. \\1\\ Equitrans, L.P., 85 FERC ] 61,089 (1998). Equitrans proposes to...

  14. 78 FR 36770 - ANR Pipeline Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-19

    ... Energy Regulatory Commission ANR Pipeline Company; Notice of Request Under Blanket Authorization Take that on May 31, 2013, ANR Pipeline Company (ANR), 717 Texas Street, Houston, Texas 77002-2761, filed in... Commission's Regulations under the Natural Gas Act, and ANR's blanket certificate issued in Docket No....

  15. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 18 Conservation of Power and Water Resources 1 2014-04-01 2014-04-01 false Blanket certificates for unbundled sales services. 284.284 Section 284.284 Conservation of Power and Water Resources... granted a blanket certificate of public convenience and necessity pursuant to section 7 of the Natural Gas...

  16. ITER Central Solenoid Module Fabrication

    SciTech Connect

    Smith, John

    2016-09-23

    The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first

  17. Effective Thermal Conductivity of Lithium Ceramic Pebble Beds for Fusion Blankets: A Review

    SciTech Connect

    Abou-Sena, A.; Ying, A.; Abdou, M.

    2005-05-15

    The use of lithium ceramic pebble beds has been considered in many blanket designs for the fusion reactors. Lithium ceramics have received a significant interest as tritium breeders for the fusion blankets during the last three decades. The thermal performance of the lithium ceramic pebble beds plays a key role for the fusion blankets. In order to study the heat transfer in the blanket, the effective thermal conductivity of the lithium ceramics pebble beds has to be well measured and characterized. The data of effective thermal conductivity of lithium ceramic pebble beds is important for the blanket design. Several studies have been dedicated to investigate the effective conductivity of the lithium ceramics pebble beds. The objective of this work is to review and compare the available data, presented by various studies, of effective conductivity of lithium ceramic pebble beds in order to address the current status of these data.

  18. Modelling of tritium transport in a pin-type solid breeder blanket

    SciTech Connect

    Martin, R.; Ghoniem, N.M.

    1986-02-01

    This study supplements a larger study of a solid breeder blanket design featuring lithium ceramic pins. This aspect of the study looks at tritium transport, release, and inventory within this blanket design. Li/sub 2/O and ..gamma..-LiAlO/sub 2/ are the two primary candidates for ceramic solid breeders. ..gamma..-LiAlO/sub 2/ was chosen for this blanket design due to its higher structural stability. Analysis of tritium behavior in solid breeder blankets is of great importance due to its impact on several critical issues: the generation of an adequate amount of fusion fuel, the safety-related issue of keeping radioactive blanket inventories as low as possible, and the release, purge, and economical processing of the bred tritium without undue contamination of the coolant and other reactor structures.

  19. ITER safety challenges and opportunities

    SciTech Connect

    Piet, S.J.

    1991-01-01

    Results of the Conceptual Design Activity (CDA) for the International Thermonuclear Experimental Reactor (ITER) suggest challenges and opportunities. ITER is capable of meeting anticipated regulatory dose limits,'' but proof is difficult because of large radioactive inventories needing stringent radioactivity confinement. We need much research and development (R D) and design analysis to establish that ITER meets regulatory requirements. We have a further opportunity to do more to prove more of fusion's potential safety and environmental advantages and maximize the amount of ITER technology on the path toward fusion power plants. To fulfill these tasks, we need to overcome three programmatic challenges and three technical challenges. The first programmatic challenge is to fund a comprehensive safety and environmental ITER R D plan. Second is to strengthen safety and environment work and personnel in the international team. Third is to establish an external consultant group to advise the ITER Joint Team on designing ITER to meet safety requirements for siting by any of the Parties. The first of the three key technical challenges is plasma engineering -- burn control, plasma shutdown, disruptions, tritium burn fraction, and steady state operation. The second is the divertor, including tritium inventory, activation hazards, chemical reactions, and coolant disturbances. The third technical challenge is optimization of design requirements considering safety risk, technical risk, and cost. Some design requirements are now too strict; some are too lax. Fuel cycle design requirements are presently too strict, mandating inappropriate T separation from H and D. Heat sink requirements are presently too lax; they should be strengthened to ensure that maximum loss of coolant accident temperatures drop.

  20. Summary report for ITER Task-T19: MHD pressure drop and heat transfer study for liquid metal systems

    SciTech Connect

    Reed, C.B.; Hua, T.Q.; Natesan, K.; Kirillov, I.R.; Vitkovski, I.V.; Anisimov, A.M.

    1995-03-01

    A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the question of insulator coatings. Design calculations show that an electrically insulating layer is necessary to maintain an acceptably low MHD pressure drop. To begin experimental investigations of the MHD performance of candidate insulator materials and the technology for putting them in place, a new test section was prepared. Aluminum oxide was chosen as the first candidate insulating material because it may be used in combination with NaK in the ITER vacuum vessel and/or the divertor. Details on the methods used to produce the aluminum oxide layer as well as the microstructures of the coating and the aluminide sublayer are presented and discussed. The overall MHD pressure drop, local MHD pressure gradient, local transverse MHD pressure difference, and surface voltage distributions in both the circumferential and the axial directions are reported and discussed. The positive results obtained here for high-temperature NaK have two beneficial implications for ITER. First, since NaK may be used in the vacuum vessel and/or the divertor, these results support the design approach of using electrically insulating coatings to substantially reduce MHD pressure drop. Secondly, while Al{sub 2}O{sub 3}/SS is not the same coating/base material combination which would be used in the advanced blanket, this work nonetheless shows that it is possible to produce a viable insulating coating which is stable in contact with a high temperature alkali metal coolant.

  1. Fusion Power measurement at ITER

    SciTech Connect

    Bertalot, L.; Barnsley, R.; Krasilnikov, V.; Stott, P.; Suarez, A.; Vayakis, G.; Walsh, M.

    2015-07-01

    Nuclear fusion research aims to provide energy for the future in a sustainable way and the ITER project scope is to demonstrate the feasibility of nuclear fusion energy. ITER is a nuclear experimental reactor based on a large scale fusion plasma (tokamak type) device generating Deuterium - Tritium (DT) fusion reactions with emission of 14 MeV neutrons producing up to 700 MW fusion power. The measurement of fusion power, i.e. total neutron emissivity, will play an important role for achieving ITER goals, in particular the fusion gain factor Q related to the reactor performance. Particular attention is given also to the development of the neutron calibration strategy whose main scope is to achieve the required accuracy of 10% for the measurement of fusion power. Neutron Flux Monitors located in diagnostic ports and inside the vacuum vessel will measure ITER total neutron emissivity, expected to range from 1014 n/s in Deuterium - Deuterium (DD) plasmas up to almost 10{sup 21} n/s in DT plasmas. The neutron detection systems as well all other ITER diagnostics have to withstand high nuclear radiation and electromagnetic fields as well ultrahigh vacuum and thermal loads. (authors)

  2. Relaxation Criteria for Iterated Traffic Simulations

    NASA Astrophysics Data System (ADS)

    Kelly, Terence; Nagel, Kai

    Iterative transportation microsimulations adjust traveler route plans by iterating between a microsimulation and a route planner. At each iteration, the route planner adjusts individuals' route choices based on the preceding microsimulations. Empirically, this process yields good results, but it is usually unclear when to stop the iterative process when modeling real-world traffic. This paper investigates several criteria to judge relaxation of the iterative process, emphasizing criteria related to traveler decision-making.

  3. ITER Construction--Plant System Integration

    SciTech Connect

    Tada, E.; Matsuda, S.

    2009-02-19

    This brief paper introduces how the ITER will be built in the international collaboration. The ITER Organization plays a central role in constructing ITER and leading it into operation. Since most of the ITER components are to be provided in-kind from the member countries, integral project management should be scoped in advance of real work. Those include design, procurement, system assembly, testing, licensing and commissioning of ITER.

  4. Error Field Correction in ITER

    SciTech Connect

    Park, Jong-kyu; Boozer, Allen H.; Menard, Jonathan E.; Schaffer, Michael J.

    2008-05-22

    A new method for correcting magnetic field errors in the ITER tokamak is developed using the Ideal Perturbed Equilibrium Code (IPEC). The dominant external magnetic field for driving islands is shown to be localized to the outboard midplane for three ITER equilibria that represent the projected range of operational scenarios. The coupling matrices between the poloidal harmonics of the external magnetic perturbations and the resonant fields on the rational surfaces that drive islands are combined for different equilibria and used to determine an ordered list of the dominant errors in the external magnetic field. It is found that efficient and robust error field correction is possible with a fixed setting of the correction currents relative to the currents in the main coils across the range of ITER operating scenarios that was considered.

  5. Construction Safety Forecast for ITER

    SciTech Connect

    cadwallader, lee charles

    2006-11-01

    The International Thermonuclear Experimental Reactor (ITER) project is poised to begin its construction activity. This paper gives an estimate of construction safety as if the experiment was being built in the United States. This estimate of construction injuries and potential fatalities serves as a useful forecast of what can be expected for construction of such a major facility in any country. These data should be considered by the ITER International Team as it plans for safety during the construction phase. Based on average U.S. construction rates, ITER may expect a lost workday case rate of < 4.0 and a fatality count of 0.5 to 0.9 persons per year.

  6. ITER Disruption Mitigation System Design

    NASA Astrophysics Data System (ADS)

    Rasmussen, David; Lyttle, M. S.; Baylor, L. R.; Carmichael, J. R.; Caughman, J. B. O.; Combs, S. K.; Ericson, N. M.; Bull-Ezell, N. D.; Fehling, D. T.; Fisher, P. W.; Foust, C. R.; Ha, T.; Meitner, S. J.; Nycz, A.; Shoulders, J. M.; Smith, S. F.; Warmack, R. J.; Coburn, J. D.; Gebhart, T. E.; Fisher, J. T.; Reed, J. R.; Younkin, T. R.

    2015-11-01

    The disruption mitigation system for ITER is under design and will require injection of up to 10 kPa-m3 of deuterium, helium, neon, or argon material for thermal mitigation and up to 100 kPa-m3 of material for suppression of runaway electrons. A hybrid unit compatible with the ITER nuclear, thermal and magnetic field environment is being developed. The unit incorporates a fast gas valve for massive gas injection (MGI) and a shattered pellet injector (SPI) to inject a massive spray of small particles, and can be operated as an SPI with a frozen pellet or an MGI without a pellet. Three ITER upper port locations will have three SPI/MGI units with a common delivery tube. One equatorial port location has space for sixteen similar SPI/MGI units. Supported by US DOE under DE-AC05-00OR22725.

  7. Building an experience factory for maintenance

    NASA Technical Reports Server (NTRS)

    Valett, Jon D.; Condon, Steven E.; Briand, Lionel; Kim, Yong-Mi; Basili, Victor R.

    1994-01-01

    This paper reports the preliminary results of a study of the software maintenance process in the Flight Dynamics Division (FDD) of the National Aeronautics and Space Administration/Goddard Space Flight Center (NASA/GSFC). This study is being conducted by the Software Engineering Laboratory (SEL), a research organization sponsored by the Software Engineering Branch of the FDD, which investigates the effectiveness of software engineering technologies when applied to the development of applications software. This software maintenance study began in October 1993 and is being conducted using the Quality Improvement Paradigm (QIP), a process improvement strategy based on three iterative steps: understanding, assessing, and packaging. The preliminary results represent the outcome of the understanding phase, during which SEL researchers characterized the maintenance environment, product, and process. Findings indicate that a combination of quantitative and qualitative analysis is effective for studying the software maintenance process, that additional measures should be collected for maintenance (as opposed to new development), and that characteristics such as effort, error rate, and productivity are best considered on a 'release' basis rather than on a project basis. The research thus far has documented some basic differences between new development and software maintenance. It lays the foundation for further application of the QIP to investigate means of improving the maintenance process and product in the FDD.

  8. Short wavelength interferometer for ITER

    SciTech Connect

    Snider, R.T.; Carlstrom, T.N.

    1992-04-01

    There is a need for a real time, reliable density measurement compatible with the restricted access and radiation environment on ITER. Due to the large plasma path length, high density and field, refraction and Faraday rotation effects makes the use of contemporary long wavelength (>50{mu}m) interferometers impractical. In this paper we consider the design of a short wavelength vibration compensated interferometer which allows operation without a prohibitively large vibration isolated structure and permits the optics to be conveniently mounted directly in or on the tokamak. A density interferometer design for ITER incorporating a 10.6 {mu}m CO{sub 2} interferometer with vibration compensation provided by a 3. 39 {mu}m HeNe laser is discussed. The proposed interferometer design requires only a small intrusion into the ITER tokamak without a large support structure, refraction and Faraday rotation problems are avoided, and it provides a density resolution of at least 0.5%. Results are presented from an interferometer installed on the DIII-D tokamak incorporating essential elements of the proposed ITER design including 10.6 and 3.39 {mu}m lasers, a retro-reflector mounted on the vacuum wall of the DIII-D tokamak and real-time density feedback control. In this paper we consider a short wavelength interferometer design that incorporates vibration compensation for use on ITER. Our primary concern is to develop a interferometer design that will produce a reliable real time density monitor. We use the ITER conceptual design activity report as the basis of the design.

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

    SciTech Connect

    Bosia, G.; Ragona, R.

    2015-12-10

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

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

    NASA Astrophysics Data System (ADS)

    Bosia, G.; Ragona, R.

    2015-12-01

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

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

    SciTech Connect

    DeMuth, J. A.; Meier, W. R.; Jolodosky, A.; Frantoni, M.; Reyes, S.

    2015-10-02

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

  12. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    SciTech Connect

    Choi, B. William; Chiu, Ing L.

    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.

  13. Tailorable advanced blanket insulation using aluminoborosilicate and alumina batting

    NASA Technical Reports Server (NTRS)

    Calamito, Dominic P.

    1989-01-01

    Two types of Tailorable Advanced Blanket Insulation (TABI) flat panels for Advanced Space Transportation Systems were produced. Both types consisted of integrally woven, 3-D fluted core having parallel faces and connecting ribs of Nicalon yarns. The triangular cross section flutes of one type was filled with mandrels of processed Ultrafiber (aluminoborosilicate) stitchbonded Nextel 440 fibrous felt, and the second type wall filled with Saffil alumina fibrous felt insulation. Weaving problems were minimal. Insertion of the fragile insulation mandrels into the fabric flutes was improved by using a special insertion tool. An attempt was made to weave fluted core fabrics from Nextel 440 yarns but was unsuccessful because of the yarn's fragility. A small sample was eventually produced by an unorthodox weaving process and then filled with Saffil insulation. The procedures for setting up and weaving the fabrics and preparing and inserting insulation mandrels are discussed. Characterizations of the panels produced are also presented.

  14. Iterative Restoration Of Tomosynthetic Slices

    NASA Astrophysics Data System (ADS)

    Ruttimann, U. E.; Groenhuis, R. A.; Webber, R. L.

    1984-08-01

    Tomosynthetic reconstructions suffer from the disadvantage that blurred images of object detail lying outside the plane of interest are superimposed over the desired image of structures in the tomosynthetic plane. It is proposed to selectively reduce these undesired superimpositions by a constrained iterative restoration method. Sufficient conditions are derived ensuring the convergence of the iterations to the exact solution in the absence of noise and constraints. Although in practice the restoration process must be left incomplete because of noise and quantization artifacts, the experimental results demonstrate that for reasons of stability these convergence conditions must be satisfied.

  15. The real mission of ITER

    SciTech Connect

    Wurden, G A

    2009-01-01

    For future machines, the plasma stored energy is going up by factors of 20-40x, and plasma currents by 2-3x, while the surface to volume ratio is at the same time decreasing. Therefore the disruption forces, even for constant B, (which scale like IxB), and associated possible localized heating on machine components, are more severe. Notably, Tore Supra has demonstrated removal of more than 1 GJ of input energy, over nearly a 400 second period. However, the instantaneous stored energy in the Tore Supra system (which is most directly related to the potential for disruption damage) is quite small compared to other large tokamaks. The goal of ITER is routinely described as studying DT burning plasmas with a Q {approx} 10. In reality, ITER has a much more important first order mission. In fact, if it fails at this mission, the consequences are that ITER will never get to the eventual stated purpose of studying a burning plasma. The real mission of ITER is to study (and demonstrate successfully) plasma control with {approx}10-17 MA toroidal currents and {approx}100-400 MJ plasma stored energy levels in long-pulse scenarios. Before DT operation is ever given a go-ahead in ITER, the reality is that ITER must demonstrate routine and reliable control of high energy hydrogen (and deuterium) plasmas. The difficulty is that ITER must simultaneously deal with several technical problems: (1) heat removal at the plasma/wall interface, (2) protection of the wall components from off-normal events, and (3) generation of dust/redeposition of first wall materials. All previous tokamaks have encountered hundred's of major disruptions in the course of their operation. The consequences of a few MA of runaway electrons (at 20-50 MeV) being generated in ITER, and then being lost to the walls are simply catastrophic. They will not be deposited globally, but will drift out (up, down, whatever, depending on control system), and impact internal structures, unless 'ameliorated'. Basically, this

  16. Iterated binomial sums and their associated iterated integrals

    NASA Astrophysics Data System (ADS)

    Ablinger, J.; Blümlein, J.; Raab, C. G.; Schneider, C.

    2014-11-01

    We consider finite iterated generalized harmonic sums weighted by the binomial binom{2k}{k} in numerators and denominators. A large class of these functions emerges in the calculation of massive Feynman diagrams with local operator insertions starting at 3-loop order in the coupling constant and extends the classes of the nested harmonic, generalized harmonic, and cyclotomic sums. The binomially weighted sums are associated by the Mellin transform to iterated integrals over square-root valued alphabets. The values of the sums for N → ∞ and the iterated integrals at x = 1 lead to new constants, extending the set of special numbers given by the multiple zeta values, the cyclotomic zeta values and special constants which emerge in the limit N → ∞ of generalized harmonic sums. We develop algorithms to obtain the Mellin representations of these sums in a systematic way. They are of importance for the derivation of the asymptotic expansion of these sums and their analytic continuation to N in {C}. The associated convolution relations are derived for real parameters and can therefore be used in a wider context, as, e.g., for multi-scale processes. We also derive algorithms to transform iterated integrals over root-valued alphabets into binomial sums. Using generating functions we study a few aspects of infinite (inverse) binomial sums.

  17. Significance of ITER IWS Material Selection and Qualification

    NASA Astrophysics Data System (ADS)

    Mehta, Bhoomi K.; Raval, Jigar; Maheshwari, Abha; Laad, Rahul; Singh, Gurlovleen; Pathak, Haresh

    2017-04-01

    In-Wall Shielding (IWS) is one of the important components of ITER Vacuum Vessel (VV) which fills the space between double walls of VV with cooling water. Procurement Arrangement (PA) for IWS has been signed with Indian Domestic Agency (INDA). Procurement of IWS materials, fabrication of IWS blocks and its delivery to respective Domestic Agency (DA) and ITER Organization (IO) are the main scope of this PA. Hence, INDIA is the only country which is contributing to VV IWS among all seven ITER partners. The main functions of the IWS are to provide Neutron Shielding with blanket, VV shells and water during plasma operations and to reduce ripple of the Toroidal Magnetic Field. To meet these functional requirements IWS blocks are made up of special materials (Borated Steels SS304 B4 & SS304 B7, Ferritic Steels SS 430, Austenitic Steel SS 316 L (N)-IG, XM-19 and Inconel-625) which are qualified, reliable and traceable for the design assessment. The choice of these materials has a significant influence on performance, maintainability, licensing, detailed design parameters and waste disposal. The main reasons for the materials selected for IWS are its high mechanical strength at operating temperatures, water chemistry properties, excellent fabrication characteristics and low cost relative to other similar materials. All the materials are qualified with respect to their respective codes (ASTM/EN standards with additional requirements as described in RCC-MR code 2007) and ITER requirements. Agreed Notified Body (ANB) has control conformity of materials certificates with approved material specification and traceability procedure for Safety Important Component (SIC). The procurement strategy for all the IWS materials has been developed in close collaboration with IO, ANB and Industries as per Product Procurement Specification (PPS). The R&D for sample, bulk material production, testing, inspection and handling as required are carried out by IN DA and IO. At present almost all

  18. Possible association of mucous blanket integrity with postirradiation colonization resistance

    SciTech Connect

    Walker, R.I.; Brook, I.; Costerton, J.W.; MacVittie, T.; Myhal, M.L.

    1985-12-01

    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.

  19. Maintenance Business Plans.

    ERIC Educational Resources Information Center

    Adams, Matt

    2002-01-01

    Discusses maintenance business plans, statements which provide accountability for facilities maintenance organizations' considerable budgets. Discusses the plan's components: statement of plan objectives, macro and detailed description of the facility assets, maintenance function descriptions, description of key performance indicators, milestone…

  20. Maintenance Management Update.

    ERIC Educational Resources Information Center

    Sternloff, Robert E.

    1987-01-01

    Current trends in park maintenance are overviewed, including maintenance impact statements, avoidance of cost through efficient use and national resource conservation, horticultural accomplishments that influence maintenance management, and vandalism prevention. (CB)

  1. Maintenance Management Update.

    ERIC Educational Resources Information Center

    Sternloff, Robert E.

    1987-01-01

    Current trends in park maintenance are overviewed, including maintenance impact statements, avoidance of cost through efficient use and national resource conservation, horticultural accomplishments that influence maintenance management, and vandalism prevention. (CB)

  2. Advanced methods comparisons of reaction rates in the Purdue Fast Breeder Blanket Facility

    SciTech Connect

    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.

  3. Thermal-hydraulic design of the target/blanket for the accelerator production of tritium conceptual design

    SciTech Connect

    Willcutt, G.J.E. Jr.; Kapernick, R.J.

    1997-11-01

    A conceptual design was developed for the target/blanket system of an accelerator-based system to produce tritium. The target/blanket system uses clad tungsten rods for a spallation target and clad lead rods as a neutron multiplier in a blanket surrounding the target. The neutrons produce tritium in {sup 3}He, which is contained in aluminum tubes located in the decoupler and blanket regions. This paper presents the thermal-hydraulic design of the target, decoupler, and blanket developed for the conceptual design of the Accelerator Production of Tritium Project, and demonstrates there is adequate margin in the design at full power operation.

  4. Transport simulation of ITER (International Thermonuclear Engineering Reactor) startup

    SciTech Connect

    Attenberger, S.E.; Houlberg, W.A.

    1989-01-01

    The present International Thermonuclear Engineering Reactor (ITER) reference configurations are the Technology Phase,'' in which the plasma current is maintained noninductively at a subignition density, and the Physics Phase,'' which is ignited but requires inductive maintenance of the current. The WHIST 1.5-D transport code is used to evaluate the volt-second requirements of both configurations. A slow current ramp (60-80's) is required for fixed-radius startup in ITER to avoid hollow current density profiles. To reach the operating point requires about 203 V{center dot}s for the Technology Phase (18 MA) and about 270 V{center dot}s for the Physics Phase (22 MA). The resistive losses can be reduced with expanding-radius startup. 5 refs., 4 figs.

  5. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    SciTech Connect

    Indah Rosidah, M. Suud, Zaki; Yazid, Putranto Ilham

    2015-09-30

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  6. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    NASA Astrophysics Data System (ADS)

    Indah Rosidah, M.; Suud, Zaki; Yazid, Putranto Ilham

    2015-09-01

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  7. Operating ITER Robustly Without Disruptions

    NASA Astrophysics Data System (ADS)

    Humphreys, D. A.; Eidietis, N. W.; Hyatt, A. W.; Leuer, J. A.; Luce, T. C.; Strait, E. J.; Walker, M. L.; Welander, A. S.; Wesley, J. C.; Lodestro, L.; Pearlstein, L. D.

    2011-10-01

    Disruptivity in ITER must be minimized to limit downtime and maximize use of the limited number of discharges. Minimizing disruptivity requires sufficient control capability, including robustness to disturbances and disruption avoidance through prediction of controllability limits. Robust control implies a balance of passively stable nominal scenarios, robust operation near or beyond open loop stability limits, and responses to off-normal events to avoid disruptive termination. Such a solution is possible because disruptions result from deterministic loss of controllability due to many proximal causes (e.g. loss of hardware resources, human error, or uncontrollable disturbances), most of which can be addressed with good physics models and known control methods. We illustrate the required approach with DIII-D experiments to assess ITER controllability and pre-qualify ITER scenarios, and with design and analysis ensuring sufficiently robust vertical control for ITER. Supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  8. Networking Theories by Iterative Unpacking

    ERIC Educational Resources Information Center

    Koichu, Boris

    2014-01-01

    An iterative unpacking strategy consists of sequencing empirically-based theoretical developments so that at each step of theorizing one theory serves as an overarching conceptual framework, in which another theory, either existing or emerging, is embedded in order to elaborate on the chosen element(s) of the overarching theory. The strategy is…

  9. Iterative method for interferogram processing

    NASA Astrophysics Data System (ADS)

    Kotlyar, Victor V.; Seraphimovich, P. G.; Zalyalov, Oleg K.

    1994-12-01

    We have developed and numerically evaluated an iterative algorithm for interferogram processing including the Fourier-transform method, the Gerchberg-Papoulis algorithm and Wiener's filter-based regularization used in combination. Using a signal-to-noise ratio not less than 1, it has been possible to reconstruct the phase of an object field with accuracy better than 5%.

  10. Neutronics experiments for uncertainty assessment of tritium breeding in HCPB and HCLL blanket mock-ups irradiated with 14 MeV neutrons

    NASA Astrophysics Data System (ADS)

    Batistoni, P.; Angelone, M.; Fischer, U.; Klix, A.; Kodeli, I.; Leichtle, D.; Pillon, M.; Pohorecki, W.; Villari, R.

    2012-08-01

    Two neutronics experiments have been carried out at 14 MeV neutron sources on mock-ups of the helium cooled pebble bed (HCBP) and the helium cooled lithium lead (HCLL) variants of ITER test blanket modules (TBMs). These experiments have provided an experimental validation of the calculations of the tritium production rate (TPR) in the two blanket concepts and an assessment of the uncertainties due to the uncertainties on nuclear data. This paper provides a brief summary of the HCPB experiment and then focuses in particular on the final results of the HCLL experiment. The TPR has been measured in the HCLL mock-up irradiated for long times at the Frascati 14 MeV Neutron Generator (FNG). Redundant and well-assessed experimental techniques have been used to measure the TPR by different teams for inter-comparison. Measurements of the neutron and gamma-ray spectra have also been performed. The analysis of the experiment, carried out by the MCNP code with FENDL-2.1 and JEFF-3.1.1 nuclear data libraries, and also including sensitivity/uncertainty analysis, shows good agreement between measurements and calculations, within the total uncertainty of 5.9% at 1σ level.

  11. First wall and blanket module safety enhancement by material selection and design decision

    SciTech Connect

    Merrill, B.J.

    1980-01-01

    A thermal/mechanical study has been performed which illustrates the behavior of a fusion reactor first wall and blanket module during a loss of coolant flow event. The relative safety advantages of various material and design options were determined. A generalized first wall-blanket concept was developed to provide the flexibility to vary the structural material (stainless steel vs titanium), coolant (helium vs water), and breeder material (liquid lithium vs solid lithium aluminate). In addition, independent vs common first wall-blanket cooling and coupled adjacent module cooling design options were included in the study. The comparative analyses were performed using a modified thermal analysis code to handle phase change problems.

  12. Conceptual design of fusion experimental reactor (FER/ITER)

    NASA Astrophysics Data System (ADS)

    Kimura, Haruyuki; Saigusa, Mikio; Saitoh, Yasushi

    1991-06-01

    Conceptual design of the Ion Cyclotron Wave (ICW) system for the FER and the Japanese contribution to the conceptual design of the International Thermonuclear Experimental Reactor (ITER) Ion Cyclotron Wave (ICW) system are presented. A frequency range of the FER ICW system is 50-85 MHz, which covers 2 omega (sub cT) heating, current drive by transit time magnetic pumping (TTMP) and 2 omega (sub cD) heating. Physics analyses show that the FER and the ITER ICW systems are suitable for the central ion heating and the burn control. The launching systems of the FER ICW system and the ITER high frequency ICW system are characterized by in-port plug and ridged-waveguide-fed 5x4 phased loop array. Merits of those systems are (1) a ceramic support is not necessary inside the cryostat and (2) remote maintenance of the front end part of the launcher is relatively easy. Overall structure of the launching system is consistent with radiation shielding, cooling, pumping, tritium safety and remote maintenance. The launcher has injection capability of 20 MW in the frequency range of 50-85 MHz with the separatrix-antenna distance of 15 cm and steep scrape-off density profile of H-mode. The shape of the ridged waveguide is optimized to provide desired frequency range and power handling capability with a finite element method. Matching between the current strap and the ridged waveguide is satisfactorily good. Thermal analysis of the Faraday shield shows that high electric conductivity low Z material such as beryllium should be chosen for a protection tile of the Faraday shield. A thick Faraday shield is necessary to tolerate electromagnetic force during disruptions. R and D needs for the ITER/FER ICW systems are identified and gain from JT-60/60U ICRF experiments and operations are indicated in connection with them.

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

    NASA Astrophysics Data System (ADS)

    Griscom, D. L.

    2012-02-01

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

  14. Eigenmode analysis of the ITER ICRF antenna plug and electrical solution to the grounding of the antenna

    SciTech Connect

    Louche, F.; Messiaen, A. M.; Dumortier, P.; Durodie, F.; Koch, R.; Lamalle, P. U.

    2009-11-26

    The excitation of resonant modes in the gap between the ITER ICRH antenna plug and the vessel can considerably increase the level of RF currents and voltages on the ITER plug and disturb the radiation characteristic of the array. We report on a study of these modes and a solution to avoid them in the ITER ion cyclotron range of frequencies. From a transmission line approach we show that resonances can be avoided if the distance between the mouth of the line and an added short-circuit is sufficiently small. We therefore propose to position short-circuits at about lm from the antenna mouth, constituted by rows of closely spaced contacts all around the plug. These conclusions are further validated with Microwave Studio simulations. A simplified model of the ITER ICRF array is used for comparing various grounding solutions and it is proved that the solution already inferred from transmission line approximation suppresses the resonances from the frequency domain relevant for ICRH in ITER. The MWS calculations predict that the proposed solution also avoids large RF currents in the blanket modules and in their connectors. Their amplitude barely exceeds 5% of the strap RF currents. This solution avoids perturbation of the antenna array impedance matrix by the coaxial gap.

  15. Grounds Maintenance Cost Report.

    ERIC Educational Resources Information Center

    Joice, Donald K.

    A grounds maintenance study was accomplished over the period of March 1965 through February 1966 to--(1) determine current grounds maintenance cost distribution; (2) provide basis for future grounds maintenance budgeting, and (3) provide basic data by which future landscape project designs can be developed to project or reduce maintenance costs.…

  16. Iteration and Prototyping in Creating Technical Specifications.

    ERIC Educational Resources Information Center

    Flynt, John P.

    1994-01-01

    Claims that the development process for computer software can be greatly aided by the writers of specifications if they employ basic iteration and prototyping techniques. Asserts that computer software configuration management practices provide ready models for iteration and prototyping. (HB)

  17. Nucleonics of a Be-Li-Th blanket for the fusion breeder

    SciTech Connect

    Lee, J.D.

    1983-03-28

    The nuclear performance of a candidate fission-suppressed, U233-producing blanket is assessed. It is predicted to have a breeding ratio (fusile + fissile) of 1.68 and produce U233 at a rate of 8030 kg/year from 3140 MW of DT fusion and a blanket coverage of 96%. Blanket energy multiplication is estimated to vary between 1.3 and 2.0 as the U233/Th232 ratio varies between 0 and 0.5%. Heterogeneous effects in the blanket's pebble-bed configuration were found to be important and more detailed analysis is needed to more accurately predict Li6 content required and U233 fission power versus U233 content.

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

    SciTech Connect

    Hamm, L.L.

    1998-10-07

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

  19. Role of Fabrication on Materials Compatibility in APT Target/Blanket

    SciTech Connect

    Iyer, N.; Louthan, M.R. Jr.; Dunn, K.; Fisher, D.L.

    1998-09-01

    This paper summarizes several of the options associated with the fabrication of selected target/blanket components. In addition, the materials characterization technologies required to validate these components performance is presented.

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

  1. Development and testing of a zero stitch MLI blanket using plastic pins for space use

    NASA Astrophysics Data System (ADS)

    Hatakenaka, Ryuta; Miyakita, Takeshi; Sugita, Hiroyuki; Saitoh, Masanori; Hirai, Tomoyuki

    2014-11-01

    New types of MLI blanket have been developed to achieve high thermal performance while maintaining production and assembly workability equivalent to the conventional type. Tag-pins, which are widely used in commercial applications to hook price tags to products, are used to fix the films in place and the pin material is changed to polyetheretherketone (PEEK) for use in space. Thermal performance is measured by using a boil-off calorimeter, in which a rectangular liquid nitrogen tank is used to evaluate the degradation at the bending corner and joint of the blanket. Zero-stitch- and multi-blanket-type MLIs show significantly improved thermal performance (ɛeff is smaller than 0.0050 at room temperature) despite having the same fastener interface as traditional blankets, while the venting design and number of tag-pins are confirmed as appropriate in a depressurization test.

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

    NASA Astrophysics Data System (ADS)

    Gallego-Sala, A. V.; Charman, D. J.; Harrison, S. P.; Li, G.; Prentice, I. C.

    2016-01-01

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

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

    NASA Astrophysics Data System (ADS)

    Gallego-Sala, A. V.; Charman, D. J.; Harrison, S. P.; Li, G.; Prentice, I. C.

    2015-10-01

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

  4. Colorado Conference on iterative methods. Volume 1

    SciTech Connect

    1994-12-31

    The conference provided a forum on many aspects of iterative methods. Volume I topics were:Session: domain decomposition, nonlinear problems, integral equations and inverse problems, eigenvalue problems, iterative software kernels. Volume II presents nonsymmetric solvers, parallel computation, theory of iterative methods, software and programming environment, ODE solvers, multigrid and multilevel methods, applications, robust iterative methods, preconditioners, Toeplitz and circulation solvers, and saddle point problems. Individual papers are indexed separately on the EDB.

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

    NASA Astrophysics Data System (ADS)

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

    2012-08-01

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

  6. 32 CFR Appendix C to Part 310 - DoD Blanket Routine Uses

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 2 2014-07-01 2014-07-01 false DoD Blanket Routine Uses C Appendix C to Part...) PRIVACY PROGRAM DOD PRIVACY PROGRAM Pt. 310, App. C Appendix C to Part 310—DoD Blanket Routine Uses (See paragraph (c) of § 310.22 of subpart E) A. Routine Use—Law Enforcement If a system of records maintained by...

  7. 32 CFR Appendix C to Part 310 - DoD Blanket Routine Uses

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 2 2012-07-01 2012-07-01 false DoD Blanket Routine Uses C Appendix C to Part...) PRIVACY PROGRAM DOD PRIVACY PROGRAM Pt. 310, App. C Appendix C to Part 310—DoD Blanket Routine Uses (See paragraph (c) of § 310.22 of subpart E) A. Routine Use—Law Enforcement If a system of records maintained by...

  8. Thermal and structural design aspects of high-temperature blankets for fusion synfuel production

    SciTech Connect

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

    1981-01-01

    The most promising process, high temperature electrolysis (HTE) of steam at temperatures of greater than or equal to 1000/sup 0/C is examined. In HTE, a large fraction (up to approx. 50%) of the energy input to split water to hydrogen and oxygen comes from thermal energy. For the projected operating conditions achieved by high temperature fusion blankets, overall efficiencies for hydrogen production should be on the order of 60%. The design, thermal-hydraulics, and materials for such blankets are discussed.

  9. Boiler tube failures: Case histories of steam blanketing and underdeposit corrosion

    SciTech Connect

    Hargrave, R.E.

    1994-12-31

    Case histories in which boiler tube failures resulted from a combination of steam blanketing and underdeposit corrosion are reviewed. Steam blanketing originates from inadequate circulation and commonly occurs in horizontal or sloped boiler tube circuits in the root of a boiler. It can lead to boiler tube corrosion failures by providing a mechanism for concentration of normally soluble species at or above a liquid/vapor interface.

  10. Boiler tube failures: Case histories of steam blanketing and underdeposit corrosion

    SciTech Connect

    Hargrave, R.E. )

    1994-08-01

    Case histories in which boiler tube failures resulted from a combination of steam blanketing and underdeposit corrosion are reviewed. Steam blanketing originates from inadequate circulation and commonly occurs in horizontal or sloped boiler tube circuits in the roof of a boiler. It can lead to boiler tube corrosion failures by providing a mechanism for concentration of normally soluble species at or above a liquid/vapor interface.

  11. Damaged thermal blanket noted on port OMS Pod on STS-117 Space Shuttle Atlantis

    NASA Image and Video Library

    2007-06-19

    S117-E-08982 (15 June 2007) --- Medium close-up scene of Atlantis' port side orbital maneuvering system pod prior to repair work performed by astronaut John (Danny) Olivas, STS-117 mission specialist (out of frame). Olivas later spent two hours stapling and pinning down this piece of thermal blanket on the pod. The 4-by-6-inch corner of the blanket peeled up during the shuttle's launch last week.

  12. Study of ITER plasma position reflectometer using a two-dimensional full-wave finite-difference time domain code

    SciTech Connect

    Silva, F. da

    2008-10-15

    The EU will supply the plasma position reflectometer for ITER. The system will have channels located at different poloidal positions, some of them obliquely viewing a plasma which has a poloidal density divergence and curvature, both adverse conditions for profile measurements. To understand the impact of such topology in the reconstruction of density profiles a full-wave two-dimensional finite-difference time domain O-mode code with the capability for frequency sweep was used. Simulations show that the reconstructed density profiles still meet the ITER radial accuracy specifications for plasma position (1 cm), except for the highest densities. Other adverse effects such as multireflections induced by the blanket, density fluctuations, and MHD activity were considered and a first understanding on their impact obtained.

  13. Bioinspired iterative synthesis of polyketides

    PubMed Central

    Zheng, Kuan; Xie, Changmin; Hong, Ran

    2015-01-01

    Diverse array of biopolymers and second metabolites (particularly polyketide natural products) has been manufactured in nature through an enzymatic iterative assembly of simple building blocks. Inspired by this strategy, molecules with inherent modularity can be efficiently synthesized by repeated succession of similar reaction sequences. This privileged strategy has been widely adopted in synthetic supramolecular chemistry. Its value also has been reorganized in natural product synthesis. A brief overview of this approach is given with a particular emphasis on the total synthesis of polyol-embedded polyketides, a class of vastly diverse structures and biologically significant natural products. This viewpoint also illustrates the limits of known individual modules in terms of diastereoselectivity and enantioselectivity. More efficient and practical iterative strategies are anticipated to emerge in the future development. PMID:26052510

  14. Design of ITER Relief Lines

    NASA Astrophysics Data System (ADS)

    Shah, N.; Choukekar, K.; Jadon, M.; Sarkar, B.; Joshi, B.; Kanzaria, H.; Gehani, V.; Vyas, H.; Pandya, U.; Panjwani, R.; Badgujar, S.; Monneret, E.

    2017-02-01

    The ITER Cryogenic system is one of the most complex cryogenic systems in the world. It includes roughly 5 km of cryogenic transfer line (cryolines) having large number of layout singularities in terms of bends at odd angles and branches. The relief lines are particularly important cryolines as they collect the helium from outlet of all process safety valves of the cryogenic clients and transfers it back to cryoplant. The total length of ITER relief lines is around 1.6 km with process pipe size varying from DN 50 to DN 200. While some part of relief lines carries warm helium for the recovery system, most part of the relief line is vacuum jacketed cryoline which carries cold helium from the clients. The final detailed design of relief lines has been completed. The paper describes the major input data and constraints for design of relief lines, design steps, flexibility and structural analysis approach and major design outcome.

  15. Spectroscopic problems in ITER diagnostics

    NASA Astrophysics Data System (ADS)

    Lisitsa, V. S.; Bureyeva, L. A.; Kukushkin, A. B.; Kadomtsev, M. B.; Krupin, V. A.; Levashova, M. G.; Medvedev, A. A.; Mukhin, E. E.; Shurygin, V. A.; Tugarinov, S. N.; Vukolov, K. Yu

    2012-12-01

    Problems of spectroscopic diagnostics of ITER plasma are under consideration. Three types of diagnostics are presented: 1) Balmer lines spectroscopy in the edge and divertor plasmas; 2) Thomson scattering, 3) charge exchange recombination spectroscopy. The Zeeman-Stark structure of line shapes is discussed. The overlapping of isotopes H-D-T spectral line shapes are presented for the SOL and divertor conditions. The polarization measurements of H-alpha spectral lines for H-D mixture on T-10 tokamak are shown in order to separate Zeeman splitting in more details. The problem of plasma background radiation emission for Thomson scattering in ITER is discussed in details. The line shape of P-7 hydrogen spectral line having a wave length close to laser one is presented together with continuum radiation. The charge exchange recombination spectroscopy (CXRS) is discussed in details. The data on Dα, HeII and CVI measurements in CXRS experiments on T-10 tokamak are presented.

  16. I-mode for ITER?

    NASA Astrophysics Data System (ADS)

    Whyte, D. G.; Marmar, E.; Hubbard, A.; Hughes, J.; Dominguez, A.; Greenwald, M.

    2011-10-01

    I-mode is a recently explored confinement regime that features a temperature pedestal and H-mode energy confinement, yet with L-mode particle confinement and no density pedestal nor large ELMs. Experiments on Alcator C-Mod and ASDEX-Upgrade show this leads to a stationary collisionless pedestal that inherently does not require ELMs for core impurity and particle control, possibly making I-mode an attractive operating regime for ITER where ELM heat pulses are expected to surpass material limits. We speculate as to how I-mode could be obtained, maintained and exploited for the ITER burning plasma physics mission. Issues examined include I-mode topology and power threshold requirements, pedestal formation, density control, avoiding H-mode, and the response of I-mode to alpha self-heating. Key uncertainties requiring further investigation are identified. Supported by the US DOE Cooperative Agreement DE-FC02-99ER54512.

  17. Cryogenics maintenance strategy

    NASA Astrophysics Data System (ADS)

    Cruzat, Fabiola

    2012-09-01

    ALMA is an interferometer composed of 66 independent systems, with specific maintenance requirements for each subsystem. To optimize the observation time and reduce downtime maintenance, requirements are very demanding. One subsystem with high maintenance efforts is cryogenics and vacuum. To organize the maintenance, the Cryogenic and Vacuum department is using and implementing different tools. These are monitoring and problem reporting systems and CMMS. This leads to different maintenance approaches: Preventive Maintenance, Corrective Maintenance and Condition Based Maintenance. In order to coordinate activities with other departments the preventive maintenance schedule is kept as flexible as systems allow. To cope with unavoidable failures, the team has to be prepared to work under any condition with the spares on time. Computerized maintenance management system (CMMS) will help to manage inventory control for reliable spare part handling, the correct record of work orders and traceability of maintenance activities. For an optimized approach the department is currently evaluating where preventive or condition based maintenance applies to comply with the individual system demand. Considering the change from maintenance contracts to in-house maintenance will help to minimize costs and increase availability of parts. Due to increased number of system and tasks the cryo team needs to grow. Training of all staff members is mandatory, in depth knowledge must be built up by doing complex maintenance activities in the Cryo group, use of advanced computerized metrology systems.

  18. US ITER limiter module design

    SciTech Connect

    Mattas, R.F.; Billone, M.; Hassanein, A.

    1996-08-01

    The recent U.S. effort on the ITER (International Thermonuclear Experimental Reactor) shield has been focused on the limiter module design. This is a multi-disciplinary effort that covers design layout, fabrication, thermal hydraulics, materials evaluation, thermo- mechanical response, and predicted response during off-normal events. The results of design analyses are presented. Conclusions and recommendations are also presented concerning, the capability of the limiter modules to meet performance goals and to be fabricated within design specifications using existing technology.

  19. Iterative Methods for Parameter Estimation

    DTIC Science & Technology

    1990-12-01

    IMPULSE RESPONSE (FIR) SYSTEMS ............... 10 A. FIXED DATA ALGORITHMS .................... 10 1. Gauss- Seidel Method ....................... 10 2...potential to provide a less biased least squares solution than a correlation method formulation [Ref. 3]. A. FIXED DATA ALGORITHMS 1. Gauss- Seidel Method A...very simple and straightforward iterative algorithm is the Gauss- Seidel method [Ref. 7]. We drop the superscript M from aM for simplicity. Unless

  20. Immobilization effect of air-injected blanket (AIB) for abdomen fixation

    SciTech Connect

    Ko, Young Eun; Suh, Yelin; Ahn, Seung Do; Lee, Sang-wook; Shin, Seong Soo; Kim, Jong Hoon; Choi, Eun Kyung; Yi, Byong Yong

    2005-11-15

    A new device for reducing the amplitude of breathing motion by pressing a patient's abdomen using an air-injected blanket (AIB) for external beam radiation treatments has been designed and tested. The blanket has two layers sealed in all four sides similar to an empty pillow made of urethane. The blanket is spread over the patient's abdomen with both ends of the blanket fixed to the sides of the treatment couch or a baseboard. The inner side, or patient side, of the blanket is thinner and expands more than the outer side. When inflated, the blanket balloons and effectively puts an even pressure on the patient's abdomen. Fluoroscopic observation was performed to verify the usefulness of AIB for patients with lung, breast cancer, or abdominal cancers. Internal organ movement due to breathing was monitored and measured with and without AIB. With the help of AIB, the average range of diaphragm motion was reduced from 2.6 to 0.7 cm in the anterior-to-posterior direction and from 2.7 to 1.3 cm in the superior-to-inferior direction. The motion range in the right-to-left direction was negligible, for it was less than 0.5 cm. These initial testing demonstrated that AIB is useful for reducing patients' breathing motion in the thoracic and abdominal regions comfortably and consistently.

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

    SciTech Connect

    Gohar, Y.; Smith, D. L.

    1999-10-07

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

  2. Application of the heterogeneous (HG) blankets for controlling the base structure vibration levels

    NASA Astrophysics Data System (ADS)

    Gautam, Ashwini; Fuller, C. R.; Carneal, James

    2005-09-01

    This work presents an extensive analysis of the properties of the heterogeneous blankets (HGs) and their effectiveness in controlling the vibration of the base structures. The HG blankets act as a distributed vibration absorbers consisting of mass inhomogeneities inside a layer of porous media (acoustic foam). To asses the effectiveness of these HG blankets in controlling the vibration of the base structure (plate), detailed finite element (FE) models of the foam, the HG blanket, and the plate have been developed. The foam has been dicretized using the eight node hexahedral elements. The HG blanket model consists of the foam model with point masses attached to the nodes of the elements. The structural (plate) domain is discretized using four node rectangular plate elements. Each of the FE models has been individually validated by comparing the numerical results with their respective analytical and experimental results. The structural and the HG blanket FE models were then combined into a larger FE model comprised of a base plate with the HG treatment on its surface. The results from this numerical model have shown that there is a significant reduction in the vibration levels of the base plate due to the HG treatment on it.

  3. ITER Plasma Control System Development

    NASA Astrophysics Data System (ADS)

    Snipes, Joseph; ITER PCS Design Team

    2015-11-01

    The development of the ITER Plasma Control System (PCS) continues with the preliminary design phase for 1st plasma and early plasma operation in H/He up to Ip = 15 MA in L-mode. The design is being developed through a contract between the ITER Organization and a consortium of plasma control experts from EU and US fusion laboratories, which is expected to be completed in time for a design review at the end of 2016. This design phase concentrates on breakdown including early ECH power and magnetic control of the poloidal field null, plasma current, shape, and position. Basic kinetic control of the heating (ECH, ICH, NBI) and fueling systems is also included. Disruption prediction, mitigation, and maintaining stable operation are also included because of the high magnetic and kinetic stored energy present already for early plasma operation. Support functions for error field topology and equilibrium reconstruction are also required. All of the control functions also must be integrated into an architecture that will be capable of the required complexity of all ITER scenarios. A database is also being developed to collect and manage PCS functional requirements from operational scenarios that were defined in the Conceptual Design with links to proposed event handling strategies and control algorithms for initial basic control functions. A brief status of the PCS development will be presented together with a proposed schedule for design phases up to DT operation.

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

    SciTech Connect

    Humrickhouse, Paul Weston; Merrill, Brad Johnson

    2014-11-01

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

  5. Uncertainties Quantification of Effective Thermal Conductivity for Ceramic Fiber Blanket

    NASA Astrophysics Data System (ADS)

    Zhao, Shu-yuan; Li, Jian-jun; He, Xiao-dong

    2014-01-01

    In the present paper, a probabilistic propagation model for assessing the uncertainty of the effective thermal conductivity was developed based on a combined conduction and radiation heat transfer model of a ceramic fiber blanket composite. The Monte Carlo technique was used to cope with the uncertainties in the material density, radiative properties, and boundary temperatures observed in experimental tests. The calculated effective thermal-conductivity distribution for the sample was compared with the experimental measurements performed on multiple samples, and the predicted mean values were in good agreement with the measured data. The result validates the thermal predictive model and demonstrates the suitability of the stochastic model containing statistical distributions in the input variables. Statistical information also indicates that the uncertainty effect can be enlarged at high temperatures. Response sensitivity analysis between the random inputs and the effective thermal conductivity demonstrates that the randomness in the hot-side temperature, the cold-side temperature, and extinction coefficient of the sample has a significant influence on the variability of thermal-conductivity properties. The extinction coefficient becomes more and more important with an increase of temperature due to the dominant radiative heat transfer contribution at high temperature. The analysis provides good insight into the scattering control in the experimental measurement and theoretical prediction of the effective thermal conductivity of a ceramic fiber composite.

  6. Progress and critical issues for IFE blanket and chamber research

    SciTech Connect

    Abdou, M.; Kulcinski, G.L.; Latkowski, J.F.; Logan, B.G.; Meier, W.R.; Moir, R.W.; Nobile, A.; Peterson, P.F.; Petti, D.; Schultz, K.R.; Tillack, M.S.

    1999-06-23

    Advances in high gain target designs for Inertial Fusion Energy (IFE), and the initiation of construction of large megajoule-class laser facilities in the U.S. (National Ignition Facility) and France (Laser-Megajoule) capable of testing the requirements for inertial fusion ignition and propagating burn, have improved the prospects for IFE. Accordingly, there have recently been modest increases in the US fusion research program related to the feasibility of IFE. These research areas include heavy-ion accelerators, Krypton-Fluoride (KrF) gas lasers, diode-pumped, solid-state (DPSSL) lasers, IFE target designs for higher gains, feasibility of low cost IFE target fabrication and accurate injection, and long-lasting IFE fusion chambers and final optics. Since several studies of conceptual IFE power plant and driver designs were completed in 1992-1996 [1-5], U.S. research in the IFE blanket, chamber, and target technology areas has focused on the critical issues relating to the feasibility of IFE concepts towards the goal of achieving economically-competitive and environmentally-attractive fusion energy. This paper discusses the critical issues in these areas, and the approaches taken to address these issues. The U.S. research in these areas, called IFE Chamber and Target Technologies, is coordinated through the Virtual Laboratory for Technology (VLT) formed by the Department of Energy in December 1998.

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

    SciTech Connect

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

    1999-08-01

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

  8. Smoke from Canadian Fires Blankets Eastern U.S.

    NASA Technical Reports Server (NTRS)

    2002-01-01

    Smoke from multiple large wildfires in Quebec is blanketing the southern portion of the Canadian province and extending southward over the Great Lakes and eastern United States. This image was acquired by the Moderate Resolution Imaging Spectroradiometer (MODIS) on the Terra satellite on July 7, 2002, and shows dozens of active fire detections (red dots) east of James Bay at upper left. The enormous smoke plume is almost 200 miles wide where it enters the United States over the New York and Vermont state lines. The thick pall is affecting air quality in places well to the south, including New York, Baltimore, and Washington, D.C. The image shows the smoke drifting out over the Atlantic Ocean, and then curling back in over North Carolina (bottom right). On Sunday, July 7, the Canadian Interagency Forest Fire Center reported 15 new fires in Quebec in the preceding 24 hours, bringing the total to more than 40 fires in the region, at least 7 which were burning out of control. Most of the fires are believed to have been caused by lightning, more of which is expected on Monday. According to news reports, several hundred people remain evacuated from their homes. Image by Jesse Allen, NASA Earth Observatory, based upon data provided by the MODIS Land Rapid Response Team at NASA GSFC

  9. Iterative methods for mixed finite element equations

    NASA Technical Reports Server (NTRS)

    Nakazawa, S.; Nagtegaal, J. C.; Zienkiewicz, O. C.

    1985-01-01

    Iterative strategies for the solution of indefinite system of equations arising from the mixed finite element method are investigated in this paper with application to linear and nonlinear problems in solid and structural mechanics. The augmented Hu-Washizu form is derived, which is then utilized to construct a family of iterative algorithms using the displacement method as the preconditioner. Two types of iterative algorithms are implemented. Those are: constant metric iterations which does not involve the update of preconditioner; variable metric iterations, in which the inverse of the preconditioning matrix is updated. A series of numerical experiments is conducted to evaluate the numerical performance with application to linear and nonlinear model problems.

  10. Establishing maintenance performance indicators

    SciTech Connect

    Baca, B.

    1994-10-01

    Maintenance Performance Indicators (PI) specify where the maintenance department is and which direction it is going allowing for a quick and accurate assessment of the performance of the Maintenance Management Program (MMP). Establishing PI`s for the maintenance department will allow a measure of productivity and a means of feedback for methods improvement. Effective performance of the maintenance department directly effects plant profitability. Improvements in the quality and productivity of the maintenance work force will significantly reduce maintenance costs. The level of performance attained by the maintenance work force is usually guessed at. Guessing will not identify areas needing improvement or help to initiate a corrective action. Maintenance PI`s are required for maintenance departments whose goal is to control maintenance costs while increasing productivity. The application of basic statistical methods will allow a maintenance department to know where they are and which direction they are going. The data presented in this paper is a representation of indicators used in industry as well as developed indicators to establish a complete maintenance performance indicator program. The methodology used in developing this program can be used as a way to manage a cost effective maintenance management program.

  11. Modelling blanket peatland hydrology and Holocene peatland development in north-eastern Scotland.

    NASA Astrophysics Data System (ADS)

    Swinnen, Ward; Verstraeten, Gert; Broothaerts, Nils

    2017-04-01

    To study long-term peatland dynamics, several peatland models have been constructed in recent decades. Most modelling efforts have focussed on peat bogs, but for other peatland types, such as blanket peatlands, modelling studies are limited. Although blanket peatland is a rare ecosystem type on a global scale, 87 percent of the peat cover in the UK is of this type. Hillslope hydrology is fundamental to blanket peatland development and an improved representation and understanding of the relationships between climate, hydrology and peat growth is crucial to better understand the effects of environmental change on peatland evolution and the carbon balance. Here, a new spatially explicit process-based peat growth model is presented for blanket peatlands, which couples a detailed 2.5D-hillslope hydrology model with a peat accumulation and decomposition module. The resultant model allows to study the hillslope hydrology and blanket peatland development along topographically complex hillslopes over a Holocene timescale. Calibration and validation of the model parameters is based on a dataset of more than 250 peat thickness measurements along several hillslope transects and eight radiocarbon dated peat samples in the headwaters of the river Dee (Cairngorms National Park, north-eastern Scotland). The model results show that the topography-driven hillslope hydrology has a strong influence on the resultant peat development along the hillslope, stressing the need for spatial models in studying blanket peatlands. Model simulations for the studied area result in peat growth initiation dates situated mostly in the period 9000 - 7000 a BP, which corresponds largely to basal calibrated radiocarbon dates for peat deposits in central and north-eastern Scotland. The simulated blanket peat growth initiation occurs before the mid-Holocene forest cover decline. These results indicate that, for the studied area, the blanket peatland development is largely driven by the early

  12. Requirements for Thin Film Solar Arrays and the Development Status of a New Solar Cell Blanket Concept

    NASA Astrophysics Data System (ADS)

    Zwanenburg, R.; Kroon, M.

    2005-05-01

    A novel blanket concept for thin film solar cells has been developed by Dutch Space. This blanket design, called the "Matrix" (patent pending), is formed by just connecting the thin film solar cells to each other at their corner points with plastic connection strips. The typical mass is only 240 g/m2 with CIGS solar cells made on 25 μm titanium foil. Interconnects are established by a pressure contact between the solar cells. Blanket samples of 48 connected solar cells (partly with live prototype solar cells) have been made and will be tested soon for thermal cycling and acoustic noise environments. The Matrix blanket design appears to be a very attractive solution and can be used on solar array level for solar cell blankets tensioned in panel frames and for large fold-out blanket arrays using masts.

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

    SciTech Connect

    Jolodosky, A.; Fratoni, M.

    2015-09-22

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

  14. 48 CFR 8.405-3 - Blanket purchase agreements (BPAs).

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... considerations. (vi) Maintenance availability. (vii) Environmental and energy efficiency considerations. (viii... level of effort and the mix of labor proposed to perform, and for determining that the proposed price is... the level of effort and the mix of labor proposed to perform a specific task being ordered, and for...

  15. Composition Optimization of Lithium-Based Ternary Alloy Blankets for Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Jolodosky, Alejandra

    The goal of this dissertation is to examine the neutronic properties of a novel type of fusion reactor blanket material in the form of lithium-based ternary alloys. Pure liquid lithium, first proposed as a blanket for fusion reactors, is utilized as both a tritium breeder and a coolant. It has many attractive features such as high heat transfer and low corrosion properties, but most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns including degradation of the concrete containment structure. The work of this thesis began as a collaboration with Lawrence Livermore National Laboratory in an effort to develop a lithium-based ternary alloy that can maintain the beneficial properties of lithium while reducing the reactivity concerns. The first studies down-selected alloys based on the analysis and performance of both neutronic and activation characteristics. First, 3-D Monte Carlo calculations were performed to evaluate two main neutronics performance parameters for the blanket: tritium breeding ratio (TBR), and energy multiplication factor (EMF). Alloys with adequate results based on TBR and EMF calculations were considered for activation analysis. Activation simulations were executed with 50 years of irradiation and 300 years of cooling. It was discovered that bismuth is a poor choice due to achieving the highest decay heat, contact dose rates, and accident doses. In addition, it does not meet the waste disposal ratings (WDR). The straightforward approach to obtain Monte Carlo TBR and EMF results required 231 simulations per alloy and became computationally expensive, time consuming, and inefficient. Consequently, alternate methods were pursued. A collision history-based methodology recently developed for the Monte Carlo code Serpent, calculates perturbation effects on practically

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

    SciTech Connect

    Dunn, F. E.

    2000-12-07

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

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

    SciTech Connect

    Crosswait, Kenneth Mitchell

    1994-04-01

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

  18. Status of the ITER heating neutral beam system

    NASA Astrophysics Data System (ADS)

    Hemsworth, R.; Decamps, H.; Graceffa, J.; Schunke, B.; Tanaka, M.; Dremel, M.; Tanga, A.; DeEsch, H. P. L.; Geli, F.; Milnes, J.; Inoue, T.; Marcuzzi, D.; Sonato, P.; Zaccaria, P.

    2009-04-01

    The ITER neutral beam (NB) injectors are the first injectors that will have to operate under conditions and constraints similar to those that will be encountered in a fusion reactor. These injectors will have to operate in a hostile radiation environment and they will become highly radioactive due to the neutron flux from ITER. The injectors will use a single large ion source and accelerator that will produce 40 A 1 MeV D- beams for pulse lengths of up to 3600 s. Significant design changes have been made to the ITER heating NB (HNB) injector over the past 4 years. The main changes are: Modifications to allow installation and maintenance of the beamline components with an overhead crane. The beam source vessel shape has been changed and the beam source moved to allow more space for the connections between the 1 MV bushing and the beam source. The RF driven negative ion source has replaced the filamented ion source as the reference design. The ion source and extractor power supplies will be located in an air insulated high voltage (-1 MV) deck located outside the tokamak building instead of inside an SF6 insulated HV deck located above the injector. Introduction of an all metal absolute valve to prevent any tritium in the machine to escape into the NB cell during maintenance. This paper describes the status of the design as of December 2008 including the above mentioned changes. The very important power supply system of the neutral beam injectors is not described in any detail as that merits a paper beyond the competence of the present authors. The R&D required to realize the injectors described in this paper must be carried out on a dedicated neutral beam test facility, which is not described here.

  19. Cryogenic instrumentation for ITER magnets

    NASA Astrophysics Data System (ADS)

    Poncet, J.-M.; Manzagol, J.; Attard, A.; André, J.; Bizel-Bizellot, L.; Bonnay, P.; Ercolani, E.; Luchier, N.; Girard, A.; Clayton, N.; Devred, A.; Huygen, S.; Journeaux, J.-Y.

    2017-02-01

    Accurate measurements of the helium flowrate and of the temperature of the ITER magnets is of fundamental importance to make sure that the magnets operate under well controlled and reliable conditions, and to allow suitable helium flow distribution in the magnets through the helium piping. Therefore, the temperature and flow rate measurements shall be reliable and accurate. In this paper, we present the thermometric chains as well as the venturi flow meters installed in the ITER magnets and their helium piping. The presented thermometric block design is based on the design developed by CERN for the LHC, which has been further optimized via thermal simulations carried out by CEA. The electronic part of the thermometric chain was entirely developed by the CEA and will be presented in detail: it is based on a lock-in measurement and small signal amplification, and also provides a web interface and software to an industrial PLC. This measuring device provides a reliable, accurate, electromagnetically immune, and fast (up to 100 Hz bandwidth) system for resistive temperature sensors between a few ohms to 100 kΩ. The flowmeters (venturi type) which make up part of the helium mass flow measurement chain have been completely designed, and manufacturing is on-going. The behaviour of the helium gas has been studied in detailed thanks to ANSYS CFX software in order to obtain the same differential pressure for all types of flowmeters. Measurement uncertainties have been estimated and the influence of input parameters has been studied. Mechanical calculations have been performed to guarantee the mechanical strength of the venturis required for pressure equipment operating in nuclear environment. In order to complete the helium mass flow measurement chain, different technologies of absolute and differential pressure sensors have been tested in an applied magnetic field to identify equipment compatible with the ITER environment.

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

    SciTech Connect

    W. JOHNSON; R. RYDER; P. RITTENHOUSE

    2001-01-01

    design data needs to evolve as the design progresses, operating conditions are refined, and materials characteristics in the unique environment are established. This paper develops the relationship between the designers' data needs and the structural design criteria recently adopted for the Target Blanket System of the APT. The latter, the newly-developed APT Supplemental Structural Design Requirements (APT SSDR), was patterned after the design criteria developed for the International Thermonuclear Experimental (Fusion) Reactor (ITER). A summary description of the design rules based on the APT SSDR is presented, and the impact of these rules of changes in materials properties resulting from exposure in the APT proton/neutron irradiation environment are discussed.

  1. Sensitivity of blanket peat vegetation and hydrochemistry to local disturbances.

    PubMed

    Robroek, Bjorn J M; Smart, Richard P; Holden, Joseph

    2010-10-01

    At the ecosystem scale, peatlands can be extremely resilient to perturbations. Yet, they are very sensitive to local disturbances, especially mechanical perturbations (e.g. trampling). The effects of these disturbances on vegetation, and potential effects on hydrochemical conditions along the peat surface, however, are largely unknown. We used three research tracks (paths researchers use to access their study sites) differing in time of abandonment to investigate the impact of local disturbance (trampling) on the vegetation and its short-term (< or = 2 year) recovery in a flagship research blanket peatland. Additionally, we examined the effects of local disturbance on fluvial runoff events and the concentrations of dissolved organic carbon (DOC) and particulate organic carbon (POC) in runoff water. Local disturbance heavily impacted peat vegetation, resulting in large areas of scarred and churned peat. Recovery of vascular plants along abandoned tracks was slow, but a functional Sphagnum layer re-established after just one year. The absence of vegetation elicited an increase in the number of runoff events along the tracks, by which POC runoff from the tracks increased. POC concentrations were highest in the surface water from the recently abandoned track, while they were low in the runoff water from the track abandoned longest and the undisturbed control track. We attribute this to the relatively fast recovery of the Sphagnum vegetation. DOC concentrations did not differ significantly either spatially or temporally in surface runoff or soil solution waters. While at an ecosystem scale local disturbances may be negligible in terms of carbon loss, our data points to the need for further research on the potential long-term effects of local disturbance on the vegetation, and significant effects on local scale carbon fluxes. Moreover, the effects of disturbances could be long-lasting and their role on ecosystem processes should not be underestimated.

  2. Satellite Shows an "Arctic Blanket" Over the U.S.

    NASA Image and Video Library

    2017-09-27

    View detail image here: bit.ly/1bvJlaN Arctic air has surged into the U.S. pushing into the Southeastern states and dropping high temperatures there into the 20s with colder wind chills. This NOAA GOES-East satellite image was captured at 1445 UTC/9:45 a.m. EST on January 28, and between the clouds and the snow on the ground with cold air overhead, it appears as if much of the U.S. has been covered by an "Arctic Blanket." According to NOAA's National Weather Service (NWS), the Gulf coast states from southern Louisiana east to the Carolinas are facing a wintry mix of precipitation along the southern edge of the Arctic air. Meanwhile, NWS notes that wind chills throughout much of the central and eastern U.S. are in single and negative numbers during the day on January 28. The GOES-East satellite is managed and operated by NOAA. This image was created by the NASA/NOAA GOES Project at NASA's Goddard Space Flight Center in Greenbelt, Md. Rob Gutro NASA's Goddard Space Flight Center Credit: NOAA/NASA GOES Project NASA image use policy. NASA Goddard Space Flight Center enables NASA’s mission through four scientific endeavors: Earth Science, Heliophysics, Solar System Exploration, and Astrophysics. Goddard plays a leading role in NASA’s accomplishments by contributing compelling scientific knowledge to advance the Agency’s mission. Follow us on Twitter Like us on Facebook Find us on Instagram

  3. Iterates of maps with symmetry

    NASA Technical Reports Server (NTRS)

    Chossat, Pascal; Golubitsky, Martin

    1988-01-01

    Fixed-point bifurcation, period doubling, and Hopf bifurcation (HB) for iterates of equivariant mappings are investigated analytically, with a focus on HB in the presence of symmetry. An algebraic formulation for the hypotheses of the theorem of Ruelle (1973) is derived, and the case of standing waves in a system of ordinary differential equations with O(2) symmetry is considered in detail. In this case, it is shown that HB can lead directly to motion on an invariant 3-torus, with an unexpected third frequency due to drift of standing waves along the torus.

  4. The physics role of ITER

    SciTech Connect

    Rutherford, P.H.

    1997-04-01

    Experimental research on the International Thermonuclear Experimental Reactor (ITER) will go far beyond what is possible on present-day tokamaks to address new and challenging issues in the physics of reactor-like plasmas. First and foremost, experiments in ITER will explore the physics issues of burning plasmas--plasmas that are dominantly self-heated by alpha-particles created by the fusion reactions themselves. Such issues will include (i) new plasma-physical effects introduced by the presence within the plasma of an intense population of energetic alpha particles; (ii) the physics of magnetic confinement for a burning plasma, which will involve a complex interplay of transport, stability and an internal self-generated heat source; and (iii) the physics of very-long-pulse/steady-state burning plasmas, in which much of the plasma current is also self-generated and which will require effective control of plasma purity and plasma-wall interactions. Achieving and sustaining burning plasma regimes in a tokamak necessarily requires plasmas that are larger than those in present experiments and have higher energy content and power flow, as well as much longer pulse length. Accordingly, the experimental program on ITER will embrace the study of issues of plasma physics and plasma-materials interactions that are specific to a reactor-scale fusion experiment. Such issues will include (i) confinement physics for a tokamak in which, for the first time, the core-plasma and the edge-plasma are simultaneously in a reactor-like regime; (ii) phenomena arising during plasma transients, including so-called disruptions, in regimes of high plasma current and thermal energy; and (iii) physics of a radiative divertor designed for handling high power flow for long pulses, including novel plasma and atomic-physics effects as well as materials science of surfaces subject to intense plasma interaction. Experiments on ITER will be conducted by researchers in control rooms situated at major

  5. Facilities maintenance handbook

    NASA Technical Reports Server (NTRS)

    1991-01-01

    This handbook is a guide for facilities maintenance managers. Its objective is to set minimum facilities maintenance standards. It also provides recommendations on how to meet the standards to ensure that NASA maintains its facilities in a manner that protects and preserves its investment in the facilities in a cost-effective manner while safely and efficiently performing its mission. This handbook implements NMI 8831.1, which states NASA facilities maintenance policy and assigns organizational responsibilities for the management of facilities maintenance activities on all properties under NASA jurisdiction. It is a reference for facilities maintenance managers, not a step-by-step procedural manual. Because of the differences in NASA Field Installation organizations, this handbook does not assume or recommend a typical facilities maintenance organization. Instead, it uses a systems approach to describe the functions that should be included in any facilities maintenance management system, regardless of its organizational structure. For documents referenced in the handbook, the most recent version of the documents is applicable. This handbook is divided into three parts: Part 1 specifies common definitions and facilities maintenance requirements and amplifies the policy requirements contained in NMI 8831. 1; Part 2 provides guidance on how to meet the requirements of Part 1, containing recommendations only; Part 3 contains general facilities maintenance information. One objective of this handbook is to fix commonality of facilities maintenance definitions among the Centers. This will permit the application of uniform measures of facilities conditions, of the relationship between current replacement value and maintenance resources required, and of the backlog of deferred facilities maintenance. The utilization of facilities maintenance system functions will allow the Centers to quantitatively define maintenance objectives in common terms, prepare work plans, and

  6. Maintenance Management System

    DTIC Science & Technology

    1990-06-01

    maintenance opera- tions. d. Available national maintenance management system (MMS) software be utilized to develop the planning, organizing...portland cement concrete pavements to level and realign faulted areas between slabs or craks within the slab by grinding the high side. MAINTENANCE ITEM

  7. Asphalt in Pavement Maintenance.

    ERIC Educational Resources Information Center

    Asphalt Inst., College Park, MD.

    Maintenance methods that can be used equally well in all regions of the country have been developed for the use of asphalt in pavement maintenance. Specific information covering methods, equipment and terminology that applies to the use of asphalt in the maintenance of all types of pavement structures, including shoulders, is provided. In many…

  8. Defer Maintenance, Invite Disaster

    ERIC Educational Resources Information Center

    Bowman, William W.

    1977-01-01

    An AGB- and NACUBO-sponsored survey showed that "wish lists" are accumulating overdue major maintenance projects because energy costs are consuming physical plant budgets. Problem areas are discussed: budget "guesstimation," preventive maintenance, deferred maintenance inventory, the APPA accounting format, resource allocation,…

  9. Microform Reader Maintenance.

    ERIC Educational Resources Information Center

    Hall, Hal W.; Michaels, George H.

    1985-01-01

    Describes experiences in organizing a program of microform reader and reader/printer maintenance at Texas A & M's Sterling C. Evans Library and offers guidelines for regular machine maintenance and repair. Guidelines discussed relate to maintenance philosophy, general machine cleaning, troubleshooting, service contracts, supplies,…

  10. Light Vehicle Preventive Maintenance.

    ERIC Educational Resources Information Center

    Marine Corps Inst., Washington, DC.

    This correspondence course, originally developed for the Marine Corps, is designed to instruct students in the performance of preventive maintenance on motor vehicles. Instructional materials are presented in three chapters as follows: (1) Major Maintenance Areas (maintenance system, tires, batteries, cooling systems, and vehicle lubrication; (2)…

  11. Iterated Stretching of Viscoelastic Jets

    NASA Technical Reports Server (NTRS)

    Chang, Hsueh-Chia; Demekhin, Evgeny A.; Kalaidin, Evgeny

    1999-01-01

    We examine, with asymptotic analysis and numerical simulation, the iterated stretching dynamics of FENE and Oldroyd-B jets of initial radius r(sub 0), shear viscosity nu, Weissenberg number We, retardation number S, and capillary number Ca. The usual Rayleigh instability stretches the local uniaxial extensional flow region near a minimum in jet radius into a primary filament of radius [Ca(1 - S)/ We](sup 1/2)r(sub 0) between two beads. The strain-rate within the filament remains constant while its radius (elastic stress) decreases (increases) exponentially in time with a long elastic relaxation time 3We(r(sup 2, sub 0)/nu). Instabilities convected from the bead relieve the tension at the necks during this slow elastic drainage and trigger a filament recoil. Secondary filaments then form at the necks from the resulting stretching. This iterated stretching is predicted to occur successively to generate high-generation filaments of radius r(sub n), (r(sub n)/r(sub 0)) = square root of 2[r(sub n-1)/r(sub 0)](sup 3/2) until finite-extensibility effects set in.

  12. Challenges for Cryogenics at Iter

    NASA Astrophysics Data System (ADS)

    Serio, L.

    2010-04-01

    Nuclear fusion of light nuclei is a promising option to provide clean, safe and cost competitive energy in the future. The ITER experimental reactor being designed by seven partners representing more than half of the world population will be assembled at Cadarache, South of France in the next decade. It is a thermonuclear fusion Tokamak that requires high magnetic fields to confine and stabilize the plasma. Cryogenic technology is extensively employed to achieve low-temperature conditions for the magnet and vacuum pumping systems. Efficient and reliable continuous operation shall be achieved despite unprecedented dynamic heat loads due to magnetic field variations and neutron production from the fusion reaction. Constraints and requirements of the largest superconducting Tokamak machine have been analyzed. Safety and technical risks have been initially assessed and proposals to mitigate the consequences analyzed. Industrial standards and components are being investigated to anticipate the requirements of reliable and efficient large scale energy production. After describing the basic features of ITER and its cryogenic system, we shall present the key design requirements, improvements, optimizations and challenges.

  13. ETR/ITER systems code

    SciTech Connect

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L.

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  14. ITER Port Interspace Pressure Calculations

    SciTech Connect

    Carbajo, Juan J; Van Hove, Walter A

    2016-01-01

    The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB) of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.

  15. Communication-optimal iterative methods

    NASA Astrophysics Data System (ADS)

    Demmel, J.; Hoemmen, M.; Mohiyuddin, M.; Yelick, K.

    2009-07-01

    Data movement, both within the memory system of a single processor node and between multiple nodes in a system, limits the performance of many Krylov subspace methods that solve sparse linear systems and eigenvalue problems. Here, s iterations of algorithms such as CG, GMRES, Lanczos, and Arnoldi perform s sparse matrix-vector multiplications and Ω(s) vector reductions, resulting in a growth of Ω(s) in both single-node and network communication. By reorganizing the sparse matrix kernel to compute a set of matrix-vector products at once and reorganizing the rest of the algorithm accordingly, we can perform s iterations by sending O(log P) messages instead of Ω(s·log P) messages on a parallel machine, and reading the on-node components of the matrix A from DRAM to cache just once on a single node instead of s times. This reduces communication to the minimum possible. We discuss both algorithms and an implementation of GMRES on a single node of an 8-core Intel Clovertown. Our implementations achieve significant speedups over the conventional algorithms.

  16. Electric blanket use and breast cancer in the Nurses' Health Study.

    PubMed

    Laden, F; Neas, L M; Tolbert, P E; Holmes, M D; Hankinson, S E; Spiegelman, D; Speizer, F E; Hunter, D J

    2000-07-01

    Electric and magnetic fields (EMFs) have been hypothesized to increase the risk of breast cancer, and electric blankets represent an important source of exposure to EMFs. The authors examined the relation between electric blanket use and invasive breast cancer in the Nurses' Health Study. On the biennial questionnaire in 1992, 87,497 women provided information on this exposure during three consecutive time periods. In a prospective analysis with 301,775 person-years of follow-up through 1996 (954 cases), the relative risk for any electric blanket use was not elevated (relative risk (RR) = 1.08, 95% confidence interval (CI): 0.95, 1.24) after controlling for breast cancer risk factors. There was a weak association between breast cancer and electric blanket use at least 16 years before diagnosis and long-term use in age-adjusted analyses but not in multivariate models. In a retrospective analysis of 1,318,683 person-years of follow-up (2,426 cases), the multivariate relative risk associated with use before disease follow-up began was null (RR = 1.05, 95% CI: 0.95, 1.16). Similar results were obtained in analyses stratified by menopause and restricted to estrogen receptor-positive breast cancers. While 95% confidence intervals for these estimates did not exclude small risks, overall, results did not support an association between breast cancer risk and exposure to EMFs from electric blankets.

  17. Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

    SciTech Connect

    Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

    1980-09-01

    The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design.

  18. ITER neutral beam system US conceptual design

    SciTech Connect

    Purgalis, P.

    1990-09-01

    In this document we present the US conceptual design of a neutral beam system for International Thermonuclear Experimental Reactor (ITER). The design incorporates a barium surface conversion D{sup {minus}} source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to water-cooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules which can be removed for remote maintenance. The neutral beam system delivers 75 MW of DO at 1.3 MeV, into three ports with a total of 9 modules arranged in stacks of three modules per port . To increase reliability each module is designed to deliver up to 10 MW; this allows eight modules operating at partial capacity to deliver the required power in the event one module is out of service, and provides 20% excess capacity to improve availability. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 46.5 m from the torus centerline. Neutron shielding in the drift duct and neutralizer provides the added feature of limiting conductance and thus reducing gas flow to and from the torus.

  19. Experimental Evidence on Iterated Reasoning in Games.

    PubMed

    Grehl, Sascha; Tutić, Andreas

    2015-01-01

    We present experimental evidence on two forms of iterated reasoning in games, i.e. backward induction and interactive knowledge. Besides reliable estimates of the cognitive skills of the subjects, our design allows us to disentangle two possible explanations for the observed limits in performed iterated reasoning: Restrictions in subjects' cognitive abilities and their beliefs concerning the rationality of co-players. In comparison to previous literature, our estimates regarding subjects' skills in iterated reasoning are quite pessimistic. Also, we find that beliefs concerning the rationality of co-players are completely irrelevant in explaining the observed limited amount of iterated reasoning in the dirty faces game. In addition, it is demonstrated that skills in backward induction are a solid predictor for skills in iterated knowledge, which points to some generalized ability of the subjects in iterated reasoning.

  20. Preconditioned iterations to calculate extreme eigenvalues

    SciTech Connect

    Brand, C.W.; Petrova, S.

    1994-12-31

    Common iterative algorithms to calculate a few extreme eigenvalues of a large, sparse matrix are Lanczos methods or power iterations. They converge at a rate proportional to the separation of the extreme eigenvalues from the rest of the spectrum. Appropriate preconditioning improves the separation of the eigenvalues. Davidson`s method and its generalizations exploit this fact. The authors examine a preconditioned iteration that resembles a truncated version of Davidson`s method with a different preconditioning strategy.

  1. Preliminary Neutronics Analysis of the ITER Toroidal Interferometer and Polarimeter Diagnostic Corner Cube Retroreflectors

    SciTech Connect

    Tresemer, K. R.

    2015-07-01

    ITER is an international project under construction in France that will demonstrate nuclear fusion at a power plant-relevant scale. The Toroidal Interferometer and Polarimeter (TIP) Diagnostic will be used to measure the plasma electron line density along 5 laser-beam chords. This line-averaged density measurement will be input to the ITER feedback-control system. The TIP is considered the primary diagnostic for these measurements, which are needed for basic ITER machine control. Therefore, system reliability & accuracy is a critical element in TIP’s design. There are two major challenges to the reliability of the TIP system. First is the survivability and performance of in-vessel optics and second is maintaining optical alignment over long optical paths and large vessel movements. Both of these issues greatly depend on minimizing the overall distortion due to neutron & gamma heating of the Corner Cube Retroreflectors (CCRs). These are small optical mirrors embedded in five first wall locations around the vacuum vessel, corresponding to certain plasma tangency radii. During the development of the design and location of these CCRs, several iterations of neutronics analyses were performed to determine and minimize the total distortion due to nuclear heating of the CCRs. The CCR corresponding to TIP Channel 2 was chosen for analysis as a good middle-road case, being an average distance from the plasma (of the five channels) and having moderate neutron shielding from its blanket shield housing. Results show that Channel 2 meets the requirements of the TIP Diagnostic, but barely. These results suggest other CCRs might be at risk of exceeding thermal deformation due to nuclear heating.

  2. Looking beyond ITER: Toroidal concept improvement

    SciTech Connect

    Sheffield, J.

    1993-01-01

    The International Thermonuclear Experimental Reactor (ITER) will demonstrate ignition and undertake fusion reactor systems integration and testing. Its design will be based upon relatively conservative physics assumptions. More attractive reactor configurations than those based upon ITER physics and the ITER configuration appear possible. The Toroidal Physics Experiment (TPX) will study tokamak behavior under a variety of conditions which may allow operation of improved combinations of beta, transport, and recirculating power. However, TPX still retains a configuration similar to ITER. Two alternative approaches appear interesting, based upon recent experimental results: the very low aspect ratio tokamak, or spherical torus, and the stellarator.

  3. Looking beyond ITER: Toroidal concept improvement

    SciTech Connect

    Sheffield, J.

    1993-06-01

    The International Thermonuclear Experimental Reactor (ITER) will demonstrate ignition and undertake fusion reactor systems integration and testing. Its design will be based upon relatively conservative physics assumptions. More attractive reactor configurations than those based upon ITER physics and the ITER configuration appear possible. The Toroidal Physics Experiment (TPX) will study tokamak behavior under a variety of conditions which may allow operation of improved combinations of beta, transport, and recirculating power. However, TPX still retains a configuration similar to ITER. Two alternative approaches appear interesting, based upon recent experimental results: the very low aspect ratio tokamak, or spherical torus, and the stellarator.

  4. Warm Ocean Temperatures Blanket the Far-Western Pacific

    NASA Technical Reports Server (NTRS)

    2001-01-01

    These data, taken during a 10-day collection cycle ending March 9, 2001, show that above-normal sea-surface heights and warmer ocean temperatures(indicated by the red and white areas) still blanket the far-western tropical Pacific and much of the north (and south) mid-Pacific. Red areas are about 10centimeters (4 inches) above normal; white areas show the sea-surface height is between 14 and 32 centimeters (6 to 13 inches) above normal.

    This build-up of heat dominating the Western Pacific was first noted by TOPEX/Poseidon oceanographers more than two years ago and has outlasted the El Nino and La Nina events of the past few years. See: http://www.jpl.nasa.gov/elnino/990127.html . This warmth contrasts with the Bering Sea, Gulf of Alaska and tropical Pacific where lower-than-normal sea levels and cool ocean temperatures continue (indicated by blue areas). The blue areas are between 5 and 13centimeters (2 and 5 inches) below normal, whereas the purple areas range from 14 to 18 centimeters (6 to 7 inches) below normal. Actually, the near-equatorial ocean cooled through the fall of 2000 and into mid-winter and continues almost La Nina-like.

    Looking at the entire Pacific basin, the Pacific Decadal Oscillation's warm horseshoe and cool wedge pattern still dominates this sea-level height image. Most recent National Oceanic and Atmospheric Administration (NOAA) sea-surface temperature data also clearly illustrate the persistence of this basin-wide pattern. They are available at http://psbsgi1.nesdis.noaa.gov:8080/PSB/EPS/SST/climo.html

    The U.S.-French TOPEX/Poseidon mission is managed by JPL for NASA's Earth Science Enterprise, Washington, D.C. JPL is a division of the California Institute of Technology in Pasadena. For more information on the TOPEX/Poseidon project, see: http://topex-www.jpl.nasa.gov

  5. Warm Ocean Temperatures Blanket the Far-Western Pacific

    NASA Technical Reports Server (NTRS)

    2001-01-01

    These data, taken during a 10-day collection cycle ending March 9, 2001, show that above-normal sea-surface heights and warmer ocean temperatures(indicated by the red and white areas) still blanket the far-western tropical Pacific and much of the north (and south) mid-Pacific. Red areas are about 10centimeters (4 inches) above normal; white areas show the sea-surface height is between 14 and 32 centimeters (6 to 13 inches) above normal.

    This build-up of heat dominating the Western Pacific was first noted by TOPEX/Poseidon oceanographers more than two years ago and has outlasted the El Nino and La Nina events of the past few years. See: http://www.jpl.nasa.gov/elnino/990127.html . This warmth contrasts with the Bering Sea, Gulf of Alaska and tropical Pacific where lower-than-normal sea levels and cool ocean temperatures continue (indicated by blue areas). The blue areas are between 5 and 13centimeters (2 and 5 inches) below normal, whereas the purple areas range from 14 to 18 centimeters (6 to 7 inches) below normal. Actually, the near-equatorial ocean cooled through the fall of 2000 and into mid-winter and continues almost La Nina-like.

    Looking at the entire Pacific basin, the Pacific Decadal Oscillation's warm horseshoe and cool wedge pattern still dominates this sea-level height image. Most recent National Oceanic and Atmospheric Administration (NOAA) sea-surface temperature data also clearly illustrate the persistence of this basin-wide pattern. They are available at http://psbsgi1.nesdis.noaa.gov:8080/PSB/EPS/SST/climo.html

    The U.S.-French TOPEX/Poseidon mission is managed by JPL for NASA's Earth Science Enterprise, Washington, D.C. JPL is a division of the California Institute of Technology in Pasadena. For more information on the TOPEX/Poseidon project, see: http://topex-www.jpl.nasa.gov

  6. Program status 4. quarter -- FY 1995: ITER and technology

    SciTech Connect

    1995-10-24

    A scoping level analysis of heat exhaust for the DEMO divertor strawman design was performed. It was concluded that to reduce the surface heat flux to the divertor, a combination of impurities radiation, compression, and increase in SOL electron density will have to be used. GA completed the preparation on the description of the helium-cooled, lithium breeder and V-alloy/ferritic steel DEMO blanket design focusing on addressing the issues of materials compatibility and reliability of high pressure gas system. Using ferritic steel, a helium Li-breeder design can handle neutron wall loading in the range of 6 to 8 MW/M{sup 2}. With advanced material like V-alloy, neutron wall loading in the range of 9 to 11 MW/m{sup 2} can be handled by similar design. The design of all components for a combline antenna for JFT-2M has been completed. Completion of the fabrication is scheduled for 11/95. The 4in x 4in 110 GHz window was fabricated and delivered to CPI (Varian) for testing. The second DiMES sample with Be/W striped coatings was exposed to four DIII-D plasma shots. Deuterium areal density measurements by W. Wampler of SNL showed the results from Be/W and Mo/V coated DiMES samples. High uptake of about 2x10{sup 17} atoms/cm{sup 2} was recorded for V. Be and C were lower by a factor of two, and Mo and W were lowest by about a factor of ten. An ITER magnetic coil has been fabricated and shipped to Brookhaven awaiting time for radiation testing. High-Nicalon and Super-Nicalon SiC fibers were received from PNL. Fabrication of different composite samples with different coatings is proceeding.

  7. First wall structural analysis of the aqueous self-cooled blanket concept

    SciTech Connect

    O'Brien, D.A.; Steiner, D.; Embrechts, M.J.; Deutch, L.; Gierszewski, P.

    1986-11-01

    A recently proposed blanket concept using water coolant with dissolved lithium compounds for breeding employs water cooled first walls. Water cooled first walls for blankets have also been proposed for some solid breeder blankets. Design options for water cooled first walls are examined in this paper. Four geometries and three materials are analyzed for water coolant at 300/sup 0/C and 13.8 MPa (2000 psi). Maximum neutron wall loads (with surface heat loads being 25% of neutron wall load) are determined for each geometry and material combination. Of the materials studied, only vanadium alloy is found to be capable of withstanding high wall loads (>10MW/m/sup 2/ neutron and >2.5 MW/m/sup 2/ heat).

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

    NASA Astrophysics Data System (ADS)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

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

  9. Results of parametric neutronic studies for high-temperature blankets with breeding

    SciTech Connect

    Makowitz, H; Lazareth, O W; Powell, J R; Fillo, J A

    1981-01-01

    Previous conceptual studies have been extended to examine a wide range of possible blanket configurations. The overall goals of this study have been to achieve neutronically viable blanket combinations that obtain an average breeding ratio (BR) of at least 1.1 and a sufficiently high fraction of fusion energy, Q/sub FRAC//sup HI/ (20% HTE High Temperature Heat) in the HTE process steam. The present studies have converged on the most attractive blanket option. After a preliminary and a secondary round of parametric calculations, ZrO/sub 2/ was selected as the best ceramic material; and a shell design with a flux trap type of back breeding zone with a Zr or Graphite plug, was chosen as optimal for the HTE module.

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

    SciTech Connect

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

    2000-05-17

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

  11. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    SciTech Connect

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  12. Divertor interferometer diagnostic for ITER

    SciTech Connect

    Brower, D. L.; Deng, B. H.; Ding, W. X.

    2006-10-15

    In the harsh environment of the divertor region in ITER, plasmas spanning a huge density range from 10{sup 19} to 10{sup 22} m{sup -3} are anticipated making measurement of the electron density particularly challenging. For any reasonable wavelength choice, the total phase measured by a conventional two-color interferometer system is always >>2{pi} and therefore subject to fringe counting errors. This problem can be remedied by adding a polarimeter capability whereby the Cotton-Mouton effect is measured or by employing differential interferometry. Using either approach, the total phase is always <<2{pi}. The conceptual design of an interferometer system along with possible wavelength choices will be explored.

  13. Iterative phase retrieval without support.

    PubMed

    Wu, J S; Weierstall, U; Spence, J C H; Koch, C T

    2004-12-01

    An iterative phase retrieval method for nonperiodic objects has been developed from the charge-flipping algorithm proposed in crystallography. A combination of the hybrid input-output (HIO) algorithm and the flipping algorithm has greatly improved performance. In this combined algorithm the flipping algorithm serves to find the support (object boundary) dynamically, and the HIO part improves convergence and moves the algorithm out of local minima. It starts with a single intensity measurement in the Fourier domain and does not require a priori knowledge of the support in the image domain. This method is suitable for general image recovery from oversampled diffuse elastic x-ray and electron-diffraction intensities. The relationship between this algorithm and the output-output algorithm is elucidated.

  14. Iterative phase retrieval without support

    NASA Astrophysics Data System (ADS)

    Wu, J. S.; Weierstall, U.; Spence, J. C. H.; Koch, C. T.

    2004-12-01

    An iterative phase retrieval method for nonperiodic objects has been developed from the charge-flipping algorithm proposed in crystallography. A combination of the hybrid input-output (HIO) algorithm and the flipping algorithm has greatly improved performance. In this combined algorithm the flipping algorithm serves to find the support (object boundary) dynamically, and the HIO part improves convergence and moves the algorithm out of local minima. It starts with a single intensity measurement in the Fourier domain and does not require a priori knowledge of the support in the image domain. This method is suitable for general image recovery from oversampled diffuse elastic x-ray and electron-diffraction intensities. The relationship between this algorithm and the output-output algorithm is elucidated.

  15. Planning as an Iterative Process

    NASA Technical Reports Server (NTRS)

    Smith, David E.

    2012-01-01

    Activity planning for missions such as the Mars Exploration Rover mission presents many technical challenges, including oversubscription, consideration of time, concurrency, resources, preferences, and uncertainty. These challenges have all been addressed by the research community to varying degrees, but significant technical hurdles still remain. In addition, the integration of these capabilities into a single planning engine remains largely unaddressed. However, I argue that there is a deeper set of issues that needs to be considered namely the integration of planning into an iterative process that begins before the goals, objectives, and preferences are fully defined. This introduces a number of technical challenges for planning, including the ability to more naturally specify and utilize constraints on the planning process, the ability to generate multiple qualitatively different plans, and the ability to provide deep explanation of plans.

  16. Benchmarking ICRF simulations for ITER

    SciTech Connect

    R. V. Budny, L. Berry, R. Bilato, P. Bonoli, M. Brambilla, R.J. Dumont, A. Fukuyama, R. Harvey, E.F. Jaeger, E. Lerche, C.K. Phillips, V. Vdovin, J. Wright, and members of the ITPA-IOS

    2010-09-28

    Abstract Benchmarking of full-wave solvers for ICRF simulations is performed using plasma profiles and equilibria obtained from integrated self-consistent modeling predictions of four ITER plasmas. One is for a high performance baseline (5.3 T, 15 MA) DT H-mode plasma. The others are for half-field, half-current plasmas of interest for the pre-activation phase with bulk plasma ion species being either hydrogen or He4. The predicted profiles are used by seven groups to predict the ICRF electromagnetic fields and heating profiles. Approximate agreement is achieved for the predicted heating power partitions for the DT and He4 cases. Profiles of the heating powers and electromagnetic fields are compared.

  17. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  18. Concept for testing fusion first wall/blanket systems in existing nuclear facilities

    SciTech Connect

    Hsu, P.Y.S.; Bohn, T.S.; Deis, G.A.; Judd, J.L.; Longhurst, G.R.; Miller, L.G.; Millsap, D.A.; Scott, A.J.; Wessol, D.E.

    1980-10-01

    A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment (except the 14 MeV neutron component) employing an existing nuclear facility is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of a test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module.

  19. Size limitations for microwave cavity to simulate heating of blanket material in fusion reactor

    SciTech Connect

    Wolf, D.

    1987-01-01

    The power profile in the blanket material of a nuclear fusion reactor can be simulated by using microwaves at 200 MHz. Using these microwaves, ceramic breeder materials can be thermally tested to determine their acceptability as blanket materials without entering a nuclear fusion environment. A resonating cavity design is employed which can achieve uniform cross sectional heating in the plane transverse to the neutron flux. As the sample size increases in height and width, higher order modes, above the dominant mode, are propagated and destroy the approximation to the heating produced in a fusion reactor. The limits at which these modes develop are determined in the paper.

  20. Innovation for maintenance technology improvements

    NASA Technical Reports Server (NTRS)

    Shives, T. R. (Editor); Willard, W. A. (Editor)

    1982-01-01

    A group of 34 submitted entries (32 papers and 2 abstracts) from the 33rd meeting of the Mechanical Failures Prevention Group whose subject was maintenance technology improvement through innovation. Areas of special emphasis included maintenance concepts, maintenance analysis systems, improved maintenance processes, innovative maintenance diagnostics and maintenance indicators, and technology improvements for power plant applications.

  1. 76 FR 2093 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-12

    ... Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY... November 30, 2010, by Eni USA Gas Marketing LLC (Eni USA), requesting blanket authorization to export... purchasing and marketing supplies of natural gas and LNG. Eni USA is a customer of the Cameron Terminal in...

  2. Thermo-mechanical Modelling of Pebble Beds in Fusion Blankets and its Implementation by a Return-Mapping Algorithm

    SciTech Connect

    Gan, Yixiang; Kamlah, Marc

    2008-07-01

    In this investigation, a thermo-mechanical model of pebble beds is adopted and developed based on experiments by Dr. Reimann at Forschungszentrum Karlsruhe (FZK). The framework of the present material model is composed of a non-linear elastic law, the Drucker-Prager-Cap theory, and a modified creep law. Furthermore, the volumetric inelastic strain dependent thermal conductivity of beryllium pebble beds is taken into account and full thermo-mechanical coupling is considered. Investigation showed that the Drucker-Prager-Cap model implemented in ABAQUS can not fulfill the requirements of both the prediction of large creep strains and the hardening behaviour caused by creep, which are of importance with respect to the application of pebble beds in fusion blankets. Therefore, UMAT (user defined material's mechanical behaviour) and UMATHT (user defined material's thermal behaviour) routines are used to re-implement the present thermo-mechanical model in ABAQUS. An elastic predictor radial return mapping algorithm is used to solve the non-associated plasticity iteratively, and a proper tangent stiffness matrix is obtained for cost-efficiency in the calculation. An explicit creep mechanism is adopted for the prediction of time-dependent behaviour in order to represent large creep strains in high temperature. Finally, the thermo-mechanical interactions are implemented in a UMATHT routine for the coupled analysis. The oedometric compression tests and creep tests of pebble beds at different temperatures are simulated with the help of the present UMAT and UMATHT routines, and the comparison between the simulation and the experiments is made. (authors)

  3. Iterative methods for weighted least-squares

    SciTech Connect

    Bobrovnikova, E.Y.; Vavasis, S.A.

    1996-12-31

    A weighted least-squares problem with a very ill-conditioned weight matrix arises in many applications. Because of round-off errors, the standard conjugate gradient method for solving this system does not give the correct answer even after n iterations. In this paper we propose an iterative algorithm based on a new type of reorthogonalization that converges to the solution.

  4. An accelerated subspace iteration for eigenvector derivatives

    NASA Technical Reports Server (NTRS)

    Ting, Tienko

    1991-01-01

    An accelerated subspace iteration method for calculating eigenvector derivatives has been developed. Factors affecting the effectiveness and the reliability of the subspace iteration are identified, and effective strategies concerning these factors are presented. The method has been implemented, and the results of a demonstration problem are presented.

  5. Rater Variables Associated with ITER Ratings

    ERIC Educational Resources Information Center

    Paget, Michael; Wu, Caren; McIlwrick, Joann; Woloschuk, Wayne; Wright, Bruce; McLaughlin, Kevin

    2013-01-01

    Advocates of holistic assessment consider the ITER a more authentic way to assess performance. But this assessment format is subjective and, therefore, susceptible to rater bias. Here our objective was to study the association between rater variables and ITER ratings. In this observational study our participants were clerks at the University of…

  6. Colorado Conference on iterative methods. Volume 2

    SciTech Connect

    1994-12-31

    The conference provided a forum for many topics in iterative methods. Volume II presents sessions on these topics: nonsymmetric solvers, parallel computation, ODE solvers, multigrid and multilevel methods, applications, robust iterative methods, preconditioners, Toeplitz and circulant matrix solvers, and saddle point problems. Individual papers are indexed separately on the EDB.

  7. New concurrent iterative methods with monotonic convergence

    SciTech Connect

    Yao, Qingchuan

    1996-12-31

    This paper proposes the new concurrent iterative methods without using any derivatives for finding all zeros of polynomials simultaneously. The new methods are of monotonic convergence for both simple and multiple real-zeros of polynomials and are quadratically convergent. The corresponding accelerated concurrent iterative methods are obtained too. The new methods are good candidates for the application in solving symmetric eigenproblems.

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

    SciTech Connect

    Hamm, L.L.

    1998-10-07

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

  9. Industrial Maintenance Strategies

    SciTech Connect

    Sajjad Akbar

    2006-07-01

    Industrial plants have become more complex due to technological advancement. This has made the task of maintenance more difficult. The maintenance costs in terms of resources and downtime loss are so high that maintenance function has become a critical factor in a plant's profitability. Industry should devote as much forethought to the management of maintenance function as to production. Maintenance has grown from an art to a precise, technical engineering science. Planning, organizing scheduling and control of maintenance using modern techniques pays dividends in the form of reduced costs and increased reliability. The magnitude and the dimension of maintenance have multiplied due to development in the engineering technologies. Production cost and capacities are directly affected by the breakdown time. Total operating cost including the maintenance cost plays an important role in replacement dimension. The integrated system approach would bring forth the desired results of high maintenance standards. The standards once achieved and sustained, would add to the reliability of the plan and relieve heavy stresses and strains on the engineering logistic support. (author)

  10. Examining maintenance responsibilities.

    PubMed

    Lam, K C

    2001-06-01

    This paper has examined the important responsibilities of the two organisations involved in the provision of maintenance service for the vital building services in many of our highly serviced buildings. The issues raised could be put to beneficial use in both clients and maintenance providers. All in all, the clients should work closely with their maintenance providers. Engineering services in buildings will not perform satisfactorily and efficiently if both parties do not work together and understand the maintenance tasks based on a business partnering mode. Put forward is the view that the management of the activities involved in the operation and maintenance process is a "shared commitment/involvement" between the client and the maintenance provider. It is obvious that many factors can influence the continued effectiveness of a quality maintenance scheme set up by client and provider. Some of these factors are: Change in key personnel Updates in technology Amendments to engineering practice Implementation of legislative requirements Changes in operation by client or provider Change of use of building Passage of time These factors must be fully reviewed by both parties from time to time, and necessary actions taken. A cooperative team working relationship and improved communication should be fostered by the client and his provider for the best management of services maintenance. This arrangement will contribute to better building services systems with continuous improvement; improved value for clients and higher return for the maintenance provider.

  11. [Medical Equipment Maintenance Methods].

    PubMed

    Liu, Hongbin

    2015-09-01

    Due to the high technology and the complexity of medical equipment, as well as to the safety and effectiveness, it determines the high requirements of the medical equipment maintenance work. This paper introduces some basic methods of medical instrument maintenance, including fault tree analysis, node method and exclusive method which are the three important methods in the medical equipment maintenance, through using these three methods for the instruments that have circuit drawings, hardware breakdown maintenance can be done easily. And this paper introduces the processing methods of some special fault conditions, in order to reduce little detours in meeting the same problems. Learning is very important for stuff just engaged in this area.

  12. TRIPOLI-4® - MCNP5 ITER A-lite neutronic model benchmarking

    NASA Astrophysics Data System (ADS)

    Jaboulay, J.-C.; Cayla, P.-Y.; Fausser, C.; Lee, Y.-K.; Trama, J.-C.; Li-Puma, A.

    2014-06-01

    The aim of this paper is to present the capability of TRIPOLI-4®, the CEA Monte Carlo code, to model a large-scale fusion reactor with complex neutron source and geometry. In the past, numerous benchmarks were conducted for TRIPOLI-4® assessment on fusion applications. Experiments (KANT, OKTAVIAN, FNG) analysis and numerical benchmarks (between TRIPOLI-4® and MCNP5) on the HCLL DEMO2007 and ITER models were carried out successively. In this previous ITER benchmark, nevertheless, only the neutron wall loading was analyzed, its main purpose was to present MCAM (the FDS Team CAD import tool) extension for TRIPOLI-4®. Starting from this work a more extended benchmark has been performed about the estimation of neutron flux, nuclear heating in the shielding blankets and tritium production rate in the European TBMs (HCLL and HCPB) and it is presented in this paper. The methodology to build the TRIPOLI-4® A-lite model is based on MCAM and the MCNP A-lite model (version 4.1). Simplified TBMs (from KIT) have been integrated in the equatorial-port. Comparisons of neutron wall loading, flux, nuclear heating and tritium production rate show a good agreement between the two codes. Discrepancies are mainly included in the Monte Carlo codes statistical error.

  13. Development of laser beam injection system for the Edge Thomson Scattering (ETS) in ITER

    NASA Astrophysics Data System (ADS)

    Yatsuka, E.; Hatae, T.; Suitoh, S.; Ohara, M.; Hagita, K.; Inoue, K.; Bassan, M.; Walsh, M.; Itami, K.

    2016-01-01

    This paper focuses on the design and development of the laser injection system for the ITER Edge Thomson Scattering system (ETS). The ITER ETS achieves a temporal resolution of 100 Hz by firing two 50 Hz laser beams alternatively. The use of dual lasers enables us to perform the Thomson scattering measurements at a temporal resolution of 50 Hz in case that one of the laser systems stops functioning. A new type of beam combiner was developed to obtain a single beam that is collinear and fixed linearly polarized from two laser beams using a motor-driven rotating half-wave plate. The rotating half-wave plate method does not induce misalignment even if the rotating mechanism malfunctions. The combined beam is relayed from the diagnostic hall to the plasma using mirror optics and is absorbed at the beam dump integrated on the inner blanket. The beam alignment system was designed to direct the laser beam onto the center of the beam dump head. The beam position at the beam dump is monitored by four alignment laser beams which propagate parallel to the diagnostic Nd:YAG laser beam and imaging systems installed outside the diagnostic port.

  14. Fusion Materials Science and Technology Research Needs: Now and During the ITER era

    SciTech Connect

    Wirth, Brian D.; Kurtz, Richard J.; Snead, Lance L.

    2013-09-30

    The plasma facing components, first wall and blanket systems of future tokamak-based fusion power plants arguably represent the single greatest materials engineering challenge of all time. Indeed, the United States National Academy of Engineering has recently ranked the quest for fusion as one of the top grand challenges for engineering in the 21st Century. These challenges are even more pronounced by the lack of experimental testing facilities that replicate the extreme operating environment involving simultaneous high heat and particle fluxes, large time varying stresses, corrosive chemical environments, and large fluxes of 14-MeV peaked fusion neutrons. This paper will review, and attempt to prioritize, the materials research and development challenges facing fusion nuclear science and technology into the ITER era and beyond to DEMO. In particular, the presentation will highlight the materials degradation mechanisms we anticipate to occur in the fusion environment, the temperature- displacement goals for fusion materials and plasma facing components and the near and long-term materials challenges required for both ITER, a fusion nuclear science facility and longer term ultimately DEMO.

  15. Inter-code comparison benchmark between DINA and TSC for ITER disruption modelling

    NASA Astrophysics Data System (ADS)

    Miyamoto, S.; Isayama, A.; Bandyopadhyay, I.; Jardin, S. C.; Khayrutdinov, R. R.; Lukash, V. E.; Kusama, Y.; Sugihara, M.

    2014-08-01

    Results of 2D disruption modelling for validation of benchmark ITER scenarios using two established codes—DINA and TSC, are compared. Although the simulation models employed in those two codes ought to be equivalent in the resistive time scale, quite different defining equations and formulations are adopted in their approaches. Moreover there are considerable differences in the implemented model of solid conducting structures placed on the periphery of the plasma such as the vacuum vessel and blanket modules. Thus it has long been unanswered whether the one of the two codes is really able to reproduce the other's results correctly, since a large number of code-wise differences render the comparison task exceedingly complicated. In this paper, it is demonstrated that after the simulations are set up accounting for the model differences, a reasonably good agreement is generally obtained, corroborating the correctness of the code results. When the halo current generation and its poloidal path in the first wall are included, however, the situation is more complicated. Because of the surface averaged treatment of the magnetic field (current density) diffusion equation, DINA can only approximately handle the poloidal electric currents in the first wall that cross the field lines. Validation is carried out for DINA simulations of the halo current generation by comparing with TSC simulations, where the treatment of halo current dynamics is more justifiable. The specific details of each code, affecting the consequence in ITER disruption prediction, are highlighted and discussed.

  16. 47 CFR 25.138 - Blanket Licensing provisions of GSO FSS Earth Stations in the 18.3-18.8 GHz (space-to-Earth), 19...

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 47 Telecommunication 2 2011-10-01 2011-10-01 false Blanket Licensing provisions of GSO FSS Earth... Earth Stations § 25.138 Blanket Licensing provisions of GSO FSS Earth Stations in the 18.3-18.8 GHz... (Earth-to-space) bands. (a) All applications for a blanket earth station license in the GSO FSS in the 18...

  17. Iterants, Fermions and Majorana Operators

    NASA Astrophysics Data System (ADS)

    Kauffman, Louis H.

    Beginning with an elementary, oscillatory discrete dynamical system associated with the square root of minus one, we study both the foundations of mathematics and physics. Position and momentum do not commute in our discrete physics. Their commutator is related to the diffusion constant for a Brownian process and to the Heisenberg commutator in quantum mechanics. We take John Wheeler's idea of It from Bit as an essential clue and we rework the structure of that bit to a logical particle that is its own anti-particle, a logical Marjorana particle. This is our key example of the amphibian nature of mathematics and the external world. We show how the dynamical system for the square root of minus one is essentially the dynamics of a distinction whose self-reference leads to both the fusion algebra and the operator algebra for the Majorana Fermion. In the course of this, we develop an iterant algebra that supports all of matrix algebra and we end the essay with a discussion of the Dirac equation based on these principles.

  18. Magnetic fusion and project ITER

    SciTech Connect

    Park, H.K.

    1992-09-01

    It has already been demonstrated that our economics and international relationship are impacted by an energy crisis. For the continuing prosperity of the human race, a new and viable energy source must be developed within the next century. It is evident that the cost will be high and will require a long term commitment to achieve this goal due to a high degree of technological and scientific knowledge. Energy from the controlled nuclear fusion is a safe, competitive, and environmentally attractive but has not yet been completely conquered. Magnetic fusion is one of the most difficult technological challenges. In modem magnetic fusion devices, temperatures that are significantly higher than the temperatures of the sun have been achieved routinely and the successful generation of tens of million watts as a result of scientific break-even is expected from the deuterium and tritium experiment within the next few years. For the practical future fusion reactor, we need to develop reactor relevant materials and technologies. The international project called ``International Thermonuclear Experimental Reactor (ITER)`` will fulfill this need and the success of this project will provide the most attractive long-term energy source for mankind.

  19. On the interplay between inner and outer iterations for a class of iterative methods

    SciTech Connect

    Giladi, E.

    1994-12-31

    Iterative algorithms for solving linear systems of equations often involve the solution of a subproblem at each step. This subproblem is usually another linear system of equations. For example, a preconditioned iteration involves the solution of a preconditioner at each step. In this paper, the author considers algorithms for which the subproblem is also solved iteratively. The subproblem is then said to be solved by {open_quotes}inner iterations{close_quotes} while the term {open_quotes}outer iteration{close_quotes} refers to a step of the basic algorithm. The cost of performing an outer iteration is dominated by the solution of the subproblem, and can be measured by the number of inner iterations. A good measure of the total amount of work needed to solve the original problem to some accuracy c is then, the total number of inner iterations. To lower the amount of work, one can consider solving the subproblems {open_quotes}inexactly{close_quotes} i.e. not to full accuracy. Although this diminishes the cost of solving each subproblem, it usually slows down the convergence of the outer iteration. It is therefore interesting to study the effect of solving each subproblem inexactly on the total amount of work. Specifically, the author considers strategies in which the accuracy to which the inner problem is solved, changes from one outer iteration to the other. The author seeks the `optimal strategy`, that is, the one that yields the lowest possible cost. Here, the author develops a methodology to find the optimal strategy, from the set of slowly varying strategies, for some iterative algorithms. This methodology is applied to the Chebychev iteration and it is shown that for Chebychev iteration, a strategy in which the inner-tolerance remains constant is optimal. The author also estimates this optimal constant. Then generalizations to other iterative procedures are discussed.

  20. 78 FR 62015 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-11

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on September 25, 2013, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe...

  1. 78 FR 26772 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-08

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on April 24, 2013, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe...

  2. 78 FR 2394 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-11

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on December 19, 2012, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe...

  3. 78 FR 20315 - Columbia Gas Transmission, LLC; Notice of Request under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-04

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request under Blanket Authorization Take notice that on Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe, Suite 2500...

  4. 77 FR 26544 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-04

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on April 18, 2012 Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe...

  5. 77 FR 36532 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-19

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on June 1, 2012, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe...

  6. 77 FR 42302 - Texas Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-18

    ... Energy Regulatory Commission Texas Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on July 2, 2012 Texas Gas Transmission, LLC (Texas Gas), 3800 Frederica Street... this application should be directed to Kathy D. Fort, Manager of Certificates and Tariffs, Texas Gas...

  7. European ceramic B.I.T. blanket for DEMO: Recent development for the zirconate version

    SciTech Connect

    Bielak, B.; Eid, M.; Fuetterer, M.

    1994-12-31

    Within the framework of the European test-blanket program, CEA and ENEA are jointly developing a DEMO-relevant, helium-cooled, Breeder-Inside-Tube (BIT) ceramic blanket. Two ceramics are possible breeder material candidate: LiAlO{sub 2} and Li{sub 2}ZrO{sub 3}. Despite the design has been originally developed for aluminate, the CEA has recently focused its work on zirconate. This concept blanket segments are made by a directly-cooled vacuum-tight steel box which contains banana-shaped poloidal breeder modules arranged in rows (6 rows in an outboard segment and 4 rows in an inboard one). A breeder module consists of a pressure vessel containing a bundle of breeder rods surrounded by baffles. Each one of the rods is made-up of a steel tube containing a stack of annular pellets of sintered lithium-zirconate, through which flows helium (the tritium purge gas). The thermo-mechanical analysis has shown that the thermal gradient in the ceramics can be kept at acceptable levels despite the poorer out-of-pile thermo-mechanical properties of zirconate compared to aluminate. Moreover, the neutronic analysis has shown that, besides the maintained tritium-breeding self-sufficiency capability of this blanket, the lower lithium burn-up could be an indication that the zirconate characteristics remains more stable after long term irradiation (i.e., close to the end-of-life fluence of 5 MWa/m{sup 2}).

  8. Chromospheric Line Blanketing and the Hydrogen Spectrum of dM Stars

    NASA Astrophysics Data System (ADS)

    Short, C. I.; Doyle, J. G.

    1996-12-01

    We present non-LTE calculations of the Hi spectrum in a grid of chromospheric models that represents a dM0 star in which the activity level ranges from quiescent to very active. We investigate three different treatments of the background opacity: 1) continuous opacity only, 2) blanketing due to lines that form in the photosphere below Tmin, and 3) blanketing by lines that form throughout the entire outer atmosphere. We show that the predicted W_lambda of Lyalpha in all models, and of Hα in very active (dMe) stars, is reduced by as much as a factor of ~ 4 by the inclusion of background line opacity. A consistent treatment of line blanketing that includes the effect of the chromospheric and transition region temperature structure in the calculation of background line opacity is necessary for the accurate calculation of Lyalpha , and in some cases Hα , in thes e stars. The Hα line in less active models, and the Pabeta line in all models, is negligibly affected by the treatment of background opacity. We also show that the broad-band continuum flux in regions where lambda < 2000 A is increased by as much as a factor of ~ 3 in some models by the inclusion of chromospheric line blanketing.

  9. 75 FR 8327 - Golden Pass Pipeline LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-24

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

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

    NASA Astrophysics Data System (ADS)

    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)

  11. 75 FR 8053 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-23

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization February 16, 2010. Take notice that on January 29, 2010, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 State Highway 56, Owensboro, Kentucky 42301, filed in Docket No. CP10-48-000, a prior...

  12. 78 FR 53746 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-30

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on August 13, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star..., Chautauqua and Montgomery Counties, Kansas. Southern Star's prior notice request is more fully set forth in...

  13. 78 FR 13663 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-28

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 11, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star... is on file with the Commission and open for public inspection. Specifically, Southern Star proposes...

  14. 78 FR 66915 - Notice of Request Under Blanket Authorization; Southern Star Central Gas Pipeline, Inc.

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-07

    ... Energy Regulatory Commission Notice of Request Under Blanket Authorization; Southern Star Central Gas Pipeline, Inc. Take notice that on October 21, 2013 Southern Star Central Gas Pipeline, Inc. (Southern Star... in Johnson and Pettis Counties, Missouri, under authorization issued to Southern Star in Docket No...

  15. 77 FR 14517 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-12

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 21, 2012 Southern Star Central Gas Pipeline, Inc. (Southern Star... City, Missouri. Specifically, Southern Star proposes to replace 3 miles of 12-inch diameter XT pipeline...

  16. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY OTHER REGULATIONS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES CERTAIN SALES AND TRANSPORTATION OF NATURAL GAS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES Blanket Certificates Authorizing Certain Natural Gas...

  17. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY OTHER REGULATIONS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES CERTAIN SALES AND TRANSPORTATION OF NATURAL GAS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES Blanket Certificates Authorizing Certain Natural Gas...

  18. 75 FR 13535 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-22

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization March 16, 2010. Take notice that on March 12, 2010, Northern Natural Gas Company (Northern), 1111 South... External Affairs, Northern Natural Gas Company, 1111 South 103rd Street, Omaha, Nebraska 68124, at (402...

  19. 75 FR 3232 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-01-20

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

  20. 78 FR 44111 - Florida Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-23

    ... Energy Regulatory Commission Florida Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization Take notice that on July 1, 2013, Florida Gas Transmission Company, LLC (Florida Gas), 1300 Main... Electronic Flow Measurement, all located in Palm Beach County, Florida. Florida Gas' prior notice request is...