Science.gov

Sample records for iter blanket maintenance

  1. ITER convertible blanket evaluation

    SciTech Connect

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

    1995-09-01

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

  2. 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.

  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. 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.

  6. U.S. Plans and Strategy for ITER Blanket Testing

    SciTech Connect

    Abdou, M.; Sze, D.; Wong, C.; Sawan, M.; Ying, A.; Morley, N.B.; Malang, S

    2005-04-15

    Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation in the ITER Test Blanket Module (TBM) Program. A US strategy for ITER-TBM has evolved that emphasizes international collaboration. A study was initiated to select the two blanket options for the US ITER-TBM in light of new R and D results from the US and world programs over the past decade. The study is led by the Plasma Chamber community in partnership with the Materials, PFC, Safety, and physics communities. The study focuses on assessment of the critical feasibility issues for candidate blanket concepts and it is strongly coupled to R and D of modeling and experiments. Examples of issues are MHD insulators, SiC insert viability and compatibility with PbLi, tritium permeation, MHD effects on heat transfer, solid breeder 'temperature window' and thermomechanics, and chemistry control of molten salts. A dual coolant liquid breeder and a helium-cooled solid breeder blanket concept have been selected for the US ITER-TBM.

  7. US technical report for the ITER blanket/shield

    NASA Astrophysics Data System (ADS)

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li2O) and lithium zirconate (Li2ZrO3) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  8. 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.

  9. 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.

  10. Tritium processing system for the ITER Li/V blanket test module

    SciTech Connect

    Sze, D.K.; Hua, T.Q.; Abdou, M.A.; Dagher, M.A.; Waganer, L.M.

    1997-04-01

    The purpose of the ITER Blanket Testing Module is to test the operating and performance of candidate blanket concepts under a real fusion environment. To assure fuel self-sufficiency the tritium breeding, recovery and processing have to be demonstrated. The tritium produced in the blanket has to be processed to a purity which can be used for refueling. All these functions need to be accomplished so that the tritium system can be scaled to a commercial fusion power plant from a safety and reliability point of view. This paper summarizes the tritium processing steps, the size of the equipment, power requirements, space requirements, etc. for a self-cooled lithium blanket. This information is needed for the design and layout of the test blanket ancillary system and to assure that the ITER guidelines for remote handling of ancillary equipment can be met.

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

    SciTech Connect

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

    2006-07-01

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

  12. 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.

    2015-11-01

    m/n=2/1 tearing modes can be slowed and stop rotating (lock) by eddy currents induced in resistive walls, a particular issue in ITER with large inertia and low applied torque. Previous estimates of tolerable 2/1 island widths in ITER, based on a forecast of initial island rotation, the n=1 resistive penetration time of the innervacuumvessel wall and benchmarked to DIII-D high-torque plasmas, found that the ITER ECCD system could catch and subdue such islands before they persisted long and grew large enough to lock. 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 will allow several times smaller islands to lock several times faster in ITER. Recent DIII-D ITER baseline scenario plasmas with low-applied torque allow better modeling and scaling to ITER with the blanket as the first resistive wall. This motivates using the ITER ECCD system in a CW preemptive operation of as little as a well-aligned 3 MW to avoid destabilizing the 2/1 NTM. Work supported by the US DOE under DE-FC02-04ER54698.

  13. Liquid metal blanket module testing and design for ITER/TIBER II

    SciTech Connect

    Mattas, R.F.; Cha, Y.; Finn, P.A.; Majumdar, S.; Picologlou, B.; Stevens, H.; Turner, L.

    1988-05-01

    A major goal for ITER is the testing of nuclear components to demonstrate the integrated performance of the most attractive concepts that can lead to a commercial fusion reactor. As part of the ITER/TIBER II study, the test program and design of test models were examined for a number of blanket concepts. The work at Argonne National Laboratory focused on self-cooled liquid metal blankets. A test program for liquid metal blankets was developed based upon the ITER/TIBER II operating schedule and the specific data needs to resolve the key issues for liquid metals. Testing can begin early in reactor operation with liquid metal MHD tests to confirm predictive capability. Combined heat transfer/MHD tests can be performed during initial plasma operation. After acceptable heat transfer performance is verified, tests to determine the integrated high temperature performance in a neutron environment can begin. During the high availability phase operation, long term performance and reliability tests will be performed. It is envisioned that a companion test program will be conducted outside ITER to determine behavior under severe accident conditions and upper performance limits. A detailed design of a liquid metal test module and auxiliary equipment was also developed. The module followed the design of the TPSS blanket. Detailed analysis of the heat transfer and tritium systems were performed, and the overall layout of the systems was determined. In general, the blanket module appears to be capable of addressing most of the testing needs. 8 refs., 27 figs., 11 tabs.

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

    NASA Astrophysics Data System (ADS)

    Shiba, K.; Enoeda, M.; Jitsukawa, S.

    2004-08-01

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

  15. ITER fast ion confinement in the presence of the European test blanket module

    NASA Astrophysics Data System (ADS)

    Äkäslompolo, Simppa; Kurki-Suonio, Taina; Asunta, Otto; Cavinato, Mario; Gagliardi, Mario; Hirvijoki, Eero; Saibene, Gabriella; Sipilä, Seppo; Snicker, Antti; Särkimäki, Konsta; Varje, Jari

    2015-09-01

    This paper addresses the confinement of thermonuclear alpha particles and neutral beam injected deuterons in the 15 MA Q = 10 inductive scenario in the presence of the magnetic perturbation caused by the helium cooled pebble bed test blanket module using the vacuum approximation. Both the flat top phase and plasma ramp-up are studied. The transport of fast ions is calculated using the Monte Carlo guiding center orbit-following code ASCOT. A detailed three-dimensional wall, derived from the ITER blanket module CAD data, is used for evaluating the fast ion wall loads. The effect of the test blanket module is studied for both overall confinement and possible hot spots. The study indicates that the test blanket modules do not significantly deteriorate the fast ion confinement.

  16. 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

  17. 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.

  18. 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

  19. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

    SciTech Connect

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

    1994-12-31

    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 or 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 austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250{degrees}C is more damaging than at 90{degrees}C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature.

  20. 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.

  1. A robust helium-cooled shield/blanket design for ITER

    NASA Astrophysics Data System (ADS)

    Wong, C. P. C.; Bourque, R. F.; Baxi, C. B.; Colleraine, A. P.; Grunloh, H. J.; Letchenberg, T.; Leuer, J. A.; Reis, E. E.; Redler, K.; Will, R.

    1993-11-01

    General Atomics Fusion and Reactor Groups have completed a helium-cooled, conceptual shield/blanket design for ITER. The configuration selected is a pressurized tubes design embedded in radially oriented plates. This plate can be made from ferritic steel or from V-alloy. Helium leakage to the plasma chamber is eliminated by conservative, redundant design and proper quality control and inspection programs. High helium pressure at 18 MPa is used to reduce pressure drop and enhance heat transfer. This high gas pressure is believed practical when confined in small diameter tubes. Ample industrial experience exists for safe high gas pressure operations. Inboard shield design is highlighted in this study since the allowable void fraction is more limited. Lithium is used as the thermal contacting medium and for tritium breeding; its safety concerns are minimized by a modular, low inventory design that requires no circulation of the liquid metal for the purpose of heat removal. This design is robust, conservative, reliable, and meets all design goals and requirements. It can also be built with present-day technology.

  2. Future technological tests on large-scale mock-ups of ITER blanket modules at IVV-2M reactor

    SciTech Connect

    Zyrianov, A.P.; Tokarev, V.I.; Zlokazov, S.B.

    1994-12-31

    A multisection core of water-cooled water-moderated reactor IVV-2M facilities testing of large scale mock-ups of ITER breeder blanket modules, the reactor arrangement in a building provides a maximum close position of tritium {open_quotes}in-pile{close_quotes} measurement station {open_quotes}RITM{close_quotes} to the core (in-pile testing of tritium producing mock-ups). Mock-ups of ceramic and liquid metal blankets are planned to be tested complying the following requirements: mock-up dimensions maximum close to those of ITER, distributions of nuclear power density, temperature fields, tritium release modes at continuous helium purging, provision of cyclic neutron and thermal loading variations. Variants of location of large ({approximately}150x200 mm) mock-up of ceramic blanket and a submerged loop facility containing liquid lithium and a vanadium alloy as a structure material are described. A technological scheme of {open_quotes}RITM{close_quotes} measurement station to study tritium system operation modes are presented.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  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. 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.

  11. ITER tokamak device

    NASA Astrophysics Data System (ADS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-07-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER, a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fueling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  12. 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

  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. Fusion Blanket Development in FDF

    NASA Astrophysics Data System (ADS)

    Wong, C. P. C.; Smith, J. P.; Stambaugh, R. D.

    2008-11-01

    To satisfy the electricity and tritium self-sufficiency missions of a Fusion Development Facility (FDF), suitable blanket designs will need to be evaluated, selected and developed. To demonstrate closure of the fusion fuel cycle, 2-3 main tritium breeding blankets will be used to cover most of the available chamber surface area in order to reach the project goal of achieving a tritium breeding ratio, TBR > 1. To demonstrate the feasibility of electricity and tritium production for subsequent devices such as the fusion demonstration power reactor (DEMO), several advanced test blankets will need to be selected and tested on the FDF to demonstrate high coolant outlet temperature necessary for efficient electricity production. Since the design goals for the main and test blankets are different, the design criteria of these blankets will also be different. The considerations in performing the evaluation of blanket and structural material options in concert with the maintenance approach for the FDF will be reported in this paper.

  16. 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.

  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. 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.

  19. 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.

  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. 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.

  3. Radiation analysis of the ITER pellet injection system

    SciTech Connect

    Gouge, M.J. ); Gomes, I.C.; Gomes, L.T.; Stevens, P.N, )

    1991-03-01

    The results of neutronics calculations for the pellet injection system of the International Thermonuclear Experimental Reactor (ITER) are described. Hands-on maintenance of components in the pellet injection room results in a considerable simplification of maintenance support equipment and in greater system availability. The basic configuration of the pellet injection system includes small-diameter guide tubes with which the pellet may have several small-angle collisions before reaching the plasma. The pellet injector port through which the guide tubes pass will be shared with ITER plasma diagnostics, so the calculation takes into account penetrations to accommodate numerous channels for a neutron spectrometer and neutron and gamma-ray cameras. The conservative assumption of steady-state operation of ITER for 1000 days was taken as the baseline for calculating the activation of components in the pellet injection room. The plasma configuration is based on the current ITER guidelines, the first wall configuration is based on the most recently updated configuration, and the blanket configuration is based on the US proposal for the blanket. The plasma, coils, and blanket regions were analyzed with the Monte Carlo code MCNP. The transport of neutrons through the penetrations was also performed with MCNP. The pellet injection room was modeled with the two-dimensional discrete ordinates code DORT, which was also used for the transport of neutrons during operation and of gamma rays caused by activation. The activation calculations were carried out with the REBATE code. Results from this study indicate that restricted personnel access to the pellet injection room is possible, so limited hands-on maintenance can be performed on the majority of the components in the room.

  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. Materials for breeding blankets

    SciTech Connect

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

    1995-09-01

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

  7. ITER EDA project status

    NASA Astrophysics Data System (ADS)

    Chuyanov, V. A.

    1996-10-01

    The status of the ITER design is as presented in the Interim Design Report accepted by the ITER council for considerations by ITER parties. Physical and technical parameters of the machine, conditions of operation of main nuclear systems, corresponding design and material choices are described, with conventional materials selected. To fully utilize the safety and economical potential of fusion advanced materials are necessary. ITER shall and can be built with materials already available. The ITER project and advanced fusion material developments can proceed in parallel. The role of ITER is to establish (experimentally) requirements to these materials and to provide a test bed for their final qualification in fusion reactor environment. To achieve this goal, the first wall/blanket modules test program is foreseen.

  8. 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.

  9. 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. PMID:22112596

  10. 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.

  11. Blanket technology workshop report

    NASA Technical Reports Server (NTRS)

    Scott-Monck, J. A.

    1980-01-01

    The solar array blanket, defined as a substrate covered with interconnected and glassed solar cells, but excluding the necessary support structure, deployment, and orientation devices is considered. The interactions between the blanket and the structure that is used to package, deploy, support and, if necessary restow it, are addressed along with systems constraints such as spacecraft configuration, size, and payload requirements. The influence on blanket design is emphasized. The three main mission classes considered are low Earth orbital (LEO), intermediate, or LEO to GEO transfer, and geosynchronous (GEO). Although interplanetary missions could be considered to be a separate class, their requirements, primarily power per unit mass, are generally close enough to geosynchronous missions to allow this mission class to be included within the third type. Examination of the critical elements of each class coupled with considerations of the shuttle capabilities is used to define the type of blanket technology most likely required to support missions that will be flown starting in 1990.

  12. 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.

  13. 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.

  14. Novel method for sludge blanket measurements.

    PubMed

    Schewerda, J; Förster, G; Heinrichmeier, J

    2014-01-01

    The most widely used methods for sludge blanket measurements are based on acoustic or optic principles. In operation, both methods are expensive and often maintenance-intensive. Therefore a novel, reliable and simple method for sludge blanket measurement is proposed. It is based on the differential pressure measurement in the sludge zone compared with the differential pressure in the clear water zone, so that it is possible to measure the upper and the lower sludge level in a tank. Full-scale tests of this method were done in the secondary clarifier at the waste water treatment plant in Hecklingen, Germany. The result shows a good approximation of the manually measured sludge level. PMID:24569276

  15. Blanket integrated blocking diodes

    NASA Astrophysics Data System (ADS)

    Uebele, P.; Kasper, C.; Rasch, K.-D.

    1986-11-01

    Two types of large area protection diodes for integration in solar arrays were developed in planar technology. For application in a bus voltage concept of V sub bus = 80 V a p-doped blanket integrated blocking diode (p-IBD) was developed with V sub rev = 120 V, whereas for the high voltage concept of V sub bus = 160 V a n-IBD with V sub rev = 250 V was developed. Application as blanket integrated shunt diodes is recommended. The optimized rearside diffusion provides a low forward voltage drop in the temperature range of minus 100 to plus 150 C. As a consequence of planar technology metallized coverglasses have to be used to minimize the photocurrent.

  16. 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.

  17. Evaluation of the Environmental Gamma-ray Dose Rate by Skyshine Analysis During the Maintenance of an Activated TFC in ITER

    NASA Astrophysics Data System (ADS)

    Sato, S.; Takatsu, H.; Maki, K.; Yamada, K.; Mori, S.; Iida, H.; Santoro, R. T.

    1997-09-01

    Gamma-ray exposure dose rates at the ITER site boundary were estimated for the cases of removal of a failed activated Toroidal Field (TF) coil from the torus and removal of a failed activated TF coil together with a sector of the activated Vacuum Vessel (VV). Skyshine analyses were performed using the two-dimensional SN radiation transport code, DOT3.5. The exposure gamma-ray dose rates on the ground at the site boundary (presently assumed to be 1 km from the ITER building), were calculated to be 1.1 and 84 μSv/year for removal of the TF coil without and with a VV sector, respectively. The dose rate level for the latter case is close to the tentative radiation limit of 100 μSv/year so an additional ˜14 cm of concrete is required in the ITER building roof to satisfy the criterion for a safety factor often for the site boundary dose rate.

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. ITER's woes

    NASA Astrophysics Data System (ADS)

    jjeherrera; Duffield, John; ZoloftNotWorking; esromac; protogonus; mleconte; cmfluteguy; adivita

    2014-07-01

    In reply to the physicsworld.com news story “US sanctions on Russia hit ITER council” (20 May, http://ow.ly/xF7oc and also June p8), about how a meeting of the fusion experiment's council had to be moved from St Petersburg and the US Congress's call for ITER boss Osamu Motojima to step down.

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

    NASA Astrophysics Data System (ADS)

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

    2014-04-01

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

  4. 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.

  5. ITER EDA Newsletter. Volume 3, no. 2

    NASA Astrophysics Data System (ADS)

    1994-02-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the Fifth ITER Council Meeting held in Garching, Germany, January 27-28, 1994, a visit (January 28, 1994) of an international group of Harvard Fellows to the San Diego Joint Work Site, the Inauguration Ceremony of the EC-hosted ITER joint work site in Garching (January 28, 1994), on an ITER Technical Meeting on Assembly and Maintenance held in Garching, Germany, January 19-26, 1994, and a report on a Technical Committee Meeting on radiation effects on in-vessel components held in Garching, Germany, November 15-19, 1993, as well as an ITER Status Report.

  6. Design of the Remote Steerable ECRH launching system for the ITER upper ports

    NASA Astrophysics Data System (ADS)

    Verhoeven, A. G. A.; Elzendoorn, B. S. Q.; Bongers, W. A.; Bruschi, A.; Cirant, S.; Danilov, I.; Fernandez, A.; Gantenbein, G.; Graswinckel, M. F.; Heidinger, R.; Kasparek, W.; Kleefeldt, K.; Kruijt, O. G.; Lamers, B.; Piosczyk, B.; Plaum, B.; Ronden, D. M. S.; Saibene, G.; Zohm, H.

    2005-01-01

    An ECRH (electron-cyclotron resonance heating) launching system for the ITER upper ports is being designed. The aim of the system is to inject Electron Cyclotron Waves (ECW) in the ITER plasma in order to stabilize neoclassical tearing modes (NTM). Each of the four upper-port launchers consists of six mm-wave lines capable of transmitting high power up to 2 MW per line at 170 GHz. In order to exploit the capability of ECW for localized heating and current drive over a range of plasma radii in ITER, the ECH&CD upper port launcher must have a beam steering capability. The Remote Steering (RS) principle has great advantages, because it enables to avoid steerable mirrors with flexible cooling lines at the plasma-facing end of the launcher. The principle consists of a long, corrugated, square waveguide having the steerable optics placed outside of the first confinement boundary of the vacuum vessel. All vulnerable components are far away from the hostile plasma environment. Furthermore, the RS launching system enables to do maintenance on the system during shutdown, without affecting the torus vacuum and the blanket cooling circuits.

  7. 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.

  8. 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.

  9. Line Blanketing in Przybylski's Star

    NASA Astrophysics Data System (ADS)

    Cowley, C. R.; Kupka, F.; Mathys, G.

    1999-12-01

    Przybylski's star (HD 101065) may be the most heavily blanketed star known. It therefore provides a test of our techniques for line blanketing. The current abstract draws on a paper in preparation by CRC, T. Ryabchikova, F. Kupka, G. Mathys, and D. J. Bord, based on ESO spectra obtained by GM. Unfortunately, the atomic species that provide the majority of the line blanketing in Przybylski's star does not have enough atomic data for realistic calculations of the blanketing. We therefore discuss three models in which iron-group elements were articifically elevated in abundance in the calculation of opacity used to construct the models. We thank Drs. R. L. Kurucz, and Bengt Edvardsson for calculating respectively Models 1 (dashed [Fe/H]=+3) and 2 (dot-dash, [Fe/H]=+2) at our request. Model 3 (line, [Fe/H]) was calculated by FK, using the Canuto-Mazzitelli formalism. Figure 1 (www.astro.lsa.umich.edu/usrs/cowley/models.gif), shows these 3 models in good agreement with one another, and clearly different from a standard solar-abundance Atlas9 model (dashed) with the same effective temperature. All three models are scaled to Te=6600K. The blanketed models have little or no convection, and show the lowered boundary temperature of classical picket-fence models. The true boundary temperature may be still lower than in these numerical models. Abundances from Pr I and Nd I are systematically higher than those from the corresponding second spectra, as are those from Pr III and Nd III. It was noted long ago by Przybylski and others that the Balmer profiles had cores indicative of temperatures of some 6000K; the wings could be fit with much higher temperatures--perhaps as high as 7500K. Molecular species have been sought but not identified. Calculations show CN and CH lines would be very weak, even if the temperature between log(tau5000)=-3.5 and -5.4 were allowed to drop to 3000K.

  10. Technical evaluation of major candidate blanket systems for fusion power reactor

    NASA Astrophysics Data System (ADS)

    Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

    1987-03-01

    The key functions required for tritium breeding blankets for a fusion power reactor are ; (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li/sub 2/O/H/sub 2/O/Be, Mo-alloy/Li/sub 2/O/He/Be, Mo-alloy/LiAlO/sub 2//He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies.

  11. ITER Diagnostic First Wal

    SciTech Connect

    G. Douglas Loesser, et. al.

    2012-09-21

    The ITER Diagnostic Division is responsible for designing and procuring the First Wall Blankets that are mounted on the vacuum vessel port plugs at both the upper and equatorial levels This paper will discuss the effects of the diagnostic aperture shape and configuration on the coolant circuit design. The DFW design is driven in large part by the need to conform the coolant arrangement to a wide variety of diagnostic apertures combined with the more severe heating conditions at the surface facing the plasma, the first wall. At the first wall, a radiant heat flux of 35W/cm2 combines with approximate peak volumetric heating rates of 8W/cm3 (equatorial ports) and 5W/cm3 (upper ports). Here at the FW, a fast thermal response is desirable and leads to a thin element between the heat flux and coolant. This requirement is opposed by the wish for a thicker FW element to accommodate surface erosion and other off-normal plasma events.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. ARIES-IV Nested Shell Blanket Design

    SciTech Connect

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

    1993-11-01

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

  17. 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.

  18. Multivariable optimization of fusion reactor blankets

    SciTech Connect

    Meier, W.R.

    1984-04-01

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% /sup 6/Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO/sub 2/ breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO/sub 2/ breeding blanket enriched to 34% /sup 6/Li.

  19. Ferritic-Martensitic steel Test Blanket Modules: Status and future needs for design criteria requirements and fabrication validation

    NASA Astrophysics Data System (ADS)

    Salavy, J.-F.; Aiello, G.; Aubert, P.; Boccaccini, L. V.; Daichendt, M.; De Dinechin, G.; Diegele, E.; Giancarli, L. M.; Lässer, R.; Neuberger, H.; Poitevin, Y.; Stephan, Y.; Rampal, G.; Rigal, E.

    2009-04-01

    The Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble Bed are the two breeding blankets concepts for the DEMO reactor which have been selected by EU to be tested in ITER in the framework of the Test Blanket Module projects. They both use a 9%CrWVTa Reduced Activation Ferritic-Martensitic steel, called EUROFER, as structural material and helium as coolant. This paper gives an overview of the status of the EUROFER qualification program and discusses the future needs for design criteria requirements and fabrication validation.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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

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

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

    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.

  6. 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.

  7. 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)

  8. Current Trends of Blanket Research and Deveopment in Japan 3.Blanket Designs in Fusion Power Reactors

    NASA Astrophysics Data System (ADS)

    Sagara, Akio; Enoeda, Mikio; Nishio, Satoshi; Kozaki, Yasuji

    The main functions of the blanket in fusion power reactors are basically independent of the type of magnetic fusion reactor (tokamak, helical, etc.) and inertia fusion. However, from technical point of view, many candidate designs of blanket have been proposed depending on the particular reactor concepts. Their main features are characterized for the recent typical designs, and key issues are defined.

  9. Development of welding technologies for the manufacturing of European Tritium Breeder blanket modules

    NASA Astrophysics Data System (ADS)

    Poitevin, Y.; Aubert, Ph.; Diegele, E.; de Dinechin, G.; Rey, J.; Rieth, M.; Rigal, E.; von der Weth, A.; Boutard, J.-L.; Tavassoli, F.

    2011-10-01

    Europe has developed two reference Tritium Breeder Blankets concepts for a DEMO fusion reactor: the Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble-Bed. Both are using the reduced-activation ferritic-martensitic EUROFER-97 steel as structural material and will be tested in ITER under the form of test blanket modules. The fabrication of their EUROFER structures requires developing welding processes like laser, TIG, EB and diffusion welding often beyond the state-of-the-art. The status of European achievements in this area is reviewed, illustrating the variety of processes and key issues behind retained options, in particular with respect to metallurgical aspects and mechanical properties. Fabrication of mock-ups is highlighted and their characterization and performances with respect to design requirements are reviewed.

  10. Development and qualification of functional materials for the EU Test Blanket Modules: Strategy and R&D activities

    NASA Astrophysics Data System (ADS)

    Zmitko, M.; Poitevin, Y.; Boccaccini, L.; Salavy, J.-F.; Knitter, R.; Möslang, A.; Magielsen, A. J.; Hegeman, J. B. J.; Lässer, R.

    2011-10-01

    Europe has developed two reference tritium breeder blankets concepts for a DEMO fusion reactor: the Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble-Bed. Both will be tested in ITER under the form of Test Blanket Modules (TBMs). The paper reviews the current status of development and qualification of the EU TBMs functional materials; i.e. ceramic solid breeder materials, beryllium/beryllides multiplier materials and Lithium-Lead liquid metal breeder material Pb-15.7Li. For each functional material the main functional/performance requirements with key qualification issues, current status of the R&D activities and the EU development strategy are presented. In the development strategy major steps considered are listed pointing out importance of the 'Development/qualification/procurement plan', currently under elaboration, for definition of a roadmap of further activities aiming at delivery of qualified functional materials to be used in the European TBMs in ITER.

  11. 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.

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

    NASA Astrophysics Data System (ADS)

    1993-08-01

    The Committee's evaluation of vanadium alloys as a structural material for fusion reactors was constrained by limited data and time. The design of the International Thermonuclear Experimental Reactor is still in the concept stage, so meaningful design requirements were not available. The data on the effect of environment and irradiation on vanadium alloys were sparse, and interpolation of these data were made to select the V-5Cr-5Ti alloy. With an aggressive, fully funded program it is possible to qualify a vanadium alloy as the principal structural material for the ITER blanket in the available 5 to 8-year window. However, the data base for V-5Cr-5Ti is limited 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 working 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, it is obvious that there is a need to move toward integrating fabrication, welding, and materials engineers into the ITER design team. If the vanadium alloy 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.

  13. 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.

  14. 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.

  15. 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.

  16. 32 CFR 318.14 - Blanket routine uses.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 2 2010-07-01 2010-07-01 false Blanket routine uses. 318.14 Section 318.14 National Defense Department of Defense (Continued) OFFICE OF THE SECRETARY OF DEFENSE (CONTINUED) PRIVACY PROGRAM DEFENSE THREAT REDUCTION AGENCY PRIVACY PROGRAM § 318.14 Blanket routine uses. (a) Blanket...

  17. Advanced Polymer For Multilayer Insulating Blankets

    NASA Technical Reports Server (NTRS)

    Haghighat, R. Ross; Shepp, Allan

    1996-01-01

    Polymer resisting degradation by monatomic oxygen undergoing commercial development under trade name "Aorimide" ("atomic-oxygen-resistant imidazole"). Intended for use in thermal blankets for spacecraft in low orbit, useful on Earth in outdoor applications in which sunlight and ozone degrades other plastics. Also used, for example, to make threads and to make films coated with metals for reflectivity.

  18. Fidget Blankets: A Sensory Stimulation Outreach Program.

    PubMed

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

    2016-01-01

    Behavioral and Psychological Symptoms of Dementia (BPSD) include behaviors such as aberrant motor behavior, agitation, anxiety, apathy, delusions, depression, disinhibition, elation, hallucinations, irritability, and sleep or appetite changes. A student-led project to provide sensory stimulation in the form of "fidget blankets" developed into a community outreach program. The goal was to decrease the use of antipsychotics used for BPSD. PMID:27250073

  19. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    .... Where a manufacturer of tobacco products operates more than one factory he may, in lieu of filing... provisions of § 40.134, for any or all of the factories. The total amount of any blanket bond given under this section shall be available for the satisfaction of any liability incurred at any factory...

  20. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    .... Where a manufacturer of tobacco products operates more than one factory he may, in lieu of filing... provisions of § 40.134, for any or all of the factories. The total amount of any blanket bond given under this section shall be available for the satisfaction of any liability incurred at any factory...

  1. 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)...

  2. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    .... Where a manufacturer of tobacco products operates more than one factory he may, in lieu of filing... provisions of § 40.134, for any or all of the factories. The total amount of any blanket bond given under this section shall be available for the satisfaction of any liability incurred at any factory...

  3. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    .... Where a manufacturer of tobacco products operates more than one factory he may, in lieu of filing... provisions of § 40.134, for any or all of the factories. The total amount of any blanket bond given under this section shall be available for the satisfaction of any liability incurred at any factory...

  4. 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

  5. Thermal insulation blanket material. Final Report

    SciTech Connect

    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.

  6. 18 CFR 157.203 - Blanket certification.

    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 certification. 157.203 Section 157.203 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY REGULATIONS UNDER NATURAL GAS ACT APPLICATIONS FOR CERTIFICATES OF PUBLIC CONVENIENCE AND NECESSITY AND FOR ORDERS...

  7. US ITER Moving Forward

    ScienceCinema

    US ITER / ORNL

    2012-03-16

    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.

  8. 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.

  9. 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).

  10. Performance assessment of the ITER ICRF antenna

    NASA Astrophysics Data System (ADS)

    Durodié, F.; Vrancken, M.; Bamber, R.; Colas, L.; Dumortier, P.; Hancock, D.; Huygen, S.; Lockley, D.; Louche, F.; Maggiora, R.; Milanesio, D.; Messiaen, A.; Nightingale, M. P. S.; Shannon, M.; Tigwell, P.; van Schoor, M.; Wilson, D.; Winkler, K.; Cycle Team

    2014-02-01

    ITER's Ion Cyclotron Range of Frequencies (ICRF) system [1] comprises two antenna launchers designed by CYCLE (a consortium of European associations listed in the author affiliations above) on behalf F4E for the ITER Organisation (IO), each inserted as a Port Plug (PP) into one of ITER's Vacuum Vessel (VV) ports. Each launcher is an array of 4 toroidal by 6 poloidal RF current straps specified to couple up to 20 MW in total to the plasma in the frequency range of 40 to 55 MHz but limited to a maximum system voltage of 45 kV and limits on RF electric fields depending on their location and direction with respect to respectively the torus vacuum and the toroidal magnetic field. A crucial aspect of coupling ICRF power to plasmas is the knowledge of the plasma density profiles in the Scrape-Off Layer (SOL) and the location of the RF current straps with respect to the SOL. The launcher layout and details were optimized and its performance estimated for a worst case SOL provided by the IO. The paper summarizes the estimated performance obtained within the operational parameter space specified by IO. Aspects of the RF grounding of the whole antenna PP to the VV port and the effect of the voids between the PP and the Blanket Shielding Modules (BSM) surrounding the antenna front are discussed.

  11. Chicxulub Ejecta Blanket Deposits From Belize

    NASA Technical Reports Server (NTRS)

    Ocampo, A.

    1995-01-01

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

  12. A light blanket for intraoperative photodynamic therapy

    NASA Astrophysics Data System (ADS)

    Hu, Yida; Wang, Ken; Zhu, Timothy C.

    2009-06-01

    A novel light source - light blanket composed of a series of parallel cylindrical diffusing fibers (CDF) is designed to substitute the hand-held point source in the PDT treatment of the malignant pleural or intraperitoneal diseases. It achieves more uniform light delivery and less operation time in operating room. The preliminary experiment was performed for a 9cmx9cm light blanket composed of 8 9-cm CDFs. The linear diffusers were placed in parallel fingerlike pockets. The blanket is filled with 0.2 % intralipid scattering medium to improve the uniformity of light distribution. 0.3-mm aluminum foil is used to shield and reflect the light transmission. The full width of the profile of light distribution at half maximum along the perpendicular direction is 7.9cm and 8.1cm with no intralipid and with intralipid. The peak value of the light fluence rate profiles per input power is 11.7mW/cm2/W and 8.6mW/cm2/W respectively. The distribution of light field is scanned using the isotropic detector and the motorized platform. The average fluence rate per input power is 8.6 mW/cm2/W and the standard deviation is 1.6 mW/cm2/W for the scan in air, 7.4 mW/cm2/W and 1.1 mW/cm2/W for the scan with the intralipid layer. The average fluence rate per input power and the standard deviation are 20.0 mW/cm2/W and 2.6 mW/cm2/W respectively in the tissue mimic phantom test. The light blanket design produces a reasonably uniform field for effective light coverage and is flexible to confirm to anatomic structures in intraoperative PDT. It also has great potential value for superficial PDT treatment in clinical application.

  13. 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

  14. 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…

  15. 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…

  16. 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. ...

  17. 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…

  18. 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.

  19. Neutronics analysis of deuterium-tritium-driven experimental hybrid blankets

    SciTech Connect

    Sahin, S.; Kumar, A.

    1984-07-01

    At the Swiss Federal Institute of Technology, an experimental fusion and fusion-fission (hybrid) reactor facility is near completion. Experiments are scheduled to begin in February 1984. The experimental cavity leads one to plan experiments mostly with blankets in plane geometry. Five different hybrid blanket modules in plane geometry are analyzed with two different left boundary conditions representing varying experimental situations. Numbers I and II represent energy and fissile fuel producing blankets, whereas number III is mainly a fissile fuel producing blanket. Numbers IV and V are actinide burning blankets. It is shown that the overall neutronic performance, such as k /sub eff/ , energy multiplication factor M, fusile and fissile breeding, of a hybrid blanket with transplutonium actinide fuel is already better than that of a UO/sub 2/ or ThO/sub 2/ hybrid blanket. Furthermore, the transplutonium actinide waste is partly converted into precious nuclear fuel of a new type, such as /sup 242m/ Am and /sup 245/Cm. An experimental blanket with a vacuum left boundary has a harder neutron spectrum, and also excessive neutron leakage from the front surface and the lateral surfaces, as compared to that in the blanket in confinement geometry. It leads to the poorer neutronic performance of the former.

  20. 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.

  1. 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.

  2. Conceptual study of fusion-driven transmutation reactor with ITER physics and engineering constraints

    NASA Astrophysics Data System (ADS)

    Hong, Bong

    2011-10-01

    A conceptual study of fusion-driven transmutation reactor was performed based on ITER physics and engineering constraints. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

  3. Preconditioned Iterative Solver

    Energy Science and Technology Software Center (ESTSC)

    2002-08-01

    AztecOO contains a collection of preconditioned iterative methods for the solution of sparse linear systems of equations. In addition to providing many of the common algebraic preconditioners and basic iterative methods, AztecOO can be easily extended to interact with user-provided preconditioners and matrix operators.

  4. 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…

  5. Development of advanced blanket materials for a solid breeder blanket of a fusion reactor

    NASA Astrophysics Data System (ADS)

    Kawamura, H.; Ishitsuka, E.; Tsuchiya, K.; Nakamichi, M.; Uchida, M.; Yamada, H.; Nakamura, K.; Ito, H.; Nakazawa, T.; Takahashi, H.; Tanaka, S.; Yoshida, N.; Kato, S.; Ito, Y.

    2003-08-01

    The design of an advanced solid breeding blanket in a DEMO reactor requires a tritium breeder and a neutron multiplier that can withstand high temperatures and high neutron fluences, and the development of such advanced blanket materials has been carried out by collaboration between JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by a wet process is a reference material as a tritium breeder, but its stability at high temperatures has to be improved for its application in a DEMO blanket. One of these improved materials, TiO2-doped Li2TiO3 pebbles, was successfully fabricated and studied. For the advanced neutron multiplier, beryllides that have a high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that it had lower swelling and tritium inventory than beryllium metal. Pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. These activities have shown that there is a bright prospect in realizing a DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides.

  6. 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.

  7. 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.

  8. Specific welds for test blanket modules

    NASA Astrophysics Data System (ADS)

    Rieth, Michael; Rey, Jörg

    2009-04-01

    Fabrication and assembling test blanket modules needs a variety of different welding techniques. Therefore, an evaluation of plate joining for breeder units by tungsten-inert-gas, laser, and electron beam welding was performed by qualification of relevant mechanical properties like hardness, charpy, and creep strength. The focus was laid on the study of post-weld heat treatments at lowest possible temperatures and for maximum recovery of the joints. The most important result is that thin EUROFER plates may be welded by EB or laser techniques without the necessity of post-welding heat treatments that include an austenitization step.

  9. 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

  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. 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.

  12. 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...

  13. 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...

  14. 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...

  15. 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...

  16. 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...

  17. Security Blankets and Children's Security of Attachment to Their Mothers.

    ERIC Educational Resources Information Center

    Donate-Bartfield, Evelyn L.; Passman, Richard H.

    This study investigated the relations between toddlers' degree of attachment to their mothers and their development of an attachment to a security blanket. Seventy-four 18-month-olds were separated from their mothers three times; the third time the toddlers were left for 5 minutes in an unfamiliar playroom with their blanket and with a stranger.…

  18. 75 FR 51482 - Woven Electric Blankets From China

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-20

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

  19. Overview of the TFTB lithium blanket module program

    SciTech Connect

    Jassby, D.L.

    1986-01-01

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

  20. Diffusive heat blanketing envelopes of neutron stars

    NASA Astrophysics Data System (ADS)

    Beznogov, M. V.; Potekhin, A. Y.; Yakovlev, D. G.

    2016-06-01

    We construct new models of outer heat blanketing envelopes of neutron stars composed of binary ion mixtures (H-He, He-C, C-Fe) in and out of diffusive equilibrium. To this aim, we generalize our previous work on diffusion of ions in isothermal gaseous or Coulomb liquid plasmas to handle non-isothermal systems. We calculate the relations between the effective surface temperature Ts and the temperature Tb at the bottom of heat blanketing envelopes (at a density ρb ˜ 108 - 1010 g cm-3) for diffusively equilibrated and non-equilibrated distributions of ion species at different masses ΔM of lighter ions in the envelope. Our principal result is that the Ts-Tb relations are fairly insensitive to detailed distribution of ion fractions over the envelope (diffusively equilibrated or not) and depend almost solely on ΔM. The obtained relations are approximated by analytic expressions which are convenient for modelling the evolution of neutron stars.

  1. 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.

  2. The TFTR lithium blanket module program

    SciTech Connect

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

    1985-02-01

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

  3. 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.

  4. 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.

  5. Neutron dosimetry for the Lithium-Blanket-Module program

    SciTech Connect

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

    1982-01-01

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

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

    SciTech Connect

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

    1981-01-01

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

  7. Neutronic design for the TFTR lithium blanket module

    SciTech Connect

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

    1981-01-01

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

  8. 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.

  9. 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.

  10. ITER nominates next leader

    NASA Astrophysics Data System (ADS)

    Clery, Daniel

    2015-01-01

    Bernard Bigot, chair of France’s Alternative Energies and Atomic Energy Commission (CEA), has been chosen as the next director-general of ITER - the experimental fusion reactor currently being built in Cadarache, France.

  11. Status and perspective of the R&D on ceramic breeder materials for testing in ITER

    NASA Astrophysics Data System (ADS)

    Ying, A.; Akiba, M.; Boccaccini, L. V.; Casadio, S.; Dell'Orco, G.; Enoeda, M.; Hayashi, K.; Hegeman, J. B.; Knitter, R.; van der Laan, J.; Lulewicz, J. D.; Wen, Z. Y.

    2007-08-01

    The main line of ceramic breeder materials research and development is based on the use of the breeder material in the form of pebble beds. At present, there are three candidate pebble materials (Li 4SiO 4, and two forms of Li 2TiO 3) for DEMO reactors that will be used for testing in ITER. This paper reviews the R&D of as-fabricated pebble materials against the blanket performance requirements and makes recommendations on necessary steps toward the qualification of these materials for testing in ITER.

  12. Privatizing Maintenance.

    ERIC Educational Resources Information Center

    Hounsell, Dan

    1996-01-01

    Schools and other government facilities want to see whether privatization of maintenance can provide services as efficiently and at less cost than inhouse workers. Privatization proponents say that everyone will benefit the most if the bidding process involves competition. Offers examples from the Memphis City Schools and the Union Public Schools…

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

    NASA Astrophysics Data System (ADS)

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

    2011-02-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2011-10-01

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

  15. 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.

  16. Streaming analysis of gap between blanket modules for fusion experimental reactor

    SciTech Connect

    Sato, Satoshi; Takatsu, Hideyuki; Seki, Yasushi; Utsumi, Toshihisa

    1996-12-31

    Shielding analyses of the inboard blanket, the vacuum vessel and the Toroidal Field Coil (TFC) in International Thermonuclear Experimental Reactor (ITER) were performed by Monte Carlo and 2-dimensional discrete ordinate methods taking the radiation streaming through the 20 mm wide gap between the adjacent blanket modules into account, and their peak nuclear responses were evaluated. The nuclear responses of the TFC could fully satisfy the radiation limits. On the other hand, the helium production rates of the branch pipe, the leg and the front surface of the vacuum vessel behind the gap were about 2 - 3 times higher than the radiation limit at the end of the operation, i.e. the neutron fluence of 3 MWa/m{sup 2}. So the shielding module is required to be increased by 80 mm to satisfy the radiation limit. Also, shielding analyses for 20 - 100 mm wide gaps were performed, and it was found that the gap width could be increased by up to 50 mm from the TFCs protection for their peak nuclear responses point of view. 6 refs., 8 figs., 3 tabs.

  17. 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).

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

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Jayakumar, T.

    2011-10-01

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

  19. Plasma material interaction studies on low activation materials used for plasma facing or blanket component

    NASA Astrophysics Data System (ADS)

    Hino, T.; Hirohata, Y.; Yamauchi, Y.; Hashiba, M.; Kohyama, A.; Katoh, Y.; Lee, Y.; Jinushi, T.; Akiba, M.; Nakamura, K.; Yoshida, H.; Sengoku, S.; Tsuzuki, K.; Kusama, Y.; Yamaguchi, K.; Muroga, T.

    2004-08-01

    Numerous issues on the plasma material interactions were investigated for low activation materials. Co-deposited carbon dust was prepared and the deuterium concentration was measured. The concentration was approximately half of the present design value for ITER. For ferritic steel, the deuterium retention was observed to be comparable to that of stainless steel. Physical sputtering yield was roughly the same as that for stainless steel. For the reduction of absorption rate in vanadium alloy, titanium oxide coating was conducted, and the coating was observed to be very effective for reduction of hydrogen absorption. Helium gas permeability was measured for numerous SiC/SiC composites, and the SiC/SiC composite made by the NITE process showed quite low permeability. The SiC/SiC blanket may be able to be used without helium leakage into plasma.

  20. 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.

  1. 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.

  2. Design analyses of self-cooled liquid metal blankets

    SciTech Connect

    Gohar, Y.

    1986-12-01

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

  3. 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. PMID:19737263

  4. Disinfection of woollen blankets in steam at subatmospheric pressure

    PubMed Central

    Alder, V. G.; Gillespie, W. A.

    1961-01-01

    Blankets may be disinfected in steam at subatmospheric pressures by temperatures below boiling point inside a suitably adapted autoclave chamber. The chamber and its contents are thoroughly evacuated of air so as to allow rapid heat penetration, and steam is admitted to a pressure of 10 in. Hg below atmospheric pressure, which corresponds to a temperature of 89°C. Woollen blankets treated 50 times by this process were undamaged. Vegetative organisms were destroyed but not spores. The method is suitable for large-scale disinfection of blankets and for disinfecting various other articles which would be damaged at higher temperatures. PMID:13860203

  5. The excitation of plasma lines in blanketing sporadic E

    NASA Technical Reports Server (NTRS)

    Gordon, W. E.; Carlson, H. C.

    1976-01-01

    Enhanced plasma lines in blanketing sporadic E have been excited by a powerful HF radio wave illuminating the E region over the Arecibo Observatory. The plasma lines are observed by the incoherent scatter radar at the observatory. They originate in the sporadic E layer when the blanketing frequency exceeds the exciting frequency, a result which confirms that the plasma is overdense for the exciting frequency. Around the time when the blanketing frequency falls through the exciting frequency, large fluctuations in the plasma line intensities are observed, and thus the possibility of overdense patches drifting through the sampled volume is suggested.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Iterated multidimensional wave conversion

    NASA Astrophysics Data System (ADS)

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

    2011-12-01

    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.

  11. 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.

  12. 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.

  13. 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.

  14. Flexible, Thin-Film Solar-Cell Blanket

    NASA Technical Reports Server (NTRS)

    Stella, Paul M.

    1992-01-01

    Much of available area used to absorb solar energy. Proposed blanket of solar photovoltaic cells mounted on exterior surface of equipment it powers. Readily conforms to irregular shapes. Does not require separate supporting structure and saves space. Not added on to equipment but constitutes an integral part of it. Interconnection wiring deposited on sheet photolithographically or by other suitable masking/fabrication methods. Complete blanket, including cells and interconnections, fabricated as rigid unit directly on, and supported by, nonplanar surface to be covered.

  15. 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).

  16. 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…

  17. 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`.

  18. ITER Fusion Energy

    ScienceCinema

    Dr. Norbert Holtkamp

    2010-01-08

    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.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Resent Status of ITER Equatorial Launcher Development

    NASA Astrophysics Data System (ADS)

    Takahashi, K.; Kajiwara, K.; Kasugai, A.; Oda, Y.; Kobayashi, N.; Sakamoto, K.

    2009-11-01

    The ITER equatorial launcher is divided into a front shield and a port plug. The front shield is composed of fourteen blanket shield modules so as to form three openings for the injection of mm-wave beams into plasma. Twenty-four waveguide transmission lines, internal shields, cooling pipes and so on are installed in the port plug. The transmission lines consist of the corrugated waveguides, miter bends and the free space propagation region utilizing two mirrors in front of the waveguide outlet. The analysis of mm-wave beam propagation in the region shows that the transmission efficiency more than 99.5% is attained. The high power experiments of the launcher mock-up have been carried out and the measured field patterns at each mirror and the outlet of the launcher are agreed with the calculations. It is concluded that the transmission line components in the launcher mock-up are fabricated as designed and the present mm-wave design in the launcher is feasible.

  5. Resent Status of ITER Equatorial Launcher Development

    SciTech Connect

    Takahashi, K.; Kajiwara, K.; Kasugai, A.; Oda, Y.; Kobayashi, N.; Sakamoto, K.

    2009-11-26

    The ITER equatorial launcher is divided into a front shield and a port plug. The front shield is composed of fourteen blanket shield modules so as to form three openings for the injection of mm-wave beams into plasma. Twenty-four waveguide transmission lines, internal shields, cooling pipes and so on are installed in the port plug. The transmission lines consist of the corrugated waveguides, miter bends and the free space propagation region utilizing two mirrors in front of the waveguide outlet. The analysis of mm-wave beam propagation in the region shows that the transmission efficiency more than 99.5% is attained. The high power experiments of the launcher mock-up have been carried out and the measured field patterns at each mirror and the outlet of the launcher are agreed with the calculations. It is concluded that the transmission line components in the launcher mock-up are fabricated as designed and the present mm-wave design in the launcher is feasible.

  6. 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.

  7. Overview of design activities for Li/V blankets

    SciTech Connect

    Sze, D.K.; Mattas, R.F.

    1997-12-31

    Recent fusion power plant design studies in the US have been conducted within the ARIES project. The most recent design of Li/V blankets was conducted as part of the ARIES-RS design. The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on a Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design.

  8. High Power Density Blanket Design Study for Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Huang, J. H.; Zhu, Y. K.; Deng, P. Zh.

    2003-06-01

    A conceptual design study of a high power density blanket has been carried out. The Fusion Experimental Breeder, FEB, is adopted as the reference reactor. The neutron wall loading is 0.5 MW/m2. The blanket is cooled by 10 MPa helium in tube. The concept of LiPb eutectic/transuranium oxide suspension is adopted. The neutronics design is performed to provide the design basis, and it gives an energy multiplication of 37 and a flattened power density distribution with a peak value of 70 W/m3. Multiple cooling panels are introduced to reduce the peak temperature of the blanket. In spite of up to 15 cooling panels, the blanket module is calculated using the ANSYS code and analytically as well. The results are consistent with each other and can meet the thermal criteria. However, structural calculation results from ANSYS did not satisfy the criterion: The blanket structure design is then improved by using curved cooling panels to model the structure in detail. Temperature distribution is obtained using the Pro/Mechanica code. Detailed structural analyses are also done by this code. Some satisfactory results are obtained.

  9. TIG and HIP joining of Reduced Activation Ferrite/Martensitic steel for the Korean ITER-TBM

    NASA Astrophysics Data System (ADS)

    Ku, Duck Young; Oh, Seungjin; Ahn, Mu-Young; Yu, In-Keun; Kim, Duck-Hoi; Cho, Seungyon; Choi, Im-Sub; Kwon, Ki-Bum

    2011-10-01

    Korea is developing a Helium Cooled Solid Breeder Test Blanket Module for ITER. The primary candidate structural material is a Reduced Activation Ferritic/Martensitic steel. The complex TBM structure requires developing joining technologies for successful fabrication. The characteristics of Tungsten Inert Gas welding and Hot Isostatic Pressing joining of RAFM steel were investigated. Metallurgical examinations showed that the steel grain size was increased after HIP joining and recovered by post joining heat treatment. Both TIG welding and HIP joining are found to be acceptable for ITER TBM based on mechanical tests and microstructure examination.

  10. 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.

  11. 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.

  12. Progress in developing high performance solar blankets and arrays

    NASA Technical Reports Server (NTRS)

    Scott-Monck, J.

    1982-01-01

    The development of high efficiency, ultrathin silicon solar cells offers both opportunity and challenge. It is possible to consider 400 W/kg blanket designs by using this cell in conjuction with flexible substrates, ultrathin covers and welded interconnects. By designing array structure which is mechanically and dynamically compatible with very low mass blankets, solar arrays with a specific power approaching 200 W/kg are achievable. Further improvements in blanket performance (higher power and lower mass per unit area), which could come from the implementation of higher efficiency cells operating at lower temperatures (silicon or GaAs), and the use of encapsulants, would result in the development of 300 W/kg solar arrays.

  13. The evolution of US helium-cooled blankets

    NASA Astrophysics Data System (ADS)

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

    1991-08-01

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

  14. 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. 

  15. Overview of the TFTR Lithium Blanket Module Program

    SciTech Connect

    Jassby, D.L.

    1986-01-01

    The Lithium Blanket Module (LBM) is an {approximately}80-cm{sup 3} module, representative of a helium-cooled lithium oxide fusion reactor blanket module. This paper summarizes the design, development, and construction of the LBM, and indicates the present status of the LBM program. Construction of the LBM provided unique development and manufacturing experience with the mass production of reactor-representative lithium oxide pellets and fuel rods. Neutron activation and tritium assay data from present irradiation experiments with a point-neutron source and future experiments with the TFTR geometrically extended neutron source will reveal the ability to neutronics codes and models to characterize individual blanket module performance in a fusion device assembly.

  16. 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.

  17. 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…

  18. Adaptive iterative reconstruction

    NASA Astrophysics Data System (ADS)

    Bruder, H.; Raupach, R.; Sunnegardh, J.; Sedlmair, M.; Stierstorfer, K.; Flohr, T.

    2011-03-01

    It is well known that, in CT reconstruction, Maximum A Posteriori (MAP) reconstruction based on a Poisson noise model can be well approximated by Penalized Weighted Least Square (PWLS) minimization based on a data dependent Gaussian noise model. We study minimization of the PWLS objective function using the Gradient Descent (GD) method, and show that if an exact inverse of the forward projector exists, the PWLS GD update equation can be translated into an update equation which entirely operates in the image domain. In case of non-linear regularization and arbitrary noise model this means that a non-linear image filter must exist which solves the optimization problem. In the general case of non-linear regularization and arbitrary noise model, the analytical computation is not trivial and might lead to image filters which are computationally very expensive. We introduce a new iteration scheme in image space, based on a regularization filter with an anisotropic noise model. Basically, this approximates the statistical data weighting and regularization in PWLS reconstruction. If needed, e.g. for compensation of the non-exactness of backprojector, the image-based regularization loop can be preceded by a raw data based loop without regularization and statistical data weighting. We call this combined iterative reconstruction scheme Adaptive Iterative Reconstruction (AIR). It will be shown that in terms of low-contrast visibility, sharpness-to-noise and contrast-to-noise ratio, PWLS and AIR reconstruction are similar to a high degree of accuracy. In clinical images the noise texture of AIR is also superior to the more artificial texture of PWLS.

  19. 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

    ...Based on affirmative final determinations by the Department of Commerce (the ``Department'') and the International Trade Commission (``ITC''), the Department is issuing an antidumping duty order on certain woven electric blankets (``woven electric blankets'') from the People's Republic of China...

  20. 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.

  1. Overview of the TFTR Lithium Blanket Module program

    SciTech Connect

    Jassby, D.L.

    1986-11-01

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

  2. Neutron Dosimetry Tokamak Fusion Test Reactor Lithium Blanket Module

    SciTech Connect

    Tsang, F.Y.; Harker, Y.D.; Anderl, R.A.; Nigg, D.W.; Jassby, D.L.

    1986-11-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-kind neutronics experiment involving a toroidal fusion neutron source. Qualification experiments have been conducted to develop primary measurement techniques and verify dosimetry materials that will be used to characterize the neutron environment inside and on the surfaces of the LBM. The deuterium-tritium simulation experiments utilizing a 14-MeV neutron generator and a fusion blanket mockup facility at the Idaho National Engineering Laboratory are described. Results and discussions are presented that identify the quality and limitations of the measured integral reaction data, including the minimum fluence requirement for the TFTR experiment.

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. Neutron activation system using water flow for ITER

    NASA Astrophysics Data System (ADS)

    Nishitani, T.; Ebisawa, K.; Kasai, S.; Walker, C.

    2003-03-01

    A neutron activation system with flowing water using the 16O(n,p)16N reaction has been designed for the International Thermonuclear Experimental Reaction (ITER) neutron yield monitor with temporal resolution, based on the experimental results carried out at the fusion neutronics source (FNS) facility of the Japan Atomic Energy Research Institute. On ITER, irradiation ends will be installed in the filler shielding module between the blanket modules at the equatorial ports. The gamma-ray counting stations will be installed on the upstairs of the pit outside the biological shield. BGO (Bi4Ge3O12) scintillation detectors will be employed to measure 6.13 MeV gamma rays emitted from 16N. The distance between the irradiation end and the counting station is ˜20 m. The performance of the neutron activation system has been evaluated by using the neutron Monte Carlo code MCNP-4b with the JENDL 3.2 library. The reaction rate of 16O(n,p)16N was calculated not only at the irradiation end but also along the transfer line, which showed that the temporal resolution would be less than the ITER requirement of 100 ms including turbulent diffusion effects for the flow velocity of 10 m/s. With a flow velocity of 10 m/s, this system can measure the fusion power from 50 kW to 1 GW of the ITER operation by using two gamma-ray detectors; one detector faces the water pipe directly, and another has a collimator for higher-neutron yield. Also the calculation shows that the reaction rate is relatively insensitive to the change of the plasma position.

  8. Radioactive waste produced by DEMO and commerical fusion reactors extrapolated from ITER and advanced data bases

    SciTech Connect

    Stacey, W.M.; Hertel, N.E.; Hoffman, E.A.

    1994-12-31

    The potential for providing energy with minimal environmental impact is a powerful motivation for the development of fusion and is the long-term objective of most fusion programs. However, the societal acceptability of magnetic fusion may well be decided in the near-term when decisions are taken on the construction of DEMO to follow ITER (if not when the construction decision is taken on ITER). Component wastes were calculated for DEMOs based on each data base by first calculating reactor sizes needed to satisfy the physics, stress and radiation attenuation requirements, and then calculating component replacement rates based on radiation damage and erosion limits. Then, radioactive inventories were calculated and compared to a number of international criteria for {open_quote}near-surface{close_quote} burial. None of the components in either type of design would meet the Japanese LLW criterion (<1 Ci/m{sup 3}) within 10 years of shutdown, although the advanced (V/Li) blanket would do so soon afterwards. The vanadium first wall, divertor and blanket would satisfy the IAEA LLW criterion (<2 mSv/h contact dose) within about 10 years after shutdown, but none of the stainless steel or copper components would. All the components in the advanced data base designs except the stainless steel vacuum vessel and shield readily satisfy the US extended 10CFR61 intruder dose criterion, but none of the components in the {open_quotes}ITER data base{close_quotes} designs do so. It seems unlikely that a stainless steel first wall or a copper divertor plate could satisfy the US (class C) criterion for near surface burial, much less the more stringent international, criteria. On the other hand, the first wall, divertor and blanket of the V/Li system would still satisfy the intruder dose concentration limits even if the dose criterion was reduced by two orders of magnitude.

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

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... United States, the authorization is conditioned on the holding company, consistent with 18 CFR 385.2005(b... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Applicability, definitions, and blanket authorizations. 33.1 Section 33.1 Conservation of Power and Water Resources...

  10. 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.

  11. 18 CFR 33.1 - Applicability, definitions, and blanket authorizations.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... United States, the authorization is conditioned on the holding company, consistent with 18 CFR 385.2005(b... 18 Conservation of Power and Water Resources 1 2013-04-01 2013-04-01 false Applicability, definitions, and blanket authorizations. 33.1 Section 33.1 Conservation of Power and Water Resources...

  12. GREEN CANE TRASH BLANKETS: INFLUENCE ON RATOON CROPS IN LOUISIANA

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Approximately 75% of Louisiana's 2000 sugarcane crop was harvested with a chopper harvester. A significant portion of the chopper-harvested sugarcane was harvested green, especially early in the season. Information on the impact of the post-harvest, green-cane residue blankets on subsequent ratoo...

  13. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2012-04-01 2011-04-01 true Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND...

  14. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2013-04-01 2013-04-01 false Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND...

  15. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2011-04-01 2011-04-01 false Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND TUBES, AND PROCESSED TOBACCO Bonds...

  16. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2010-04-01 2010-04-01 false Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND...

  17. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2014-04-01 2014-04-01 false Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND...

  18. 75 FR 11557 - Woven Electric Blankets From China

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-11

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

  19. 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.

  20. 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.

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

    SciTech Connect

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

    1982-11-01

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

  2. 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.

  3. 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.

  4. ECRH System For ITER

    SciTech Connect

    Darbos, C.; Henderson, M.; Gandini, F.; Albajar, F.; Bomcelli, T.; Heidinger, R.; Saibene, G.; Chavan, R.; Goodman, T.; Hogge, J. P.; Sauter, O.; Denisov, G.; Farina, D.; Kajiwara, K.; Kasugai, A.; Kobayashi, N.; Oda, Y.; Ramponi, G.

    2009-11-26

    A 26 MW Electron Cyclotron Heating and Current Drive (EC H and CD) system is to be installed for ITER. The main objectives are to provide, start-up assist, central H and CD and control of MHD activity. These are achieved by a combination of two types of launchers, one located in an equatorial port and the second type in four upper ports. The physics applications are partitioned between the two launchers, based on the deposition location and driven current profiles. The equatorial launcher (EL) will access from the plasma axis to mid radius with a relatively broad profile useful for central heating and current drive applications, while the upper launchers (ULs) will access roughly the outer half of the plasma radius with a very narrow peaked profile for the control of the Neoclassical Tearing Modes (NTM) and sawtooth oscillations. The EC power can be switched between launchers on a time scale as needed by the immediate physics requirements. A revision of all injection angles of all launchers is under consideration for increased EC physics capabilities while relaxing the engineering constraints of both the EL and ULs. A series of design reviews are being planned with the five parties (EU, IN, JA, RF, US) procuring the EC system, the EC community and ITER Organization (IO). The review meetings qualify the design and provide an environment for enhancing performances while reducing costs, simplifying interfaces, predicting technology upgrades and commercial availability. In parallel, the test programs for critical components are being supported by IO and performed by the Domestic Agencies (DAs) for minimizing risks. The wide participation of the DAs provides a broad representation from the EC community, with the aim of collecting all expertise in guiding the EC system optimization. Still a strong relationship between IO and the DA is essential for optimizing the design of the EC system and for the installation and commissioning of all ex-vessel components when several

  5. ECRH System For ITER

    NASA Astrophysics Data System (ADS)

    Darbos, C.; Henderson, M.; Albajar, F.; Bigelow, T.; Bomcelli, T.; Chavan, R.; Denisov, G.; Farina, D.; Gandini, F.; Heidinger, R.; Goodman, T.; Hogge, J. P.; Kajiwara, K.; Kasugai, A.; Kern, S.; Kobayashi, N.; Oda, Y.; Ramponi, G.; Rao, S. L.; Rasmussen, D.; Rzesnicki, T.; Saibene, G.; Sakamoto, K.; Sauter, O.; Scherer, T.; Strauss, D.; Takahashi, K.; Zohm, H.

    2009-11-01

    A 26 MW Electron Cyclotron Heating and Current Drive (EC H&CD) system is to be installed for ITER. The main objectives are to provide, start-up assist, central H&CD and control of MHD activity. These are achieved by a combination of two types of launchers, one located in an equatorial port and the second type in four upper ports. The physics applications are partitioned between the two launchers, based on the deposition location and driven current profiles. The equatorial launcher (EL) will access from the plasma axis to mid radius with a relatively broad profile useful for central heating and current drive applications, while the upper launchers (ULs) will access roughly the outer half of the plasma radius with a very narrow peaked profile for the control of the Neoclassical Tearing Modes (NTM) and sawtooth oscillations. The EC power can be switched between launchers on a time scale as needed by the immediate physics requirements. A revision of all injection angles of all launchers is under consideration for increased EC physics capabilities while relaxing the engineering constraints of both the EL and ULs. A series of design reviews are being planned with the five parties (EU, IN, JA, RF, US) procuring the EC system, the EC community and ITER Organization (IO). The review meetings qualify the design and provide an environment for enhancing performances while reducing costs, simplifying interfaces, predicting technology upgrades and commercial availability. In parallel, the test programs for critical components are being supported by IO and performed by the Domestic Agencies (DAs) for minimizing risks. The wide participation of the DAs provides a broad representation from the EC community, with the aim of collecting all expertise in guiding the EC system optimization. Still a strong relationship between IO and the DA is essential for optimizing the design of the EC system and for the installation and commissioning of all ex-vessel components when several teams

  6. 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.

  7. 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.

  8. 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.

  9. Progress and Achievements on the R&D Activities for ITER Vacuum Vessel

    SciTech Connect

    Nakahira, M.; Koizumi, K.; Takahashi, H.; Onozuka, M.; Ioki, K.; Kuzumin, E.; Krylov, V.; Maslakowski, J.; Nelson, Brad E; Jones, L.; Danner, W.; Maisonnier, D.

    2001-01-01

    The ITER vacuum vessel (VV) is designed to be large double-walled structure with a D-shaped crosssection. The achievable fabrication tolerance of this structure was unknown due to the size and complexity of shape. The Full-scale Sector Model of ITER Vacuum Vessel, which was 15m in height, was fabricated and tested to obtain the fabrication and assembly tolerances. The model was fabricated within the target tolerance of 5mm and welding deformation during assembly operation was obtained. The port structure was also connected using remotized welding tools to demonstrate the basic maintenance activity. In parallel, the tests of advanced welding, cutting and inspection system were performed to improve the efficiency of fabrication and maintenance of the Vacuum Vessel. These activities show the feasibility of ITER Vacuum Vessel as feasible in a realistic way. This paper describes the major progress, achievement and latest status of the R&D activities on the ITER vacuum vessel.

  10. Iterative modulo scheduling

    SciTech Connect

    Rau, B.R.

    1996-02-01

    Modulo scheduling is a framework within which algorithms for software pipelining innermost loops may be defined. The framework specifies a set of constraints that must be met in order to achieve a legal modulo schedule. A wide variety of algorithms and heuristics can be defined within this framework. Little work has been done to evaluate and compare alternative algorithms and heuristics for modulo scheduling from the viewpoints of schedule quality as well as computational complexity. This, along with a vague and unfounded perception that modulo scheduling is computationally expensive as well as difficult to implement, have inhibited its corporation into product compilers. This paper presents iterative modulo scheduling, a practical algorithm that is capable of dealing with realistic machine models. The paper also characterizes the algorithm in terms of the quality of the generated schedules as well as the computational incurred.

  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. 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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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).

  18. 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.

  19. 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

  20. Iterative denoising of ghost imaging.

    PubMed

    Yao, Xu-Ri; Yu, Wen-Kai; Liu, Xue-Feng; Li, Long-Zhen; Li, Ming-Fei; Wu, Ling-An; Zhai, Guang-Jie

    2014-10-01

    We present a new technique to denoise ghost imaging (GI) in which conventional intensity correlation GI and an iteration process have been combined to give an accurate estimate of the actual noise affecting image quality. The blurring influence of the speckle areas in the beam is reduced in the iteration by setting a threshold. It is shown that with an appropriate choice of threshold value, the quality of the iterative GI reconstructed image is much better than that of differential GI for the same number of measurements. This denoising method thus offers a very effective approach to promote the implementation of GI in real applications. PMID:25322001

  1. 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.

  2. Evaluation of US demo helium-cooled blanket options

    SciTech Connect

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

    1995-10-01

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

  3. Development of insulating coatings for liquid metal blankets

    SciTech Connect

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

    1994-07-01

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

  4. 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.

  5. 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.

  6. 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.

  7. MHD considerations for a self-cooled liquid lithium blanket

    NASA Astrophysics Data System (ADS)

    Sze, D. K.; Mattas, R. F.; Hull, A. B.; Picologlou, B. F.; Smith, D. L.

    1992-03-01

    The magnetohydrodynamic (MHD) effects can present a feasibility issue for a self-cooled liquid metal blanket of magnetically confined fusion reactors, especially the inboard regime of a tokamak. This pressure drop can be significantly reduced by using an insulated wall structure. A self-healing insulating coating has been identified, which will reduce the pressure drop by more than a factor of 10. The future research direction to further quantify the performance of this coating is also outlined.

  8. MFTF-B Upgrade for blanket-technology testing

    SciTech Connect

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

    1982-10-22

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

  9. 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.

  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.