WorldWideScience

Sample records for advanced blanket performance

  1. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

    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

  2. Limitations on blanket performance

    International Nuclear Information System (INIS)

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

  3. Development of advanced blanket performance under irradiation and system integration through JUPITER-II project

    International Nuclear Information System (INIS)

    The Japan-USA collaborative program, JUPITER-II, has made significant progress in a research program titled 'The irradiation performance and system integration of advanced blanket' through a six-year plan for 2001-2006. The scientific concept of this program is to study the elemental technology in macroscopic system integration for advanced fusion blankets based on an understanding of the relevant mechanics at the microscopic level. The program has four main research emphases: (1)Flibe molten salt system: Flibe handling, reduction-oxidation control by Be and Flibe tritium chemistry; thermofluid flow simulation experiment and numerical analysis. (2)Vanadium /Li system: MHD ceramics coating of vanadium alloys and compatibility with Li; neutron irradiation experiment in Li capsule and radiation creep. (3)SiC/He system: Fabrication of advanced composites and property evaluation; thermomechanics of SiC system with solid breeding materials; neutron irradiation experiment in He capsule at high temperatures. (4)Blanket system modeling: Design-based integration modeling of Flibe system and V/Li system; multiscale materials system modeling including He effects. This paper describes the perspective of the program including the historical background, the organization and facilities, and the task objectives. Important recent results are reviewed

  4. Tailorable Advanced Blanket Insulation (TABI)

    Science.gov (United States)

    Sawko, Paul M.; Goldstein, Howard E.

    1987-01-01

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

  5. Lithium-cooled blankets for advanced tokamaks

    International Nuclear Information System (INIS)

    The main objective of the Tokamak Power System Studies (TPSS) at Argonne National Lab. during fiscal year 1985 was to explore innovative design concepts that have the potential for significant enhancement of the attractiveness of a tokamak-based power plant. Activities in the area of plasma engineering resulted in a reference reactor concept, which served as a model for the impurity control and first-wall/blanket/shield studies. The liquid-metal-cooled first-wall/blanket/shield design activity was centered around the vanadium alloy structure and liquid-lithium coolant leading blanket concept as identified by the Blanket Comparison and Selection Study (BCSS). A ferritic steel structure and a LiPb breeder were considered as backup options. The magnetohydrodynamics (MHD) effects associated with self-cooled liquid-metal blanket/first-wall systems are substantially reduced by the lower magnetic fields required for higher plasmas, the lower neutron wall loading resulting from reduced power output, and the smaller reactor size of the TPSS model reactor. Therefore, improved performance characteristics of self-cooled liquid-metal blanket concepts are achievable mainly because the design constraints are more relaxed compared to the BCSS guidelines. Key aspects of the designs evaluated in the current study include the following: (1) design simplicity; (2) use of the first wall as an impurity control device; (3) modular first-wall/blanket/reflector/shield construction; and (4) integrated first-wall/blanket/reflector/shield

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

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  7. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this project advanced acoustic blankets for improved low frequency interior noise control in aircraft will be developed and demonstrated. The improved...

  8. Advanced Polymer For Multilayer Insulating Blankets

    Science.gov (United States)

    Haghighat, R. Ross; Shepp, Allan

    1996-01-01

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

  9. Range of blanket concepts from near term solutions to advanced concepts

    International Nuclear Information System (INIS)

    Breeding blankets are key components of fusion power plants and determine to a large degree their attractiveness. There is a large range of possible blanket concepts, characterized by breeding material, structural material, coolant, and geometrical arrangement. On one extreme side there are concepts requiring only a modest extrapolation of the present technology, but with limited attractiveness. On the other side there are concepts with the potential for very attractive plants, but involving a considerable development risk. Therefore, the selection of blanket concepts depends on the overall strategy for fusion power plant development. To ensure sufficient support over the long development time, it has to be shown in a credible way, that (a) there are feasible blanket concepts for fusion power plants which can be developed with a high confidence of success, and (b) the final product will be an attractive power plant which can be operated safely, with minimum impact on the environment, and at an acceptable cost of electricity. This requires an evaluation of the entire range of blanket concepts, starting from near term solutions and reaching up to very advanced designs with a potential for exceptionally high performance. Typical examples of candidate blanket concepts are compared in this study. The blanket development strategy developed in the EU is described as an example

  10. Tailorable advanced blanket insulation using aluminoborosilicate and alumina batting

    Science.gov (United States)

    Calamito, Dominic P.

    1989-01-01

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

  11. Nuclear Performance Analyses for HCPB Test Blanket Modules in ITER

    International Nuclear Information System (INIS)

    The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in the framework of the European Fusion Technology Programme for performance tests in ITER. The related efforts currently focus on the design optimisation of suitable Test Blanket Modules (TBM) and associated R-and-D activities. Four different HCPB TBM types are considered for addressing issues related to (i) electromagnetic transients (EM), (ii) neutronics and Tritium (NT), (iii) thermo-mechanical properties of the pebble beds (TM), and (iv) the integral performance of the blanket module (Plant Integration, PI). The lay-out of the NT and the PI modules has been entirely revised to represent the latest HCPB breeder blanket concept for fusion power reactors. A HCPB TBM consists of a steel box with an internal stiffening grid and small breeder units. The stiffening grid forms radially running open cells accommodating the breeder units (BU). The BU consists of a back plate with attached breeder canisters providing space for the breeder pebble beds. The space between the canisters and the stiffening plates is filled with Beryllium pebbles for the neutron multiplication. The latest design assumes two vertically arranged breeder containers per BU with a toroidal bed height of 10 and 24 mm, for NT and PI modules, respectively. Li4SiO4 is assumed as breeder material at 6Li enrichment levels between 40 at % (NT) and 90 at % (PI). This work is devoted to the neutronic, shielding and activation analyses performed recently for NT and PI variants of the HCPB TBM in ITER. The analyses are based on three-dimensional neutronic and activation calculations making use of a 20 degree torus sector model of ITER developed for Monte Carlo calculations with the MCNP code. The model includes a proper representation of the horizontal ITER test blanket port, the water cooled support frame with two integrated HCPB blanket test modules, the radiation shield and the port environment. Monte Carlo

  12. Nuclear performance analyses for HCPB test blanket modules in ITER

    International Nuclear Information System (INIS)

    Neutronic, shielding and activation analyses have been performed for recent design variants of the Helium Cooled Pebble Bed (HCPB) test blanket module (TBM) in ITER on the basis of 3D Monte Carlo calculations. The main objective has been to assess and optimise the nuclear performance of the HCPB test blanket modules in terms of the tritium generation, the nuclear heating and the radiation shielding and provide, among others, the data required for the engineering design of the test modules. The shielding efficiency of the TBM system was shown to be sufficient to allow access of work personnel to the port extension after a waiting time of 10 days after shut down as required by ITER. The activation analyses provided the afterheat and activation data for quality assured safety analyses assuming a representative irradiation scenario

  13. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the proposed Phase II research effort is to develop heterogeneous (HG) blankets for improved sound reduction in aircraft structures. Phase I...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-04

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

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

    International Nuclear Information System (INIS)

    A review of worldwide results revealed that reaction rates in the blanket region are generally underpredicted with the discrepancy increasing with penetration; however, these results vary widely. Experiments in the large uniform Purdue Fast Breeder Blanket Facility (FBBF) blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50 group cross sections), a consistent Calculated/Experimental (C/E) drop-off was observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the blanket is necessary for agreement with experiments. The usefulness of refined group constant generation utilizing specialized weighting spectra and transport theory methods in correcting this discrepancy was analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The surprising result was that transport methods had no effect on the blanket deviations; thus, transport theory considerations do not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by some approximations which are applied in all current methods. 27 refs., 3 figs., 1 tab

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

    Science.gov (United States)

    Bosia, G.; Ragona, R.

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bosia, G., E-mail: gbosia@to.infn.it [Department of Physics, University of Turin (Italy); Ragona, R. [Laboratory for Plasma Physics-LPP-ERM/KMS, Brussels (Belgium)

    2015-12-10

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

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

    International Nuclear Information System (INIS)

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

  19. A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    Highlights: • SFR burner core configurations are explored and analyzed for effective use of thorium blankets. • Thorium blankets can significantly improve SFR burner core performances. • No recycling or partial recycling of Th blankets with multi-batches is very effective. - Abstract: In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate

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

    International Nuclear Information System (INIS)

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

  1. Advanced sodium cooled reactor cores having thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    In this paper, a design concept of 400 MWe sodium cooled fast reactor (SFR) cores having thorium blankets for effective burning of TRU (Transuranics) from LWR spent fuel is described. Specifically, we considered two recycling options of thorium blankets : 1) no recycling and 2) fully recycling. The thorium blankets are loaded in the axially central regions of the core regions and their axial heights are adjusted so as to increase TRU burning rate and to reduce burnup reactivity swing. Also, we analyzed the performances of the cores having different fuel management batch sizes and different recycling options for the searched core configuration. The results show that the axial thorium blankets with no recycling option can be effectively used to increase TRU burning rate with a significant reduction of burnup reactivity swing in comparison with typical SFR burner cores having no blankets while the recycling of thorium blanket degrades TRU burning rate and burnup reactivity swing but it leads to a reduction of sodium void worth and more negative Doppler coefficient. (author)

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

    International Nuclear Information System (INIS)

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

  3. Development of tailorable advanced blanket insulation for advanced space transportation systems

    Science.gov (United States)

    Calamito, Dominic P.

    1987-01-01

    Two items of Tailorable Advanced Blanket Insulation (TABI) for Advanced Space Transportation Systems were produced. The first consisted of flat panels made from integrally woven, 3-D fluted core having parallel fabric faces and connecting ribs of Nicalon silicon carbide yarns. The triangular cross section of the flutes were filled with mandrels of processed Q-Fiber Felt. Forty panels were prepared with only minimal problems, mostly resulting from the unavailability of insulation with the proper density. Rigidizing the fluted fabric prior to inserting the insulation reduced the production time. The procedures for producing the fabric, insulation mandrels, and TABI panels are described. The second item was an effort to determine the feasibility of producing contoured TABI shapes from gores cut from flat, insulated fluted core panels. Two gores of integrally woven fluted core and single ply fabric (ICAS) were insulated and joined into a large spherical shape employing a tadpole insulator at the mating edges. The fluted core segment of each ICAS consisted of an Astroquartz face fabric and Nicalon face and rib fabrics, while the single ply fabric segment was Nicalon. Further development will be required. The success of fabricating this assembly indicates that this concept may be feasible for certain types of space insulation requirements. The procedures developed for weaving the ICAS, joining the gores, and coating certain areas of the fabrics are presented.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

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

    International Nuclear Information System (INIS)

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

  7. Neutronics performance of high-temperature refractory alloy helium-cooled blankets for fusion application

    International Nuclear Information System (INIS)

    Among the concepts considered in the advanced power extraction (Apex) study is the helium-cooled refractory metal FW and blanket concept. Refractory metals exhibit high operating temperature and can offer good capability for withstanding high power density operation that is the focus of the APEX study. In this paper, we assess the impact of using various refractory metals on the nuclear heating profiles across the blanket and power multiplication (PM) and on the tritium breeding profiles and tritium breeding ratio (TBR). The refractory metals considered with liquid lithium breeder are W, TZM, and Nb-1Zr. The impact of Li-6 enrichment on these profiles and on TBR and PM is also assessed. Comparison of these nuclear characteristics is also made to other liquid breeder (Flibe and Li-Sn). Because the moderation power of these breeders to neutron energy varies among them, the damage to the structure is different with various structure/breeder combinations. The damage parameters (DPA, helium and hydrogen production) at key locations are also compared with the corresponding values in the thick liquid FW/blanket concept; an innovative design concept under consideration within the APEX study

  8. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

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

  9. Nuclear performance optimization of the Be/Li/Th blanket for the fusion breeder

    International Nuclear Information System (INIS)

    More rigorous nuclear analysis, including treatment of resonance self-shielding effects coupled with an optimization procedure, has resulted in improved performance of the Be/Li/Th blanket. Net U-233 breeding ratio has increased 36% (to 0.84) while at an average U-233/Th ratio of 0.5 a/o average energy multiplication has increased only 12% (to 2.1) compared with earlier results

  10. Materials for breeding blankets

    International Nuclear Information System (INIS)

    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

  11. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

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

  12. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  13. Analysis of deficiencies in fast reactor blanket physics predictions

    International Nuclear Information System (INIS)

    This analysis addresses a deviation between experimental measurements and fast reactor blanket physics predictions. A review of worldwide results reveals that reaction rates in the blanket are underpredicted with the discrepancy increasing with penetration into the blanket. The analysis of this discrepancy involves two parts: quantifying possible error reductions using the most advanced methods and investigating deficiencies in current methodology. The source of these discrepancies was investigated by application of ''state-of-the-art'' group constant generation and flux prediction methodology to flux calculations for the Purdue University Fast Breeder Blanket Facility (FBBF). Refined group constant generation methods yielded a significant reduction in the blanket deviations; however, only about half of the discrepancy can be accounted for in this manner. Transport theory calculations were used to predict the blanket neutron transmission problem. The surprising result is that transport theory predictions utilizing diffusion theory group constants did not improve the blanket results. Transport theory predictions exhibited blanket underpredictions similar to the diffusion theory results. The residual blanket discrepancies not explained using advanced methods require a refinement of the theory. For this purpose an analysis of deficiencies in current methodology was performed

  14. Effect of channel wall conductance on the performance characteristics of self-cooled liquid metal fusion reactor blankets

    International Nuclear Information System (INIS)

    One of the critical issues in self-cooled liquid metal tritium breeding blankets in magnetically confined fusion reactors is strong MHD effects particularly when the channel walls are not electrically insulated from the flowing liquid metals. Another critical issue is the cooling of the first wall which is subjected to intense heat load from the fusion plasma. In this work we investigate the effect of channel wall conductance on the friction factor and Nusselt number. It is shown by solving the indication and linear momentum equations that even for relatively small channel wall conductance ratios, the friction factor increases by an order of magnitude for the typical Hartmann numbers encountered in fusion reactor blankets. Furthermore, by solving the temperature equation, it is shown that channel wall conductance has negligible effect on Nusselt number in spite of high velocity jets developing near the side walls. Taking into account these limitations, it is shown however, that the self-cooled liquid metal blankets remain a feasible proposition for both first wall heat extraction and bulk heat removal from the blanket. The most important thermal-hydraulic performance parameter -the heat removal rate to pumping power ratio- can still be kept quite high by suitably choosing the design variables of the liquid metal cooling system. The results are presented and compared for the three prime candidates for self-cooled liquid metal breeding blankets, i.e., lithium, lead-lithium, and tin-lithium alloys. (author)

  15. Conceptual study on high performance dual-cooled blanket in a spherical tokamak fusion-driven transmuter

    International Nuclear Information System (INIS)

    A preliminary conceptual design of high performance dual-cooled blanket in a spherical tokamak fusion-driven transmuter has been proposed based on the core D-T plasma parameter level achieved or to be achieved in the near future. The calculation shows that this kind of blanket is tritium self-sustainable and could safely transmute the long-lived actinides produced by 102 GWe·year PWRs, with several tons of fission products per year and 11600 MW thermal power output

  16. The effect of self-shielding of resonance cross sections on the performance of some promising fusion blanket designs

    International Nuclear Information System (INIS)

    The effect of self-shielding of resonance cross sections on the tritium breeding ratio was investigated for three promising fusion blanket designs with liquid lithium, lithium oxide and lithium-lead breeders. Calculations were performed using ANISN and MCNP transport codes with the ENDF/B-V based nuclear data libraries. It is found that the self-shielding effect cannot be neglected in the blanket design if the blanket is neutron leaky in the case when the blanket is thin or with lower Li-6 enrichment in Li. This may result in an underestimate of the tritium breeding ratio if the cross sections are infinitely diluted. This is due to the resonances in the structure materials in which the absorption cross sections are enhanced in the infinitely diluted case. Thus the effect of self-shielding of resonance cross sections should be considered in neutronics calculations of fusion reactors. It is shown that the MCNP results are better reproduced by those from the transport code with the infinitely diluted library. This is probably due to the weight function used to generate the library and to the number of groups considered. Thus for fusion applications it is recommanded to collapse broad group cross sections with the spectrum obtained from an accurate calculation based on many fine groups. (author)

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

    Science.gov (United States)

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

    2013-09-01

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

  18. Performance analysis of upflow anaerobic sludge blanket reactors in the treatment of swine wastewater

    Directory of Open Access Journals (Sweden)

    Luiz A. V. Sarmento

    2007-07-01

    Full Text Available The adoption of confined systems for swine production have been increased the use of water in these installations and, consequently, an each time greater production of wastewater. Diagnostics have been showed a high level of water pollution due the waste material release on lands without criterions and in waters without previous treatment. The utilization of anaerobic process to reduce the liquid residues pollutant power has been detaching because beyond reducing the environmental pollution they allow to recover the energetic potential as fertilizer and biogas. In this work the performance of two real scale upflow anaerobic sludge blanket reactors treating swine wastewater were evaluated through operational system analysis, physical-chemical parameters of pollution and biogas production measurement. The results permitted to verify upflow rate speeds above of the value for which these reactors were designed and hydraulic residence times under of the design value. These factors affected negatively the treatment and had reflected on the law removal of the physical-chemical parameters and biogas production. The maximum removal efficiencies reached for TSS, BOD and COD were 72,5%, 34,7% and 40,0%, respectively. The mean rate of biogas liberation was 0,011 m-³ m-².h-1.

  19. Influence of nuclear data uncertainties on thorium fusion-fission hybrid blanket nucleonic performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, E.T.; Mathews, D.R.

    1979-09-01

    The fusion-fission hybrid blanket proposed for the Tandem Mirror Hybrid Reactor employs thorium metal as the fertile material. Based on the ENDF/B-IV nuclear data, the /sup 233/U and tritium production rate and blanket energy multiplication averaged over the blanket lifetime of about 9 MW-yr/m/sup 2/ are 0.76 and 1.12 per D-T neutron and 4.8, respectively. At the time of the blanket discharge, the /sup 233/U enrichment in the thorium metal is about 3%. The thorium cross sections given by the ENDF/B-IV and V were reviewed, and the important partial cross sections such as (n,2n), (n,3n), and (n,..gamma..) were found to be known to +-10 to 20% in the respective energy range of interest. A sensitivity study showed that the /sup 233/U and tritium production rate and blanket energy multiplication are relatively sensitive to the thorium capture and fission cross section uncertainties. In order to predict the above parameters within +-1%, the Th(n,..gamma..) and Th(n,..nu..f) cross sections must be measured within about +-2% in the energy range 3 to 3000 keV and 13.5 to 15 MeV, respectively.

  20. Influence of nuclear data uncertainties on thorium fusion-fission hybrid blanket nucleonic performance

    International Nuclear Information System (INIS)

    The fusion-fission hybrid blanket proposed for the Tandem Mirror Hybrid Reactor employs thorium metal as the fertile material. Based on the ENDF/B-IV nuclear data, the 233U and tritium production rate and blanket energy multiplication averaged over the blanket lifetime of about 9 MW-yr/m2 are 0.76 and 1.12 per D-T neutron and 4.8, respectively. At the time of the blanket discharge, the 233U enrichment in the thorium metal is about 3%. The thorium cross sections given by the ENDF/B-IV and V were reviewed, and the important partial cross sections such as (n,2n), (n,3n), and (n,γ) were found to be known to +-10 to 20% in the respective energy range of interest. A sensitivity study showed that the 233U and tritium production rate and blanket energy multiplication are relatively sensitive to the thorium capture and fission cross section uncertainties. In order to predict the above parameters within +-1%, the Th(n,γ) and Th(n,νf) cross sections must be measured within about +-2% in the energy range 3 to 3000 keV and 13.5 to 15 MeV, respectively

  1. The EU advanced lead lithium blanket concept using SiCf/SiC flow channel inserts as electrical and thermal insulators

    International Nuclear Information System (INIS)

    Preparatory work on the EU advanced dual coolant (A-DC) blanket concept using SiCf/SiC flow channel inserts as electrical and thermal insulators has been carried out at the Forschungszentrum Karlsruhe in co-operation with CEA (SiCf/SiC composite-related issues) as a conceptual design proposal to the EU fusion power plant study planned to be launched in 2001 within the framework of the EU fusion programme having the main objective of specifying the characteristics of an attractive and viable commercial D-T fusion power plant. The basic principles and the study method for the A-DC blanket concept are presented in this report. The results of this study show that the A-DC blanket concept has a high potential for further development due to its high thermal efficiency and its simple concept solution

  2. TRISO-Fuel Element Performance Modeling for the Hybrid LIFE Engine with Pu Fuel Blanket

    Energy Technology Data Exchange (ETDEWEB)

    DeMange, P; Marian, J; Caro, M; Caro, A

    2010-02-18

    A TRISO-coated fuel thermo-mechanical performance study is performed for the hybrid LIFE engine to test the viability of TRISO particles to achieve ultra-high burnup of a weapons-grade Pu blanket. Our methodology includes full elastic anisotropy, time and temperature varying material properties for all TRISO layers, and a procedure to remap the elastic solutions in order to achieve fast fluences up to 30 x 10{sup 25} n {center_dot} m{sup -2} (E > 0.18 MeV). In order to model fast fluences in the range of {approx} 7 {approx} 30 x 10{sup 25} n {center_dot} m{sup -2}, for which no data exist, careful scalings and extrapolations of the known TRISO material properties are carried out under a number of potential scenarios. A number of findings can be extracted from our study. First, failure of the internal pyrolytic carbon (PyC) layer occurs within the first two months of operation. Then, the particles behave as BISO-coated particles, with the internal pressure being withstood directly by the SiC layer. Later, after 1.6 years, the remaining PyC crumbles due to void swelling and the fuel particle becomes a single-SiC-layer particle. Unrestrained by the PyC layers, and at the temperatures and fluences in the LIFE engine, the SiC layer maintains reasonably-low tensile stresses until the end-of-life. Second, the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Obtaining more reliable measurements, especially at higher fluences, is an imperative for the fidelity of our models. Finally, varying the geometry of the TRISO-coated fuel particles results in little differences in the scope of fuel performance. The mechanical integrity of 2-cm graphite pebbles that act as fuel matrix has also been studied and it is concluded that they can reliable serve the entire LIFE burnup cycle without failure.

  3. Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

    International Nuclear Information System (INIS)

    Spanish Breeding Blanket Technology Programme TECNOFUS is exploring the technological capabilities of a Dual-Coolant He/Pb15.7Li breeding blanket for DEMO and studying new breeding blanket design specifications. The progress of the channel conceptual design is being conducted in parallel with the extension of MHD computational capabilities of CFD tools and the underlying physics of MHD models. A qualification of MHD effects under present blanket design specifications and some approaches to their modelling were proposed by the authors in . The analysis was accomplished with the 2D transient algorithm from Sommeria and Moreau and implemented in the OpenFOAM CFD toolbox . The thermal coupling was implemented by means of the Boussinesq hypothesis. Previous analyses showed the need of improvement of FCI thickness and thermal properties in order to obtain a desirable liquid metal temperature gain of 300 °C. In the present study, an assessment through sensitivity and parametric analyses of the required FCI thickness is performed. Numerical simulations have been carried out considering a Robin-type thermal boundary condition which assumes 1D steady state thermal balance across the solid FCI and Eurofer layers. Such boundary condition has been validated with a fluid–solid coupled domain analysis. Results for the studied flow conditions and channel dimensions show that, in order to obtain a liquid metal temperature gain of about 300 °C, the required FCI material should have a very small effective heat transfer coefficient ((k/δ) ≤ 1 W/m2K) and fluid velocities should be about 0.2 m/s or less. Moreover, special attention has to be placed on the temperature difference across the FCI layer. However, for a maximised liquid metal thermal gain, higher velocities would be preferable, what would also imply a reduced temperature difference across the FCI layer.

  4. Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Mas de les Valls, E., E-mail: elisabet.masdelesvalls@gits.ws [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Heat Engines, Barcelona (Spain); Batet, L. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Physics and Nuclear Engineering, Barcelona (Spain); Medina, V. de [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Sediment Transport Research Group, Department of Engineering Hydraulic, Marine and Environmental Engineering, Barcelona (Spain); Fradera, J. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Physics and Nuclear Engineering, Barcelona (Spain); Sanmarti, M. [bFUS-IREC, Jardins de les Dones de Negre 1, 08930 Sant Adria del Besos (Spain); Sedano, L.A. [EURATOM-CIEMAT Association, 28040 Madrid (Spain)

    2012-08-15

    Spanish Breeding Blanket Technology Programme TECNO{sub F}US is exploring the technological capabilities of a Dual-Coolant He/Pb15.7Li breeding blanket for DEMO and studying new breeding blanket design specifications. The progress of the channel conceptual design is being conducted in parallel with the extension of MHD computational capabilities of CFD tools and the underlying physics of MHD models. A qualification of MHD effects under present blanket design specifications and some approaches to their modelling were proposed by the authors in . The analysis was accomplished with the 2D transient algorithm from Sommeria and Moreau and implemented in the OpenFOAM CFD toolbox . The thermal coupling was implemented by means of the Boussinesq hypothesis. Previous analyses showed the need of improvement of FCI thickness and thermal properties in order to obtain a desirable liquid metal temperature gain of 300 Degree-Sign C. In the present study, an assessment through sensitivity and parametric analyses of the required FCI thickness is performed. Numerical simulations have been carried out considering a Robin-type thermal boundary condition which assumes 1D steady state thermal balance across the solid FCI and Eurofer layers. Such boundary condition has been validated with a fluid-solid coupled domain analysis. Results for the studied flow conditions and channel dimensions show that, in order to obtain a liquid metal temperature gain of about 300 Degree-Sign C, the required FCI material should have a very small effective heat transfer coefficient ((k/{delta}) {<=} 1 W/m{sup 2}K) and fluid velocities should be about 0.2 m/s or less. Moreover, special attention has to be placed on the temperature difference across the FCI layer. However, for a maximised liquid metal thermal gain, higher velocities would be preferable, what would also imply a reduced temperature difference across the FCI layer.

  5. Advances in the optimisation of apparel heating products: A numerical approach to study heat transport through a blanket with an embedded smart heating system

    International Nuclear Information System (INIS)

    The optimisation of the performance of products with smart/active functionalities (e. g. in protective clothing, home textiles products, automotive seats, etc.) is still a challenge for manufacturers and developers. The aim of this study was to optimise the thermal performance of a heating product by a numerical approach, by analysing several opposing requirements and defining solutions for the identified limitations, before the construction of the first prototype. A transfer model was developed to investigate the transport of heat from the skin to the environment, across a heating blanket with an embedded smart heating system. Several parameters of the textile material and of the heating system were studied, in order to optimise the thermal performance of the heating blanket. Focus was put on the effects of thickness and thermal conductivity of each layer, and on parameters associated with the heating elements, e.g. position of the heating wires relative to the skin, distance between heating wires, applied heating power, and temperature range for operation of the heating system. Furthermore, several configurations of the blanket (and corresponding heating powers) were analysed in order to minimise the heat loss from the body to the environment, and the temperature distribution along the skin. The results show that, to ensure an optimal compromise between the thermal performance of the product and the temperature oscillation along its surface, the distance between the wires should be small (and not bigger than 50 mm), and each layer of the heating blanket should have a specific thermal resistance, based on the expected external conditions during use and the requirements of the heating system (i.e. requirements regarding energy consumption/efficiency and capacity to effectively regulate body exchanges with surrounding environment). The heating system should operate in an ON/OFF mode based on the body heating needs and within a temperature range specified based on

  6. Advanced fuel technology and performance

    International Nuclear Information System (INIS)

    The purpose of the Advisory Group Meeting on Advanced Fuel Technology and Performance was to review the experience of advanced fuel fabrication technology, its performance, peculiarities of the back-end of the nuclear fuel cycle with regard to all types of reactors and to outline the future trends. As a result of the meeting recommendations were made for the future conduct of work on advanced fuel technology and performance. A separate abstract was prepared for each of the 20 papers in this issue

  7. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    International Nuclear Information System (INIS)

    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

  8. Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

    Directory of Open Access Journals (Sweden)

    Toshio Wakabayashi

    2013-01-01

    Full Text Available An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

  9. Seed and blanket ADS using thorium–reprocessed fuel: Parametric survey on TRU transmutation performance and safety characteristics

    International Nuclear Information System (INIS)

    Highlights: • The seed and blanket ADS reference configuration without MA enrichment shows a large reactivity swing due to burnup. • The larger amount of MA introducing into the core, the more TRUs are transmuted. • With larger MA content, the smaller reactivity swing in the system can be achieved. • Increasing core size by introducing more thorium and reprocessed fuel assemblies into the system reduces reactivity swing. - Abstract: Conceptual designs of accelerator driven systems (ADS) that utilize thorium fuel as blanket and reprocessed fuel as seed for fission reaction in order to transmute the transuranic elements in spent nuclear fuel and produce energy from thorium utilization was proposed. In order to analyze the TRU transmutation performance of the system while the safety margin during the operation is satisfied, a parametric survey is done in this study. The impact of minor actinide (MA) amount and the reactor size on transmutation efficiency and safety characteristics of the ADS is investigated using MCNPX code using the ENDF/B-VII. As the results, with increasing MA content, TRU transmutation rate is increased while void reactivity of the system becomes less negative. Besides, increasing the core size by introducing more thorium and reprocessed fuel assemblies into the system reduces the TRU transmutation rate and reactivity swing due to burnup of the system. In order to match the efficient transmutation purpose and safety features during operation, higher MA content is suggested to introduce into the system

  10. Status of the extended performance tests for blanket remote maintenance in ITER L6 project

    International Nuclear Information System (INIS)

    Mechanically attached blanket module insertion tests were carried out considering the misalignment between module and back plate. Through the insertion tests, the module was successfully inserted up to the misalignment of ±10 mm under the clearance of ± 0.16 ∼ ±0.18 mm between key and groove. This was achieved by the passive compliance due to the flexibility of the manipulator through the assistance of the chamfer configuration of the key for smooth insertion. In addition, the 'correlation coefficient' based on the results obtained by the strain gages located at the end-effector was found to be useful in order to estimate the forces of the complicated end-effector during module insertion for the development of the sensor based control. (author)

  11. Status of extended performance tests for blanket remote maintenance in the ITER L6 project

    International Nuclear Information System (INIS)

    Mechanically attached blanket module insertion tests were carried out for various misalignments between the module and the back plate. In the insertion tests, the module was successfully inserted up to a misalignment of ±10 mm with a clearance of ±(0.16-0.18) mm between key and groove. This was achieved owing to the passive compliance due to the flexibility of the manipulator with the assistance of the chamfer configuration of the key for smooth insertion. In addition, the 'correlation coefficient' based on the results obtained by the strain gauges located at the end effector was found to be useful for estimating the forces of the complex end effector during module insertion for the development of sensor based control. (author)

  12. Advanced Performance Hydraulic Wind Energy

    Science.gov (United States)

    Jones, Jack A.; Bruce, Allan; Lam, Adrienne S.

    2013-01-01

    The Jet Propulsion Laboratory, California Institute of Technology, has developed a novel advanced hydraulic wind energy design, which has up to 23% performance improvement over conventional wind turbine and conventional hydraulic wind energy systems with 5 m/sec winds. It also has significant cost advantages with levelized costs equal to coal (after carbon tax rebate). The design is equally applicable to tidal energy systems and has passed preliminary laboratory proof-of-performance tests, as funded by the Department of Energy.

  13. Performance evaluation of a novel anaerobic-anoxic sludge blanket reactor for biological nutrient removal treating municipal wastewater.

    Science.gov (United States)

    Díez-Montero, Rubén; De Florio, Loredana; González-Viar, Marta; Herrero, María; Tejero, Iñaki

    2016-06-01

    A novel anaerobic-anoxic sludge blanket reactor, AnoxAn, unifies the non-aerated zones of the biological nutrient removal treatment train in a single upflow reactor, aimed at achieving high compactness and efficiency. The environmental conditions are vertically divided up inside the reactor with the anaerobic zone at the bottom and the anoxic zone above. This contribution presents the performance evaluation of the novel reactor in the removal of organic matter and nutrients from municipal wastewater, coupled with an aerobic hybrid MBR. The overall system achieved total nitrogen and phosphorus removal with average efficiencies of 75% and 89%, respectively. Separate anoxic and anaerobic conditions were maintained in AnoxAn, allowing anaerobic phosphate release and nearly complete anoxic denitrification in the single reactor operating with an HRT of 4.2h. Biomass was retained in the reactor achieving TSS concentration up to 10gL(-1) and partial hydrolysis of influent particulate organic matter. PMID:26970922

  14. Fabrication and performance of AIN insulator coatings for application in fusion reactor blankets

    International Nuclear Information System (INIS)

    The liquid-metal blanket concept for fusion reactors requires an coating on the first-wall structural material to minimize the magnetohydrodynamic pressure drop that occurs during the flow of liquid metal in a magnetic field. Based on the thermodynamics of interactions betwen the coating and the liquid lithium on one side and the structural V-base alloy on the other side, an AIN coating was selected as a candidate. Detailed investigations were conducted on the fabrication, metallurgical microstructure, compatibility in liquid Li, and electrical characteristics of AIN material obtained from several sources. Lithium compatibility was studied in static systems by exposing AIN-coated specimens to liquid Li for several time periods. Electrical resistance was measured at room temperature on the specimens before and after exposure to liquid Li. The results obtained in this study indicate that AIN is a viable coating from the standpoint of chemical compatibility in Li, electrical insulation, and ease of fabrication; for these reasons, the coating should be examined further for fusion reactor applications

  15. Performance experiments with alternative advanced teleoperator control modes for a simulated solar maximum satellite repair

    Science.gov (United States)

    Das, H.; Zak, H.; Kim, W. S.; Bejczy, A. K.; Schenker, P. S.

    1992-01-01

    Experiments are described which were conducted at the JPL Advanced Teleoperator Lab to demonstrate and evaluate the effectiveness of various teleoperator control modes in the performance of a simulated Solar Max Satellite Repair (SMSR) task. THe SMSR was selected as a test because it is very rich in performance capability requirements and it actually has been performed by two EVA astronauts in the Space Shuttle Bay in 1984. The main subtasks are: thermal blanket removal; installation of a hinge attachment for electrical panel opening; opening of electrical panel; removal of electrical connectors; relining of cable bundles; replacement of electrical panel; securing parts and cables; re-mate electrical connectors; closing of electrical panel; and reinstating thermal blanket. The current performance experiments are limited to thermal blanket cutting, electrical panel unbolting and handling electrical bundles and connectors. In one formal experiment even different control modes were applied to the unbolting and reinsertion of electrical panel screws subtasks. The seven control modes are alternative combinations of manual position and rate control with force feedback and remote compliance referenced to force-torque sensor information. Force-torque sensor and end effector position data and task completion times were recorded for analysis and quantification of operator performance.

  16. Breeding blanket for DEMO

    International Nuclear Information System (INIS)

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

  17. Fast breeder reactor blanket management: comparison of LMFBR and GCFR blankets

    International Nuclear Information System (INIS)

    The economic performance of the fast breeder reactor blanket, considering different fuel management schemes was studied. To perform this, the investigation started with a standard reactor physics calculation. Then, two economic models for evaluation of the economic performance of the radial blanket were developed. These models formed the basis of a computer code, ECOBLAN, which computes the net economic gain and the levelized fuel cost due to the radial blanket. The net gain in terms of dollars and $/kgHM-y and the levelized fuel cost in mills/kWhe were obtained as a function of blanket thickness and a residence time of the fuel in the blanket. A LMFBR and a GCFR were the reactor models considered in this study. The optimum radial blanket of a GCFR consists of two rows, that of a LMFBR consists of three rows. Regarding the different fuel management schemes, the fixed blanket was found to be more favorable than reshuffled blanket. Out-in and in-out reshuffled blanket offer almost the same net gain. In all the cases, the burnup calculated for the fuel was found to be less than the acceptable limit. There is an optimum residence time for the fuel in the blanket which depends on the position of the fuel in the blanket and the fuel management scheme studied. As expected, except for very short residence times (less than 2.5 years), the radial blanket is a net income producer. There is no significant difference between the economic performance of the blanket of a LMFBR and a GCFR

  18. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

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

  19. Neutronic evaluation of fissile fuel breeding blankets for the fission-suppressed Tandem-Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    A computational study was performed on the blanket design of the Lawrence Livermore National Laboratory (LLNL) fission-suppressed Tandem Mirror Hybrid Reactor (TMHR) to qualify the methods and data bases available at Oak Ridge National Laboratory (ORNL) for use in analyzing the neutronic performance of fissile fuel breeding blankets. The eventual goal of the study was to establish the capability for analysis and optimization of advanced fissile fuel production blanket designs. Discrete ordinates radiation transport calculations were performed in one-dimensional cylindrical geometry to obtain the blanket spatial distribution and energy spectra of the neutron and gamma-ray fluxes resulting from the monoenergetic (14.1 MeV) fusion first wall source. Key macroscopic cross sections of the blanket materials were then folded with the flux spectra to obtain reaction rates critical to evaluating blanket feasibility. Finally, a time-dependent depletion analysis was performed to evaluate the blanket performance during equilibrium cycle conditions. The results of the study are presented both as graphs and tables

  20. Blanket concept of water-cooled lithium lead with beryllium for the SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    As an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. In SlimCS, the net tritium breeding ratio (TBR) supplied from WCLL blanket was not enough because the thickness of blanket in SlimCS was limited to about 0.5 m so as to allocate the conducting shell position near the plasma for high beta access and vertical stability of plasma. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier to reach a required TBR (≥ 1.05). Considering the operating temperature of blanket materials, a double pipe structure was adopted. The nuclear and thermal analysis were carried out by a nuclear-thermal-coupled code, ANIHEAT and DOHEAT so that blanket materials were appropriately arranged to satisfy the acceptable operation temperatures. The temperatures of materials were kept in appropriate range for the neutron wall load Pn = 5 MW/m2. It was found that the local TBR of WCLL with Be blanket was comparable with that of solid breeder blanket. (author)

  1. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

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

  2. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented

  3. Design and analysis of ITER shield blanket

    International Nuclear Information System (INIS)

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

  4. Performance test of diamond-like carbon films for lubricating ITER blanket maintenance equipment under GPa-level high contact stress

    International Nuclear Information System (INIS)

    Diamond-like carbon (DLC) coating was tested as a candidate solid lubricant for transmission gears of the maintenance equipment of the blanket of the ITER instead of an oil lubricant. The wear tests using the pin-on-disk method were performed on disks with SCM440 and SNCM420 as the base materials and coated with soft, layered, and hard DLCs. All cases satisfied the required allowable contact stress (2 GPa) and lifetime (104 cycles), and therefore the feasibility of the DLC coating was validated. Among the three types of DLCs, the soft DLC showed the best performance. (author)

  5. Development of Solid Breeder Blanket at JAERI

    International Nuclear Information System (INIS)

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

  6. Development of advanced blanket performance under irradiation and system integration through JUPITER-II project

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Katsunori; Kohyama, Akira; Tanaka, Satoru; Namba, C.; Terai, T.; Kunugi, T.; Muroga, Takeo; Hasegawa, Akira; Sagara, A.; Berk, S.; Zinkle, Steven J.; Sze, Dai Kai; Petti, D. A.; Abdou, Mohamed A.; Morley, Neil B.; Kurtz, Richard J.; Snead, Lance L.; Ghoniem, Nasr M.

    2008-12-01

    This report describes an outline of the activities of the JUPITER-II collaboration (japan-USA program of Irradiation/Integration test for Fusion Research-II), Which has bee carried out through six years (2001-2006) under Phase 4 of the collabroation implemented by Amendment 4 of Annex 1 to the DOE (United States Department of Energy)-MEXT (Ministry of Education ,Culture,Sports,Science and Technology) Cooperation. This program followed the RTNS-II Program (Phase1:1982-4986), the FFTF/MOTA Program (Phase2:1987-1994) and the JUPITER Program (Phase 3: 1995-2000) [1].

  7. Advanced Modular Software Performance Monitoring

    International Nuclear Information System (INIS)

    The LHCb software is based on the Gaudi framework, on top of which are built several large and complex software applications. As the LHCb experiment is now in the active phase of collecting and analyzing data, performance problems arise in various parts of the software, from the High Level Trigger (HLT) programs to data analysis frameworks. It is not easy to find hotspots in the code — only specialized tools can help to understand where CPU or memory usage are not reasonable. There exist many performance analyzing tools, but the main problem is that they show reports in terms of class and function names and such information usually is not very useful — the majority of algorithm developers use the Gaudi framework abstractions and usually do not know about functions which lie at the lower level. We will show a new approach which adds to performance reports a higher abstraction level based on knowledge of framework architecture and run-time object properties. A set of profiling tools (based on Intel® VTune™ Amplifier XE) and visualization interfaces has been developed and deployed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA

  10. Advanced Airfoils Boost Helicopter Performance

    Science.gov (United States)

    2007-01-01

    Carson Helicopters Inc. licensed the Langley RC4 series of airfoils in 1993 to develop a replacement main rotor blade for their Sikorsky S-61 helicopters. The company's fleet of S-61 helicopters has been rebuilt to include Langley's patented airfoil design, and the helicopters are now able to carry heavier loads and fly faster and farther, and the main rotor blades have twice the previous service life. In aerial firefighting, the performance-boosting airfoils have helped the U.S. Department of Agriculture's Forest Service control the spread of wildfires. In 2003, Carson Helicopters signed a contract with Ducommun AeroStructures Inc., to manufacture the composite blades for Carson Helicopters to sell

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-09-01

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

  12. Design of test blanket system for ITER module testing

    International Nuclear Information System (INIS)

    Test blanket systems to be installed in ITER for developing demo blankets have been investigated. One of the main engineering goals of ITER is to test tritium breeding blankets relevant to a power reactor. The test foreseen on modules include the demonstration of a breeding capability that would lead to tritium self-sufficiency in a reactor and extraction of a high grade heat suitable for electricity generation. To accomplish these goals, several ITER equatorial ports are available to test the test blanket systems, both in the basic performance phase (BPP) and the enhanced performance phase (EPP). Test blanket systems for water-cooled and helium-cooled type DEMO blankets with ceramic breeders, developed in Japan have been designed. The design activities include the neutronics, thermal and hydraulic analyses, and mechanical configuration considering port sharing, cooling systems and tritium recovery systems, and test blanket system compatible with the current ITER design has been developed. (author)

  13. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

    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

  14. Multivariable optimization of fusion reactor blankets

    International Nuclear Information System (INIS)

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% 6Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO2 breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO2 breeding blanket enriched to 34% 6Li

  15. Improved dechlorinating performance of upflow anaerobic sludge blanket reactors by incorporation of Dehalospirillum multivorans into granular sludge

    DEFF Research Database (Denmark)

    Hörber, Christine; Christiansen, Nina; Arvin, Erik;

    1998-01-01

    Dechlorination of tetrachloroethene, also known as perchloroethylene (PCE), was investigated in an upflow anaerobic sludge blanket (UASB) reactor after incorporation of the strictly anaerobic, reductively dechlorinating bacterium Dehalospirillum multivorans into granular sludge. This reactor...... was compared to the reference 1 (R1) reactor, where the granules were autoclaved to remove all dechlorinating abilities before inoculation, and to the reference 2 (R2) reactor, containing only living granular sludge. All three reactors were fed mineral medium containing 3 to 57 mu M PCE, 2 mM formate, and 0.......5 mM acetate and were operated under sterile conditions. In the test reactor, an average of 93% (mole/mole) of the effluent chloroethenes was dichloroethene (DCE), compared to 99% (mole/mole) in the Ri reactor. The R2 reactor, with no inoculation, produced only trichloroethene (TCE), averaging 43...

  16. High Performance Grinding and Advanced Cutting Tools

    CERN Document Server

    Jackson, Mark J

    2013-01-01

    High Performance Grinding and Advanced Cutting Tools discusses the fundamentals and advances in high performance grinding processes, and provides a complete overview of newly-developing areas in the field. Topics covered are grinding tool formulation and structure, grinding wheel design and conditioning and applications using high performance grinding wheels. Also included are heat treatment strategies for grinding tools, using grinding tools for high speed applications, laser-based and diamond dressing techniques, high-efficiency deep grinding, VIPER grinding, and new grinding wheels.

  17. Performance Enhancements for Advanced Database Management Systems

    OpenAIRE

    Helmer, Sven

    2000-01-01

    New applications have emerged, demanding database management systems with enhanced functionality. However, high performance is a necessary precondition for the acceptance of such systems by end users. In this context we developed, implemented, and tested algorithms and index structures for improving the performance of advanced database management systems. We focused on index structures and join algorithms for set-valued attributes.

  18. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li17Pb83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li17Pb83 blankets. (author)

  19. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  20. ITER breeding blanket module design and analysis

    International Nuclear Information System (INIS)

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

  1. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m2) the average neutron power load on the first wall is below 1 MWm.2, which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  2. Monitoring performance of the Advanced Light Source

    OpenAIRE

    Byrne, Warren E.; Lampo, Edward J.; Samuelson, Bruce C.

    2001-01-01

    Providing high quality light to users in a consistent and reliable manner is one of the main goals of the accelerator physics group at the Advanced Light source (ALS). To meet this goal considerable time is spent monitoring the performance of the machine. At the Group's weekly meeting the performance of the accelerator over the previous week's run is reviewed. This paper describes the parameters that are monitored to optimize the performance of the ALS.

  3. Development of high temperature fusion blanket with LiPb-SiC and its socio-economic aspects

    International Nuclear Information System (INIS)

    This paper describes recent results of the development of SiC-LiPb blanket in Kyoto University with advanced SiC composite, and its implication for high temperature blanket preferable from socio-economic aspects. Blanket is the interface between fusion energy and outside world, i.e., industry, public and environment. Material and energy balances, such as fuel supply and waste discharge, or supply of product energy from fusion plants are characterized by the specific blanket concepts, and performance and characteristics of blankets are considered to dominate the feature of fusion energy that should respond to the requirements of the sponsors and public. Thus development strategy for blanket concepts are affected by the aspects of socio-economics such as; environmental release, assumed accidental scenario, rad-waste, and deployment to the market in each countries/area. Selections of breeders, coolants, and structural materials are based on such consideration, and ITER/TBM is expected to foster the evolving concepts. Combination of LiPb, helium and SiC is of particular interests for a demo blanket concept, because it is expected to be feasible in early stage of development such as TBM by Dual Coolant Lithium Lead concept as an intermediate step, and would eventually achieve high operating temperature above current fission reactors'. In the power plant design in Europe, US and Japan, Lithium lead will work as breeder and primary heat medium, and SiC composites are used for flow insert, enclosure and Intermediate Heat Exchanger to generate high temperature helium that is considered as a medium for power generation including possible hydrogen production. Staged development strategy can be planned; and experiments to verify compatibility, tritium permeability, MHD pressure drop and heat transfer characteristics are being evaluated with LiPb-He loop in Kyoto University. Hydrogen production process with this blanket is also experimentally studied. Based on the results

  4. Thermal-hydraulic performance and structural thermal stress analysis for ITER shield blanket module nearby NB rejoin

    International Nuclear Information System (INIS)

    Hydraulic and thermal analysis of the International Thermonuclear Experimental Reactor (ITER) standard neutral beam (NB) blanket module was carried out in order to check whether the latest design meets ITER requirements. Minor-loss coefficients were estimated with a CFD code, and friction factors of straight channels were obtained using existing formulas. The effects of different radial hole's diameter, length of the back of the radial hole, size of clearance, type of flow driver, branch velocity and flow direction on minor-loss coefficients for radial holes were investigated. Since total mass flow rate and dimensions of the cooling channels were given, when pressure drop due to intersection of the radial hole with back drilled collector was ignored, we can obtain pressure drop, flow rate, velocity and heat transfer coefficient in each radial hole. An improved calculation without neglecting the pressure drop caused by the intersection was also done to compare with the simplified one. Finally, maximum temperature, thermal stress and deformation were evaluated according to FEM thermal analysis. The results of the latest hydraulic and thermal analysis indicate that the current design meets ITER requirements well, except that flow distribution is not so uniform when different types of flow drivers are used, and temperature in the front head surface is a little high. Improved design is necessary in the further. (authors)

  5. TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

    2011-11-30

    This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

  6. Divertor and gas blanket impurity control study

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

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

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  13. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  14. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

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

  15. Neutronics Assessment of Molten Salt Breeding Blanket Design Options

    International Nuclear Information System (INIS)

    Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be ∼1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (∼3 kg) over the blanket lifetime of ∼3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied

  16. Modeling the performance of 'up-flow anaerobic sludge blanket' reactor based wastewater treatment plant using linear and nonlinear approaches-A case study

    International Nuclear Information System (INIS)

    The paper describes linear and nonlinear modeling of the wastewater data for the performance evaluation of an up-flow anaerobic sludge blanket (UASB) reactor based wastewater treatment plant (WWTP). Partial least squares regression (PLSR), multivariate polynomial regression (MPR) and artificial neural networks (ANNs) modeling methods were applied to predict the levels of biochemical oxygen demand (BOD) and chemical oxygen demand (COD) in the UASB reactor effluents using four input variables measured weekly in the influent wastewater during the peak (morning and evening) and non-peak (noon) hours over a period of 48 weeks. The performance of the models was assessed through the root mean squared error (RMSE), relative error of prediction in percentage (REP), the bias, the standard error of prediction (SEP), the coefficient of determination (R2), the Nash-Sutcliffe coefficient of efficiency (Ef), and the accuracy factor (Af), computed from the measured and model predicted values of the dependent variables (BOD, COD) in the WWTP effluents. Goodness of the model fit to the data was also evaluated through the relationship between the residuals and the model predicted values of BOD and COD. Although, the model predicted values of BOD and COD by all the three modeling approaches (PLSR, MPR, ANN) were in good agreement with their respective measured values in the WWTP effluents, the nonlinear models (MPR, ANNs) performed relatively better than the linear ones. These models can be used as a tool for the performance evaluation of the WWTPs.

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

    Directory of Open Access Journals (Sweden)

    A. Kirichek

    2013-12-01

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

  18. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

    Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10-4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

  19. Breeding blanket concepts for fusion and materials requirements

    International Nuclear Information System (INIS)

    This paper summarizes the design and performances of recent breeding blanket concepts and identifies the key material issues associated with them. An assessment of different classes of concepts is carried out by balancing out the potential performance of the concepts with the risk associated with the required material development. Finally, an example strategy for blanket development is discussed

  20. Impact of temperature on feed-flow characteristics and filtration performance of an upflow anaerobic sludge blanket coupled ultrafiltration membrane treating municipal wastewater.

    Science.gov (United States)

    Ozgun, Hale; Tao, Yu; Ersahin, Mustafa Evren; Zhou, Zhongbo; Gimenez, Juan B; Spanjers, Henri; van Lier, Jules B

    2015-10-15

    The objective of this study was to assess the operational feasibility of an anaerobic membrane bioreactor (AnMBR), consisting of an upflow anaerobic sludge blanket (UASB) reactor coupled to an ultrafiltration membrane unit, at two operational temperatures (25°C and 15°C) for the treatment of municipal wastewater. The results showed that membrane fouling at 15°C was more severe than that at 25°C. Higher chemical oxygen demand (COD) and soluble microbial products (SMP) concentrations, lower mean particle diameter, and higher turbidity in the UASB effluent at lower temperature aggravated membrane fouling compared to the 25°C operation. However, the overall AnMBR treatment performance was not significantly affected by temperature, which was attributed to the physical membrane barrier. Cake resistance was found responsible for over 40% of the total fouling in both cases. However, an increase was observed in the contribution of pore blocking resistance at 15°C related to the larger amount of fine particles in the UASB effluent compared to 25°C. Based on the overall results, it is concluded that an AnMBR, consisting of a UASB coupled membrane unit, is not found technically feasible for the treatment of municipal wastewater at 15°C, considering the rapid deterioration of the filtration performance. PMID:26141423

  1. The scaling of economic and performance parameters of DT and advanced fuel fusion reactors

    International Nuclear Information System (INIS)

    In this study, the plasma stability index beta and the fusion power density in the plasma were treated as independent variables to determine how they influenced three economic performance parameters of fusion reactors burning the DT and four advanced fusion fuel cycles. The economic/performance parameters included the total power produced per unit length of reactor; the mass per unit length, and the specific mass in kilograms/kilowatt. The scaling of these parameters with beta and fusion power density was examined for a common set of engineering assumptions on the allowable wall loading limits, the maximum magnetic field existing in the plasma, average blanket mass density, etc. It was found that the power per unit length decreased as the plasma power density and beta increased. This is a consequence of the fact that the first wall is a bottleneck in the energy flow from the plasma to the generating equipment, and the wall power flux will exceed wall loading limits if the plasma radius exceeds a critical value. If one wished to build an engineering test reactor which produced a burning plasma at the lowest possible initial cost, and without regard to whether such a reactor would ultimately produce the cheapest power, then one would minimize the mass per unit length. The mass per unit length decreases with increasing plasma power density and beta, with the DT reaction being the most expensive at a fixed plasma power density (because of its thicker blanket), and the least expensive at a fixed value of beta, at least up to values of beta of 50%. The specific mass, in kg/kw, which is a rough measure of the cost of the power generated by the reactor, shows an opposite trend. It increases with increasing plasma power density and beta. At a given plasma power density and low beta, the DT reaction gives the lowest specific mass, but at a fixed beta above 10%, the advanced fuel cycles have the lowest specific mass

  2. Performance appraisal and advancement exercise 2005

    CERN Multimedia

    Human Resources Department

    2004-01-01

    The 2005 performance appraisal and advancement exercise will start in the usual way with annual interviews between staff and their supervisors. These interviews should be held in the period from 15 November 2004 to 15 February 2005 [see Administrative Circular No 26 (Rev. 5) Annex II, § 3. For this exercise, the paper MAPS form has been replaced by an electronic form in the EDH system, which is available via EDH (on the EDH desktop under Other Tasks / HR & Training) No changes have been made to the contents of the form. HR Department will shortly provide further information on this subject. Human Resources Department Tel. 73566

  3. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

  4. Advisory group meeting on design and performance of reactor and subcritical blanket systems with lead and lead-bismuth as coolant and/or target material. Working material

    International Nuclear Information System (INIS)

    The purpose of the IAEA Advisory Group Meeting (AGM) on Design and Performance of Reactor and Sub-critical Blanket Systems with Lead and Lead-Bismuth as Coolant and/or Target Material was to provide a forum for international information exchange on all the topics relevant to Pb and Pb/Bi cooled critical and sub-critical reactors. In addition, the AGM aimed at: (1) finding ways and means to improve international co-ordination efforts in this area; (2) obtaining advice from the Member States with regard to the activities to be implemented in this area by the IAEA, in order to best meet their needs; and (3) laying out the plans for an effective co-ordination and support of the R and D activities in this area. The AGM stressed that nuclear energy is a realistic solution to satisfy the energy demand, considering the limited resources of fossil fuel, its uneven distribution in the world and the impact of its use on the planet, and taking into account the expected doubling of the world population in the 21st century and tripling of the electricity demand (especially in the developing countries). However, the AGM concluded that the development of an innovative nuclear technology meeting the following requirements must be pursued: (a) deterministic exclusion of any severe accident; (b) proliferation resistance; (c) cost competitiveness with alternative energy sources; (d) sustainable fuel supply; and (e) solution of the radioactive waste management problem

  5. Impact of ozone pre-treatment on the performance of upflow anaerobic sludge blanket treating pre-treated grain distillery wastewater.

    Science.gov (United States)

    Robertson, L; Britz, T J; Sigge, G O

    2014-01-01

    Two 2 L laboratory-scale upflow anaerobic sludge blanket (UASB) reactors were operated for 277 days. The substrate of the control reactor (Rc) contained grain distillery wastewater (GDWW) that had undergone coagulant pre-treatment, and the substrate of the second UASB reactor consisted of GDWW that had undergone coagulant pre-treatment and ozone pre-treatment (Ro). Both reactors treated pre-treated GDWW successfully at ca. 9 kgCOD m(-3) d(-1). Chemical oxygen demand (COD) reductions of ca. 96% for Rc and 93% for Ro were achieved. Fats, oils and grease (FOG) reductions (%) showed variations throughout the study, and reductions of ca. 88 and 92% were achieved for Rc and Ro, respectively. Rc produced more biogas, and the methane percentage was similar in both reactors. UASB granule washout in Rc suggested possible toxicity of unsaturated fatty acids present in non-ozonated substrate. The feasibility of FOG removal was demonstrated as both reactors successfully treated pre-treated GDWW. Better results were obtained for Ro effluent during post-ozonation. The ozone pre-treatment possibly led to easier degradable wastewater, and better results could potentially be obtained when other post-treatment steps are applied. Ozone pre-treatment did not, however, show an added benefit in the reactor performance results. PMID:25429461

  6. Performance appraisal and advancement exercise 2006 (Rev)

    CERN Multimedia

    Human Resources Department

    2005-01-01

    The 2006 performance appraisal and advancement exercise will start in the usual way with annual interviews between staff and their supervisors. This year, these interviews should be held in the period from 1 December 2005 to 15 March 2006*. In this connection, a number of improvements to the procedures relating to the interviews and performance appraisals are currently under study. Administrative Circular No. 26 (Procedures Governing the Career Development of Staff Members) and the electronic MAPS form in EDH are being reviewed and will be available from December onwards. In the meantime supervisors can start the interview procedure. HR Department will shortly provide further information on this subject. The 2005 MAPS report can be retrieved for consultation at any time via EDH. * Instead of 31 March, as indicated in the electronic Bulletin No. 47 of 21 November 2005 Human Resources Department Tel. 73566

  7. Performance appraisal and advancement exercise 2006

    CERN Multimedia

    Human Resources Department

    2005-01-01

    The 2006 performance appraisal and advancement exercise will start in the usual way with annual interviews between staff and their supervisors. This year, these interviews should be held in the period from 1 December 2005 to 31 March 2006. In this connection, a number of improvements to the procedures relating to the interviews and performance appraisal are currently under study. Administrative Circular No 26 (Procedures governing the Career Development of Staff Members) and the electronic MAPS form in EDH are being reviewed and will be available from December onwards. In the meantime supervisors can start the interview procedure. HR Department will shortly provide further information on this subject. The 2005 MAPS report can be retrieved for consultation at any time via EDH. Human Resources Department Tel. 73566

  8. Performance of the Advanced Light Source

    International Nuclear Information System (INIS)

    The Advanced Light Source (ALS) at the Lawrence Berkeley Laboratory (LBL) is the first of the lower energy (1--2 GeV) third-generation synchrotron radiation facilities to come into operation. Designed with very small electron beam emittances to operate with long insertion devices producing very high brightness beams of synchrotron radiation in the VUV and soft x-ray regions of the spectrum, these facilities are complementary to the higher energy (6--9 GeV) facilities designed for harder x-radiation. The ALS storage ring began operation in October 1993. In this paper, we will review the operational performance of the ALS, including the effects of the 4.5 m long undulators (period 5 cm), and discuss the overall performance of the facility

  9. Heterogeneous structure effect on molten salt blanket neutronics

    Energy Technology Data Exchange (ETDEWEB)

    Grebyonkin, K.F.; Kandiev, Ya.Z.; Malyshkin, G.N.; Orlov, A.I. [Inst. of Technical Pysics, Chelyabinsk (Russian Federation). Dept. of Physics

    1997-09-01

    The report presents the results of the molten salt blanket neutronics calculations performed for researchers of a facility for accelerator-driven transmutation of long-lived radioactive wastes and plutonium conversion. Heterogeneous structure effect on molten salt blanket neutronics was studied through computation. 4 refs., 1 fig., 1 tab.

  10. Evaluation of adhesion strength of Er2O3 coating layer for an advanced breeding blanket system applied to thermal cycles using nano-scratch method

    International Nuclear Information System (INIS)

    The electrical insulator and hydrogen permeation barrier coatings are important materials to realize the liquid metal and molten-salt typed breeding blanket systems. We found that erbium oxide (Er2O3) is one of the promising materials as the electrical insulator and hydrogen permeation restraint coatings. Establishing the mechanical property evaluation method for these coating is extremely important to certify the durability of coating material in the blanket systems. The adhesion strength property, which is one of the key mechanical properties of coating materials, was investigated using the nano-scratch method. From the results, it was found that the nano-scratch test was able to evaluate the adhesion strength of the Er2O3 coating synthesized by the Metal Organic Chemical Vapor Deposition (MOCVD) process with high reproducibility. Furthermore, the adhesion strength of the Er2O3 coating before and after thermal cycling was evaluated using this method. The adhesion strength after 50 thermal cycles at 700degC was kept around 70% compared with that before thermal cycling. (author)

  11. Integrating advanced facades into high performance buildings

    International Nuclear Information System (INIS)

    Glass is a remarkable material but its functionality is significantly enhanced when it is processed or altered to provide added intrinsic capabilities. The overall performance of glass elements in a building can be further enhanced when they are designed to be part of a complete facade system. Finally the facade system delivers the greatest performance to the building owner and occupants when it becomes an essential element of a fully integrated building design. This presentation examines the growing interest in incorporating advanced glazing elements into more comprehensive facade and building systems in a manner that increases comfort, productivity and amenity for occupants, reduces operating costs for building owners, and contributes to improving the health of the planet by reducing overall energy use and negative environmental impacts. We explore the role of glazing systems in dynamic and responsive facades that provide the following functionality: Enhanced sun protection and cooling load control while improving thermal comfort and providing most of the light needed with daylighting; Enhanced air quality and reduced cooling loads using natural ventilation schemes employing the facade as an active air control element; Reduced operating costs by minimizing lighting, cooling and heating energy use by optimizing the daylighting-thermal tradeoffs; Net positive contributions to the energy balance of the building using integrated photovoltaic systems; Improved indoor environments leading to enhanced occupant health, comfort and performance. In addressing these issues facade system solutions must, of course, respect the constraints of latitude, location, solar orientation, acoustics, earthquake and fire safety, etc. Since climate and occupant needs are dynamic variables, in a high performance building the facade solution have the capacity to respond and adapt to these variable exterior conditions and to changing occupant needs. This responsive performance capability

  12. Advanced-performance macroparticle accelerators: Polyphase railguns

    International Nuclear Information System (INIS)

    The initial naive design of railguns has the drawback of a geometric mismatch between the two-rail system used in such accelerators and a round-bore barrel stressed nonuniformly. The most important drawback however, is the reliance on direct current arcs drastically limiting the performance and life span of the launchers. The paper tries to advance the concept of polyphase alternating-current railguns in which multirail systems achieve a more uniform stress distribution within the barrel and in which the frequent passing of currents through zero--each half of the cycle--creates conditions for very low erosion and diminishes other unwanted effects. Such polyphase, alternating-current, electromagnetic rail accelerators hold the promise of higher velocities, higher efficiency and durability. A power supply described in a companion paper to this conference is capable of supplying the optimal voltage and current wave shapes at the necessary energy levels

  13. Neutronic implications of lead-lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W.R.

    1982-08-01

    Lead-lithium alloys have been proposed for use in several conceptual blanket designs for both inertial and magnetic confinement fusion reactors. In most cases, Pb/sub 83/Li/sub 17/, a eutectic with a melting point of 235/sup 0/C, is the chosen composition. The primary reasons for using Pb/sub 83/Li/sub 17/ instead of Li as the tritium breeding material are the perceived safety advantages, low tritium solubility, and favorable neutronic characteristics. This paper describes the neutronic characteristics of Pb/sub 83/Li/sub 17/ blankets with emphasis on the enhanced neutron leakage through chamber ports and the degradation in blanket performance parameters that occurs as a result of the enhanced leakage.

  14. Lightweight IMM PV Flexible Blanket Assembly

    Science.gov (United States)

    Spence, Brian

    2015-01-01

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

  15. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

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

  16. Development of the Helium Cooled Lithium Lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    Highlights: • The HCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • The new design has been developed with the aim to capitalize on TBM experience in ITER. • A new attachment system for the modules has been proposed. - Abstract: The Helium Cooled Lithium Lead (HCLL) blanket is one of the candidate European blanket concepts selected for the DEMOnstration fusion power plant that should follow ITER. In a fusion power plant, the blanket is one of the key components because of its impact on the plant performance, availability, safety and economics. In 2012, the European Fusion Development Agreement (EFDA) agency issued new specifications for DEMO: this paper describes the work performed to adapt the previous 2007 HCLL-DEMO blanket design to those specifications. A new segmentation has been defined assuming straight surfaces for all blanket modules. Following the Multi Module Segment (MMS) option, all modules are attached to a common back supporting structure which also serves as manifold for Helium and PbLi distribution. A detailed CAD design of the central outboard module has been defined. Thermo-hydraulic and thermo-mechanical analyses on of the First Wall and Breeder Zone have been carried out. For the attachment of the modules to the common backplate, a new solution based on the use of Tie Rods, derived from the design of the corresponding HCLL Test Blanket Module for ITER, has been proposed. This paper also identifies the priorities for further development of the HCLL blanket design.

  17. Fast Breeder Blanket Facility (FBBF). Annual report, January 31, 1976--December 31, 1977

    International Nuclear Information System (INIS)

    The work performed in the reporting period was primarily concerned with the construction of the Fast Breeder Blanket Facility (FBBF), acquisition of experimental equipment, outlining the experimental program, preanalysis of the initial loading configuration and investigation of the safety of the initial loading and advanced loadings. The detailed physical description of the FBBF, operational procedures and controls, radiation shielding and experimental equipment are presented. The ability of the FBBF to simulate the blanket spectrum of a large fast breeder reactor is illustrated by comparison of spectra. The source axial distribution, reaction rate comparisons, breeding of plutonium and gamma-ray energy deposition rates are also discussed. Some of the safety aspects of the initial loading and advanced loadings are described. Experimental capabilities of the facility are outlined

  18. High performance anode for advanced Li batteries

    Energy Technology Data Exchange (ETDEWEB)

    Lake, Carla [Applied Sciences, Inc., Cedarville, OH (United States)

    2015-11-02

    The overall objective of this Phase I SBIR effort was to advance the manufacturing technology for ASI’s Si-CNF high-performance anode by creating a framework for large volume production and utilization of low-cost Si-coated carbon nanofibers (Si-CNF) for the battery industry. This project explores the use of nano-structured silicon which is deposited on a nano-scale carbon filament to achieve the benefits of high cycle life and high charge capacity without the consequent fading of, or failure in the capacity resulting from stress-induced fracturing of the Si particles and de-coupling from the electrode. ASI’s patented coating process distinguishes itself from others, in that it is highly reproducible, readily scalable and results in a Si-CNF composite structure containing 25-30% silicon, with a compositionally graded interface at the Si-CNF interface that significantly improve cycling stability and enhances adhesion of silicon to the carbon fiber support. In Phase I, the team demonstrated the production of the Si-CNF anode material can successfully be transitioned from a static bench-scale reactor into a fluidized bed reactor. In addition, ASI made significant progress in the development of low cost, quick testing methods which can be performed on silicon coated CNFs as a means of quality control. To date, weight change, density, and cycling performance were the key metrics used to validate the high performance anode material. Under this effort, ASI made strides to establish a quality control protocol for the large volume production of Si-CNFs and has identified several key technical thrusts for future work. Using the results of this Phase I effort as a foundation, ASI has defined a path forward to commercialize and deliver high volume and low-cost production of SI-CNF material for anodes in Li-ion batteries.

  19. Performance evaluation of up-flow anaerobic sludge blanket (UASB) reactor for treatment of paper mill wastewater

    Institute of Scientific and Technical Information of China (English)

    M. Mahadevaswamy; B.M. Sadashiva Murthy; A.R. Girijamma

    2004-01-01

    The present study deals with the performance evaluation of the UASB reactor under varied organic loadingrate(OLR) for the treatment of paper mill wastewater. The sludge granulation process started after 120 days from thed) the VSS concentration was 12.86 gVSS/L, which got increased to 38.05 gVSS/L at the end of an OLR 2.1process. Many times the pH observed was between 6.5 and 7.8, which is more favorable for any anaerobicprocess. It is also found that pH within the reactor increases along with the height of reactor. The total maximumremoval at this stage was observed to be 90%.

  20. Recent advances in ECE imaging performance

    International Nuclear Information System (INIS)

    ECE Imaging (ECEI) systems have been installed and are presently operating on the KSTAR, DIII-D, ASDEX-UG, and HT-7 tokamaks. All are inherently 2-D systems, collect-ing second harmonic ECE radiation to form temporally-resolved localized Te images. System resolutions range from 16 × 8 (HT-7 and ASDEX-UG) to 20 × 16 (DIII-D) to 24 × 16 (KSTAR), with a spatial resolution as low as 1.0 cm (vertical) by 0.9 cm (radial), and with video bandwidths up to 400 kHz. Noise and drift performance of ECEI systems installed on KSTAR and DIII-D were significantly improved in 2011 with new zero bias detectors. This higher level of performance has resulted in new physics advances as ECEI is employed to visualize high temperature plasmas from the plasma edge (pedestal region) through the plasma core, with examples presented herein. In addition to these systems, a new expanded view ECEI system has been developed for the EAST tokamak that produces 24 × 16 Te images from a single imaging array and which is currently being commissioned.

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

    International Nuclear Information System (INIS)

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

  2. LMFBR Blanket Physics Project progress report No. 6

    International Nuclear Information System (INIS)

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

  3. LMFBR Blanket Physics Project progress report No. 6

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J. (ed.)

    1975-06-30

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

  4. Fast-core thermal-blanket breeder reactor

    International Nuclear Information System (INIS)

    A preliminary assessment of the performance expected from a specific type of FCTB reactor, consisting of a gas-cooled fast system for the core and natural-uranium light-water thermal system for the blanket is reported. Both the core and the blanket use the 238U-Pu fuel cycle. When all the neutrons leaking out of the core reach the blanket, the blanket-to-core power ratio is estimated to be about 1.3. By reducing its water-to-fuel volume ratio below 1.5, the light water blanket can be designed to have a higher ksub(eff), while maintaining an equilibrium fissile fuel content. Compared with conventional FBRs, having the same power output, the FCTB reactor considered offers the following advantages: a lower fissile fuel content, easier and safer control, no need for Pu separation. (B.G.)

  5. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) project began a new design phase called the Engineering Design Activity (EDA) which started in July 1992. A variety of blanket designs options were analyzed as a part of the U.S. ITER home team blanket option trade-off study (BOTS) which began in May 1993. The options considered were a self-cooled Li/V blanket, a helium cooled Li/V blanket and a water cooled 316 SS nonbreeding shield option. Detailed activation, dose rate and waste disposal rating calculations have been performed for these different ITER blanket design options based on a fluence of 3.0 MWa/m2 and an average neutron wall loading of 2.0 MW/m2. A continuous operation assumption was utilized in the analysis. The results of this work are presented in this conference

  7. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

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

  8. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Aqueous solution blanket using lithium salts such as LiNO3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    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

  11. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

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

  12. ITER reference breeding blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, M. [ENEA, Frascati (Italy); Bianchi, A. [EFET, Ansaldo Ricerche, Genova (Italy); Celentano, G. [ENEA, ERG-FUS, Centro di Frascati, Via Enrico Fermi, 27, P.O. Box 65, I-00044, Frascati (IT)] [and others

    1999-11-01

    The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions. (orig.)

  13. ITER reference breeding blanket design

    International Nuclear Information System (INIS)

    The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions. (orig.)

  14. On the numerical assessment of the thermo-mechanical performances of the DEMO Helium-Cooled Pebble Bed breeding blanket module

    International Nuclear Information System (INIS)

    Highlights: • HCPB blanket module thermo-mechanical behavior has been investigated under normal operation and over-pressurization steady state scenarios. • A theoretical–computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Under normal operation scenario, SDC-IC safety rule relevant to the loss of ductility is not fulfilled in the FW and in the hot spots of SPv. • Under over-pressurization scenario, SDC-IC safety rule relevant to the loss of ductility is not met in the hot spots of lower and upper SPv. - Abstract: Within the framework of the European DEMO Breeder Blanket Programme, a research campaign has been launched by University of Palermo, ENEA-Brasimone and Karlsruhe Institute of Technology to theoretically investigate the thermo-mechanical behavior of the Helium-Cooled Pebble Bed (HCPB) breeding blanket module of the DEMO1 blanket vertical segment, under normal operation and over-pressurization loading scenarios. The research campaign has been carried out following a theoretical–computational approach based on the finite element method (FEM) and adopting a qualified commercial FEM code. A realistic 3D FEM model of the HCPB blanket module central poloidal–radial region has been developed, including one breeder cell in the toroidal direction and all the five cells in the poloidal one. No Breeder Units have been modeled, their presence being simulated by effective thermo-mechanical loads. Two sets of uncoupled steady state thermo-mechanical analyses have been carried out with reference to the investigated loading scenarios. In particular, under normal operation scenario (level A) the module has been supposed to undergo both 8 MPa coolant pressure on its cooling channel walls and thermal deformations due to the flat-top plasma operational state thermal field, while under over-pressurization scenario (level D) it has been assumed to experience 8 MPa coolant pressure on its

  15. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    The blanket of the present invention can keep the temperature of breeding materials within a predetermined range even if the breeding materials are consumed and the amount of heat generated from the breeding materials is reduced, thereby enabling to release tritium stably. That is, a neutron incident amount control means is disposed to the blanket for controlling the amount of neutrons incident to the breeding materials. Alternatively, a material to form hollow layers are disposed to the periphery of the breeding materials. With such constitution, the neutron incident amount control means enables to control the incident amount of neutrons from plasmas to the breeding materials, thereby enabling to suppress the change of the amount of heat generated in the breeding materials. In addition, the hollow layers formed at the periphery of the breeding materials enables selective filling of fluids having different heat transfer characteristics thereby enabling to control heat resistance between the breeding materials and cooling tubes. Accordingly, temperature of the breeding materials can be kept constant even in any of the cases. (I.S.)

  16. Integral approach for neutronics analyses of the European test blanket modules in ITER

    International Nuclear Information System (INIS)

    An advanced integral approach has been implemented for neutronic analyses of the European test blanket modules (TBMs) in ITER. The central element of this approach is the use of the geometry conversion tool McCad for the generation of Monte Carlo analysis models from CAD geometry data. Following this approach, an MCNP model of the test blanket port plug with HCPB and HCLL TBM assemblies, elaborated by the European TBM Consortium of Associates (CA), was generated and integrated into the Alite MCNP model of ITER. Neutronic performance and shielding analyses were conducted on the basis of MCNP-5 calculations for the HCPB and HCLL TBMs and the entire shield system. The results indicated the need for a further optimization of the shield system complemented by a rigorous shutdown dose rate analysis.

  17. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

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

  18. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

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

  19. Blanket comparison and selection study. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1983-10-01

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

  20. Advanced transport systems analysis, modeling, and evaluation of performances

    CERN Document Server

    Janić, Milan

    2014-01-01

    This book provides a systematic analysis, modeling and evaluation of the performance of advanced transport systems. It offers an innovative approach by presenting a multidimensional examination of the performance of advanced transport systems and transport modes, useful for both theoretical and practical purposes. Advanced transport systems for the twenty-first century are characterized by the superiority of one or several of their infrastructural, technical/technological, operational, economic, environmental, social, and policy performances as compared to their conventional counterparts. The advanced transport systems considered include: Bus Rapid Transit (BRT) and Personal Rapid Transit (PRT) systems in urban area(s), electric and fuel cell passenger cars, high speed tilting trains, High Speed Rail (HSR), Trans Rapid Maglev (TRM), Evacuated Tube Transport system (ETT), advanced commercial subsonic and Supersonic Transport Aircraft (STA), conventionally- and Liquid Hydrogen (LH2)-fuelled commercial air trans...

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

    International Nuclear Information System (INIS)

    An 80-K Multi Layer Insulation (MLI) blanket in the interconnect region between magnets DD0019 and DD0027 in the Fermi National Accelerator Laboratory (FNAL) ER string was instrumented with temperature sensors to obtain the steady-state temperature gradient through the blanket after string cooldown. A thermal model of the 80-K blanket assembly was constructed to analyze the steady-state temperature gradient data. Estimates of the heat flux through the 80-K MLI blanket assembly and predicted temperature gradients were calculated. The thermal behavior of the heavy polyethylene terapthalate (PET) cover layers separating the shield and inner blanket and inner and outer blankets was derived empirically from the data. The results of the analysis predict a heat flux of 0.363--0.453 W/m2 based on the 11 sets of data. These flux values are 33--46% below the 80-K MLI blanket heat leak budget of 0.676 W/m2. The effective thermal resistance of the two heavy PET cover layers between the shield and inner blanket was found to be 2.1 times that of a single PET spacer layer, and the effective resistance of the two heavy PET cover layers between the inner blanket and outer blanket was found to be 7 times that of a single PET spacer layer. Based on these results, the 80-K MLI blanket assembly appears to be performing more than adequately to meet the 80-K static IR heat leak budget. However, these results should not be construed as a verification of the 80-K static IR heat leak, since no actual heat leak was measured. The results have been used to improve the empirically based model data in the 80-K MLI blanket thermal model, which has previously not included the effects of heavy PET cover layers on 80-K MLI blanket thermal performance

  2. Advances in Nb3Sn Performance

    Energy Technology Data Exchange (ETDEWEB)

    Godeke, Arno

    2008-05-19

    Nb{sub 3}Sn wires with non-Cu critical current densities (J{sub c}) that surpass 3 kAmm{sup -2} at 12 T and 4.2 K are commercially available in piece lengths longer than 10 km. Accelerator-type magnets that utilize these conductors have achieved record magnetic fields. This article summarizes key developments in the last decade that have led to these significant improvements in the performance of Nb{sub 3}Sn wires.

  3. Advances in Nb3Sn Performance

    International Nuclear Information System (INIS)

    Nb3Sn wires with non-Cu critical current densities (Jc) that surpass 3 kAmm-2 at 12 T and 4.2 K are commercially available in piece lengths longer than 10 km. Accelerator-type magnets that utilize these conductors have achieved record magnetic fields. This article summarizes key developments in the last decade that have led to these significant improvements in the performance of Nb3Sn wires.

  4. European reference design of the water-cooled lithium-lead blanket for a demonstration reactor

    International Nuclear Information System (INIS)

    The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European Union for DEMO-relevant design and R ampersand D activities. This paper gives a presentation of the reference conceptual design for water-cooled Pb-17Li DEMO blankets and an overview on the results of its performance assessments. Moreover, a critical discussion about the technical aspects requiring further improvements and/or modifications is performed taking into account the present status of the associated R ampersand D. This concept appears to be a very promising candidate for a DEMO reactor breeding blanket

  5. Preliminary Safety Analysis of Korea HCSB Test Blanket Module

    International Nuclear Information System (INIS)

    A Helium Cooled Solid Breeder (HCSB) blanket has been considered as one of the promising blanket for the fusion power demonstration plant. Therefore HCSB Test Blanket Module (TBM) testing in ITER is the most important milestone for the development of the blanket of the DEMO plant. Korea has developed the HCSB TBM with some features such as graphite reflector and simplified flow passage. The objective of this study was to evaluate the thermal and structural integrity of the HCSB TBM under the hypothetical accidental conditions such as cooling pipe break in TBM. The safety analysis was performed under conservative conditions based on the TBM design, which can be assumed by the similarity of the safety analysis of the ITER shielding blanket. Transient analysis model was used to calculate the temperature distribution for Loss of Coolant Accident (LOCA). Simplified analysis conditions were a) simultaneous plasma shutdown and LOCA b) LOCA and then after FW temperature reaches 1150 deg. plasma shutdown. Helium circuit behavior during the different LOCA scenarios was also evaluated. Finally the design modifications based on the analysis result and the related R-and-D of the HCSB blanket design for the application in a DEMO reactor were mentioned. (author)

  6. A European proposal for an ITER water-cooled solid breeder blanket

    International Nuclear Information System (INIS)

    The water-cooled solid breeder blanket concept proposed here aims to replace the shielding blanket for the enhanced performance phase of the international thermonuclear experimental reactor (ITER). The nominal performances are as follows: an average neutron wall load of 1 MW m-2 which corresponds to a fusion power of about 1.5 GW, and an average neutron fluence of 1 MWy m-2. The proposed blanket concept has been designed to accept a power increase of about 30% and power transients up to 3-5 GW for a short time. This blanket concept is based on a breeder inside tube (BIT)-type blanket with poloidal breeding elements made of 316 L-type stainless steel and filled with lithium metazirconate and beryllium pebbles. Inlet and outlet water temperatures of 160 and 200 C have been considered with a medium-pressure cooling system during plasma burn. The diameters of the breeding elements are compatible with the space available in test fission reactor core channels, making in-pile testing, required for blanket development and qualification, easier. A conservative approach using qualified materials, a blanket concept easily testable in fission reactors and on-going mock-up testing, which can be qualified using the blanket test module during the basic performance phase of ITER, will allow the blanket reliability required for the enhanced performance phase to be achieved. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    International Nuclear Information System (INIS)

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

  10. Thermomechanics analysis and optimization for high power density blanket

    International Nuclear Information System (INIS)

    Thermomechanics analysis, i.e. steady thermal analysis and steady thermal stress analysis have been carried out for a high power density blanket. The Fusion Experimental Breeder (FEB) is adopted as the reference reactor. The parts for the blanket module in Pro/ENGINEER were created, then turn to Pro/MECHANICA functionality for thermomechanics analysis. During analysis, the distribution of the power density in the blanket was optimized to be more flat, the arched curvature and rounds of the cooling tube panels were optimized to less stiffness, and the boundary condition at the interface of helium cooling tube panel and manifold chamber was optimized, which is reasonable by using advanced welding processes with electron beam or laser beam in a single pass. To the end, a maximum temperature Tm 350 degree C and a maximum shear stress τm 80 MPa for the helium cooling panels have been shown in the calculations. (authors)

  11. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

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

  12. A Study on Performance Requirements for Advanced Alarm System

    International Nuclear Information System (INIS)

    A design goals of advanced alarm system is providing advanced alarm information to operator in main control room. To achive this, we applied computer based system to Alarm System. Because, It should apply data management and advanced alarm processing(ie. Data Base Mangegment System and S/W module for alarm processing). These are not impossible in analog based alarm system. And, preexitance research examples are made on digital computer. We have digital systems for test of advanced alarm system table and have tested and studied using by test equipment in the view point of the system performance, stability and security. In this paper, we discribed about general software architecture of preexitance research examples. Also, CPU performance and requirements of system software that served to accommodate it, stability and security

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  14. Performance of Uplink Carrier Aggregation in LTE-Advanced Systems

    DEFF Research Database (Denmark)

    Wang, Hua; Rosa, Claudio; Pedersen, Klaus

    2010-01-01

    paper, we evaluate the performance of uplink CA in LTE-Advanced systems with different CC allocation schemes. We first present the radio resource management (RRM) framework of multi-component carrier LTE-Advanced systems, with special attention on CC selection, adaptive transmission bandwidth, and......Carrier aggregation (CA) has been proposed to aggregate two or more component carriers (CCs) to support a much wider transmission bandwidth for LTE-Advanced systems. With carrier aggregation, it is possible to schedule a user equipment (UE) on multiple component carriers simultaneously. In this...... traffic conditions....

  15. High performance installation for drill and blast advance Mitholz, Switzerland

    Institute of Scientific and Technical Information of China (English)

    Jost Wenk

    2004-01-01

    The section Mitholz of the L tschberg Alp - Transit tunnel consists basically of 3 drill & blast advances with a cross section of 63 - 69m2 and a total length of 25 km.The high - performance back - up installation in use distinguishes itself by the following substantial innovations:The joint venture SATCO ( STRABAG AG, Rothpletz, Lienhard & Cie. , Walo Bertschinger AG, Vinci Construction,Skanska Europe AB), the c ontractor in charge, is achieving very high rates of advance, thanks to the high - performance back - up installations.- The chosen heading system has a positive effect on the entire construction program. Supplementary work can be done within the planned time schedule - Owing to the excellent performance, the joint venture SATCO is ahead on the construction program by about 700m.- The high rates of advance result for the customer in a positive return on investment - The installed equipment results in a higher safety at the workplace for the workers

  16. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

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

  17. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

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

  18. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

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

  19. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

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

  20. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

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

  1. Fusion reactor blanket: neutronic studies in France

    International Nuclear Information System (INIS)

    The problem of effective tritium regeneration is a crucial issue for the fusion reactor, especially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty analysis. The results of these studies permit us to conclude that it is possible to expect an adequate tritium breeding ratio

  2. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

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

  3. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

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

  4. Performance evaluation of the sulfur-redox-reaction-activated up-flow anaerobic sludge blanket and down-flow hanging sponge anaerobic/anoxic sequencing batch reactor system for municipal sewage treatment.

    Science.gov (United States)

    Hatamoto, Masashi; Ohtsuki, Kota; Maharjan, Namita; Ono, Shinya; Dehama, Kazuya; Sakamoto, Kenichi; Takahashi, Masanobu; Yamaguchi, Takashi

    2016-03-01

    A sulfur-redox-reaction-activated up-flow anaerobic sludge blanket (UASB) and down-flow hanging sponge (DHS) system, combined with an anaerobic/anoxic sequencing batch reactor (A2SBR), has been used for municipal sewage treatment for over 2years. The present system achieved a removal rate of 95±14% for BOD, 74±22% for total nitrogen, and 78±25% for total phosphorus, including low water temperature conditions. Sludge conversion rates during the operational period were 0.016 and 0.218g-VSSg-COD-removed(-1) for the UASB, and DHS, respectively, which are similar to a conventional UASB-DHS system, which is not used of sulfur-redox-reaction, for sewage treatment. Using the sulfur-redox reaction made advanced treatment of municipal wastewater with minimal sludge generation possible, even in winter. Furthermore, the occurrence of a unique phenomenon, known as the anaerobic sulfur oxidation reaction, was confirmed in the UASB reactor under the winter season. PMID:26773951

  5. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    In nuclear fusion DEMO reactor, the blanket is required to provide the tritium breeding ratio (TBR) of more than unity by the neutron induced reaction in lithium in the blanket. To provide the TBR of more than unity is critical issue in the development of the blanket. Also in order to develop the blanket with low activation level, the evaluation of the induced activity with high precision is required by taking into account the sequential reactions induced by secondary charged particles. In order to evaluate these issues experimentally, neutronics experiments have been performed by using DT neutrons at JAERI/FNS. From the results of TBR experiment by using the mockup relevant to the DEMO blanket with multilayered structure composed of Be, Li2TiO3 and F82H, it was clarified that the TPR can be evaluated within 10 % uncertainty by using the Monte Carlo calculation. From the results of sequential reactions experiment for the test specimens simulating the cooling water pipe, it was found that the effective cross-sections due to the sequential reactions were increased in a form close to an exponential curve in the cooling water pipe with reducing the distance to the water. (author)

  6. Comparison analysis of fusion breeder blanket concepts

    International Nuclear Information System (INIS)

    Based on the wide survey, the development status and key issues of fusion breeder blanket concepts are summarized. Two types of blanket concepts, i.e. solid and liquid breeder blanket, were compared and assessed in terms of engineering feasibility, tritium recovery and control, economic and safety aspects, etc. The advantages and disadvantages of the two types of blanket concepts were clarified from the viewpoint of technology realization and development potential. This study may act as a valuable reference for fusion blanket concept selection and design. (authors)

  7. The Advanced Photon Source: Performance and results from early operation

    International Nuclear Information System (INIS)

    The Advanced Photon Source at Argonne National Laboratory is now providing researchers with extreme-brilliance undulator radiation in the hard x-ray region of the spectrum. All technical facilities and components are operational and have met design specifications. Fourteen research teams, occupying 20 sectors on the APS experiment hall floor, are currently installing beamline instrumentation or actively taking data. An overview is presented for the first operational years of the Advanced Photon Source. Emphasis is on the performance of accelerators and insertion devices, as well as early scientific results and future plans

  8. Initial performance of Los Alamos Advanced Free Electron Laser

    International Nuclear Information System (INIS)

    We report recent results on the high-brightness electron linac and initial performance of the Advanced Free-electron Laser (AFEL) at Los Alamos. The design and construction of the AFEL beamline are based upon integration of advanced technologies such as the high-brightness rf linac, a brightness-preserving beamline with permanent magnet components, and a pulsed electromagnet microwiggler. With a compact optical resonator, the AFEL will be the first of its kind small enough to be mounted on an optical table, yet capable of providing high-power optical output scanning the near-IR and visible regions. (Author) 3 refs., 3 figs

  9. Performance validation of advanced safety features of APR

    International Nuclear Information System (INIS)

    This paper briefly introduces recent progress, which is mainly being performed at KAERI, in design and performance validation tests of advanced safety features for the APR+ (Advanced Power Reactor plus). The main safety technologies included in the APR+ standard design improvement are an advanced Direct Vessel Injection Plus (DVI+), Fluidic Device Plus (FD+), and Passive Auxiliary Feedwater System (PAFS). The validation tests of advanced safety features are performed using scaled down model testes. They include: 1/5-Scale air-water test, DVI+ duct vibration test, full scale duct injection test, 1/240-Scale PAFS separate effect test, and ATLAS-PAFS integral effect test. The other design verification tests are also performed. They include: Full scale 3-Fuel assembly cross flow test, 1/5-Scale reactor flow distribution test (to be performed, under construction), Core simulator calibration developing test for core flow test, Full scale advanced Fluidic Device (FD+), CEDM test-to be performed, and IVR and core capture test. The core flow and reactor flow tests are being performed by both experimental and CFD (Computational Fluid Dynamics) analysis for the test facility development. The ECC bypass test is performed using 1/5-scale air-water test facility. The duct test facility. The 1/5-scale core barred vibration test with DVI duct is also performed under the both wet and dry condition. The design goal of PAFS is replacing an auxiliary feedwater system with a passive Steam Generator (SG) secondary cooling system of APR+. In PAFS, the core decay heat is ideogram in the Passive Condensation Cooling Tank (PCCT). The PCCT is located on the top of Auxiliary building. To validate the cooling performance of PAFS, a separate effect test loop has been constructed at KAERI (Korea Atomic Energy Research Institute), which is named PASCAL (PAFS Condensing heat removal Assessment Loop). To quantify the tube heat transfer coefficient of the horizontal heat exchanger, a full scale

  10. Resonance self-shielding in the blanket of a hybrid reactor

    International Nuclear Information System (INIS)

    Three sets of energy group cross sections were obtained using various approximations for resonance self shielding. The three models used in obtaining the cross sections were: (a) infinitely dilute model, (b) homogeneous-medium resonance self shielding, and (c) heterogeneous-medium resonance self shielding. The effects on the blanket performance of fusion--fission hybrid reactors, and in particular, on the performance of the current reference Westinghouse Demonstration Tokamak Hybrid Reactor blanket, were compared and analyzed for a variety of fuel-coolant combinations. It has been concluded that (1) the infinitely dilute cross sections can be used to produce preliminary crude estimates for beginning-of-life (BOL) only, (2) the resonance absorber finite dilution should be considered for BOL, poorly moderated blankets and well moderated blankets with low fissile material content situations, and (3) the spacial details should be considered in high fissile content, well moderated blanket situations

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

    International Nuclear Information System (INIS)

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

  12. A Consistent Comparative Study of Advanced Sodium-cooled Fast Burner Cores loaded with Thorium and Uranium-based Metallic Fuels

    International Nuclear Information System (INIS)

    We considered uranium-based metallic fuel of TRU-U-10Zr for driver fuel and thorium was considered as blanket because thorium blanket produces less amount of TRU than uranium blanket and use of thorium blanket leads to smaller sodium void worth than the use of uranium blanket due to the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. However, these cores using thorium blanket still have a large amount of TRU production from the driver fuels because the driver fuels contain a large amount of depleted uranium which leads to the production of TRU through neutron capture. The objective of this work is to consistently compare the neutronic performances of advanced sodium cooled fast reactor cores loaded with thorium and uraniumbased metallic fuels as driver fuel for TRU burning. Our main emphasis is given on the analyses of the differences in the core performance parameters. For consistent comparison, we used the same core configuration and all the same design parameters except for the fact that depleted uranium in uraniumbased fuel is replaced with thorium. We considered the cores having no thorium blanket and the cores having thorium blanket that were designed in our previous works

  13. 2000 GUIDELINES FOR ADVANCEMENT AND EXCEPTIONAL PERFORMANCE AWARDS

    CERN Multimedia

    2000-01-01

    1. GeneralAs indicated in last year's guidelines, studies on CERN's career structures are continuing in the context of the work of the Tripartite Forum on Employment Conditions (TREF) and should be completed during this year's 5-Yearly Remuneration Review for the application of any changes in 2001. In the meantime, the 2000 annual advancement exercise will be conducted on similar basic lines to those of previous years, i.e. within the procedures of the Administrative Circular N° 26 (Rev. 2), thus promoting the aspect of individual merit in advancement.The present guidelines continue to aim at the best possible degree of flexibility of application for each division within the present Staff Rules and Regulations in order to take full account of both the different levels of individual performance and also the different staff profiles.All advancement must be justified on the grounds of correctly judged merit: exceptional advancement (i.e. second step or entering the exceptional advancement grade) for par...

  14. Performance and Weight Estimates for an Advanced Open Rotor Engine

    Science.gov (United States)

    Hendricks, Eric S.; Tong, Michael T.

    2012-01-01

    NASA s Environmentally Responsible Aviation Project and Subsonic Fixed Wing Project are focused on developing concepts and technologies which may enable dramatic reductions to the environmental impact of future generation subsonic aircraft. The open rotor concept (also historically referred to an unducted fan or advanced turboprop) may allow for the achievement of this objective by reducing engine fuel consumption. To evaluate the potential impact of open rotor engines, cycle modeling and engine weight estimation capabilities have been developed. The initial development of the cycle modeling capabilities in the Numerical Propulsion System Simulation (NPSS) tool was presented in a previous paper. Following that initial development, further advancements have been made to the cycle modeling and weight estimation capabilities for open rotor engines and are presented in this paper. The developed modeling capabilities are used to predict the performance of an advanced open rotor concept using modern counter-rotating propeller designs. Finally, performance and weight estimates for this engine are presented and compared to results from a previous NASA study of advanced geared and direct-drive turbofans.

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

    International Nuclear Information System (INIS)

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

  16. Performance of LTE ADVANCED Uplink in a Flat Rayleigh Channel

    Directory of Open Access Journals (Sweden)

    Edward Kasem

    2013-01-01

    Full Text Available This paper describes the performance of LTE advanced uplink transmission in a flat Rayleigh channel. The uplink is simulated using a modified version of the Vienna uplink link level matlab code simulator. This modified version supports two transmission antennas instead of one. Moreover, it includes two extra processes; layer mapping and precoding. In addition, the demodulation reference signal is presented and employed to allow channel estimation. In this paper, the structure of the LTE advanced system is described. Furthermore, we present generation of the demodulation reference signal. Four combinations of two distinct channel estimation and two signal detection methods are used to provide the simulation results of performance evaluation in term of the BER and throughput curves for selected scenarios.

  17. Maximizing Cochlear Implant Patients’ Performance with Advanced Speech Training Procedures

    OpenAIRE

    Fu, Qian-Jie; Galvin, John J.

    2007-01-01

    Advances in implant technology and speech processing have provided great benefit to many cochlear implant patients. However, some patients receive little benefit from the latest technology, even after many years’ experience with the device. Moreover, even the best cochlear implant performers have great difficulty understanding speech in background noise, and music perception and appreciation remain major challenges. Recent studies have shown that targeted auditory training can significantly i...

  18. Performance of a transmutation advanced device for sustainable energy application

    OpenAIRE

    García, C.; Rosales, J.; García, L.; Pérez-Navarro, A.; Escrivá, A.; Abánades Velasco, Alberto

    2011-01-01

    Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas-cooled pebble bed accelerator driven system, TADSEA (Transmutation Advanced Device for Sustainable Energy Application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, cooled by helium which enables high temperatures (...

  19. Conceptual design of the dual-coolant blanket within the framework of the EU Power Plant Conceptual Study (TW2-TRP-PPCS12). Final report

    International Nuclear Information System (INIS)

    The dual-coolant (DC) blanket - characterized by its simple construction, simple function, and high thermal efficiency - is one of the EU advanced blanket concepts to be investigated within the framework of the power plant conceptual study (PPCS). Its basic concept is based on the use of helium-cooled ferritic steel structure, a self-cooled Pb-17Li breeding zone, and SiCf/SiC flow channel inserts serving as electrical and thermal insulators. The present PPCS stage III is largely based on the preparatory study on fusion plant availability (PPA) carried out in 1999 with the objective to perform the conceptual design of the DC blanket (model C). Some details are to be selected in accordance with the overall strategy, which allows an extrapolation of the present knowledge between the near-term blanket concept solutions: water-cooled lead-lithium (WCLL, model A), helium-cooled pebble bed (HCPB, model B), and the very advanced concept of self-cooled Pb-17Li SiCf/SiC (SCLL, model D). In the PPCS the electrical reactor power is normalized to a typical value of commercial reactors of 1,500 MW, which requires iterative calculations between the blanket layout and the system code analysis. This work is coordinated by FZK in co-operation with CEA, EFET/IBERTEF, UKAEA, VR, and VTT Processes. In this report, the present state of development of model C shall be described; the final results of the study shall be summarized and discussed. (orig.)

  20. Advanced fuel technology and performance: Current status and trends

    International Nuclear Information System (INIS)

    During the last years the Nuclear Fuel Cycle and Waste Management Division of the IAEA has been giving great attention to the collection, analysis and exchange of information in the field of reactor fuel technology. Most of these activities are being conducted in the framework of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). The purpose of this Advisory Group Meeting on Advanced Fuel Technology and Performance was to update and to continue the previous work, and to review the experience of advanced fuel technology, its performance with regard to all types of reactors and to outline the future trends on the basis of national experience and discussions during the meeting. As a result of the meeting a Summary Report was prepared which reflected the status of the advanced nuclear fuel technology up to 1990. The 10 papers presented by participants of this meeting are also published here. A separate abstract was prepared for each of these papers. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

    Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li4SiO4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.)

  2. Initial performance of Los Alamos Advanced Free Electron Laser

    International Nuclear Information System (INIS)

    We report recent results on the high-brightness electron linac and initial performance of the Advanced FEL at Los Alamos. The design and construction of the Advanced FEL beamline are based upon integration of advanced technologies such as high-brightness photoinjector, high-gradient compact linac, and permanent-magnet beamline components. With the use of microwiggler, both permanent magnet and pulsed electromagnet, and compact optical resonator, the Advanced FEL will be the first of its kind small enough to be mounted on an optical table and yet capable of providing highpower optical output spanning the near-ir and visible regions. A schematic of the Advanced FEL is shown in. The source of high-current electron pulses is a laser-gated photoelectron injector which forms-an integral part of a high-gradient 1.2-m long rf linear accelerator. The latter is capable of accelerating electrons up to 20 MeV with room temperature operation and 25 MeV at 77K. The electrons are produced in 10-ps pulses with peak currents as high as 300 A. These electron pulses are transported in a brightness-preserving beamline consisting of permanent magnet dipoles and quadrupoles. The beamline has three 30 degrees bends. The first bend allows for the photocathode drive laser input; the second allows for the FEL output and the third turns the electron beam into the floor for safety reasons. Additional information on the design physics of the Advanced FEL can be found elsewhere

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  4. Activation analyses for the different options considered in the US ITER blanket trade-off study

    International Nuclear Information System (INIS)

    Detailed activation analyses were performed for the different blanket design options considered in the ITER blanket option trade-off study. The options considered included a self-cooled Li/V option, a helium-cooled Li/V option and a water-cooled 316 SS non-breeding shield option. A vacuum vessel made of double-wall Inconel 625 and water-cooled 316 SS balls was used with all options. The He-cooled blanket activity is higher than that of the self-cooled blanket due to the larger structure content. Meanwhile, the vacuum vessel activity is lower for the He-cooled blanket option due to the larger neutron attenuation in the blanket. The shield activity and decay heat of the 316 SS/H2O option are higher than those of the Li/V blankets due to the large amount (80%) of 316 SS used. In both Li/V options the blanket qualifies as class C low-level waste. On the other hand, the 316 SS/H2O shield does not qualify for disposal as low- level waste. The 316 SS/H2O option produces the highest off-site doses in the case of accidental release of 100% of its radioactive inventory. Only remote maintenance would be allowed for all options. (orig.)

  5. The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code

    International Nuclear Information System (INIS)

    Highlights: ► New, three-dimensional, parallel, multi-physics code to simulate fuel behavior in nominal operation. ► Fully-coupled thermomechanics for nominal operation and operation during transients. ► Isotopic depletion using Scale/ORIGEN-S within a fuel performance code. ► Leveraging of existing, validated material models from existing fuel performance codes. ► Initial validation evaluation of an advanced modeling and simulation code for fuel performance. - Abstract: The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code is a new, three-dimensional, multi-physics tool that uses state-of-the-art solution methods and validated nuclear fuel models to simulate the nominal operation and anticipated operational transients of nuclear fuel. The AMP Nuclear Fuel Performance code leverages existing validated material models from traditional fuel performance codes and the Scale/ORIGEN-S spent-fuel characterization code to provide an initial capability that is shown to be sufficiently accurate for a single benchmark problem and anticipated to be accurate for a broad range of problems. The thermomechanics foundation can be solved in a time-dependent or quasi-static approach with any variation of operator-split or fully-coupled solutions at each time step through interoperable interfaces to leading computational mathematics tools, including PETSc, Trilinos, and SUNDIALS. A baseline validation of the AMP Nuclear Fuel Performance code has been performed through the modeling of an experiment in the Halden Reactor Project (IFA-432) that demonstrates the integrated capability and provides a baseline of the initial accuracy of the software.

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

    International Nuclear Information System (INIS)

    A trade-off study of liquid-metal self-cooled blankets was carried out to define the performance of these blankets with respect to the main functions in a fusion reactor, and to determine the potential to operate at the maximum possible values of the performance parameters. The main purpose is to improve the reactor economics by maximizing the blanket energy multiplication factor, reduce the capital cost of the reactor, and satisfy the design requirements. The main parameters during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the 6Li enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, the impact of different reactor design choices on the performance parameters was analyzed. The effect of the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, the coolant choice for the nonbreeding inboard blanket, and the neutron source distribution were part of the trade-off study. In addition, tritium breeding benchmark calculations were performed to study the impact of the use of different transport codes and nuclear data libraries. The importance and the negative effect of high TBR on the energy multiplication motivated the benchmark calculations

  7. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

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

  8. On-site evaluation of the performance of a full-scale down-flow hanging sponge reactor as a post-treatment process of an up-flow anaerobic sludge blanket reactor for treating sewage in India.

    Science.gov (United States)

    Okubo, Tsutomu; Onodera, Takashi; Uemura, Shigeki; Yamaguchi, Takashi; Ohashi, Akiyoshi; Harada, Hideki

    2015-10-01

    A down-flow hanging sponge (DHS) reactor is a novel, unaerated, aerobic, biofilm reactor that is used to polish effluent received from an up-flow anaerobic sludge blanket (UASB) reactor for treating municipal sewage. A full-scale DHS reactor was constructed for post-treatment of a full-scale UASB reactor at a municipal sewage treatment plant in India. Performance of the DHS reactor was evaluated with respect to organic removal over 1800 days of continuous operation. The UASB+DHS system consistently produced effluent with chemical oxygen demand (COD), biochemical oxygen demand (BOD), and suspended solids (SS) values of 37, 6.0 and 19 mg L(-1), on average, respectively. The sludge yield of the DHS reactor was estimated to be 0.04 kg SS kg(-1) COD removed or 0.12 kg SS kg(-1) BOD removed, which is considerably lower than other aerobic treatment methods that have been employed for polishing UASB effluent. PMID:26188558

  9. Advanced non-destructive methods for an efficient service performance

    International Nuclear Information System (INIS)

    Due to the power generation industry's desire to decrease outage time and extend inspection intervals for highly stressed turbine parts, advanced and reliable Non-destructive methods were developed by Siemens Non-destructive laboratory. Effective outage performance requires the optimized planning of all outage activities as well as modern Non-destructive examination methods, in order to examine the highly stressed components (turbine rotor, casings, valves, generator rotor) reliably and in short periods of access. This paper describes the experience of Siemens Energy with an ultrasonic Phased Array inspection technique for the inspection of radial entry pinned turbine blade roots. The developed inspection technique allows the ultrasonic inspection of steam turbine blades without blade removal. Furthermore advanced Non-destructive examination methods for joint bolts will be described, which offer a significant reduction of outage duration in comparison to conventional inspection techniques. (authors)

  10. Advanced Wet Tantalum Capacitors: Design, Specifications and Performance

    Science.gov (United States)

    Teverovsky, Alexander

    2016-01-01

    Insertion of new types of commercial, high volumetric efficiency wet tantalum capacitors in space systems requires reassessment of the existing quality assurance approaches that have been developed for capacitors manufactured to MIL-PRF-39006 requirements. The specifics of wet electrolytic capacitors is that leakage currents flowing through electrolyte can cause gas generation resulting in building up of internal gas pressure and rupture of the case. The risk associated with excessive leakage currents and increased pressure is greater for high value advanced wet tantalum capacitors, but it has not been properly evaluated yet. This presentation gives a review of specifics of the design, performance, and potential reliability risks associated with advanced wet tantalum capacitors. Problems related to setting adequate requirements for DPA, leakage currents, hermeticity, stability at low and high temperatures, ripple currents for parts operating in vacuum, and random vibration testing are discussed. Recommendations for screening and qualification to reduce risks of failures have been suggested.

  11. Compatibility problems in tritium breeding blankets

    International Nuclear Information System (INIS)

    Compatibility between tritium breeding materials (liquid or solid), neutron multiplier and structural steels is a concern for the choice of a tritium breeding blanket for NET. For solid tritium breeding blanket, it seems that the more severe compatibility problem is due to the interaction of beryllium with steel. As for the water-cooled Pb17Li blanket, the first results obtained in experimental conditions closed to the concept have evidenced lower corrosion rates than those measured in thermal convection loops

  12. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

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

  13. Nuclear analyses of Indian LLCB test blanket system in ITER

    International Nuclear Information System (INIS)

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no. 2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radio-active waste management, equipments maintenance and replacement strategies and nuclear safety. To predict the nuclear behaviour of LLCB test blanket module in ITER, nuclear responses like tritium production, nuclear heating, neutron fluxes and radiation damages are estimated. As a part of ITER machine, LLCB TBS has to follow certain nuclear shielding requirements i.e. shutdown dose rates should not exceed the defined limits in ITER premises (inside bio-shield ∼100 μSv/hr after 12 days cooling and outside bio-shield ∼10 μSv/hr after 1 day cooling). Hence nuclear analyses are performed to assess and optimize the shielding capability of LLCB TBS inside and outside bio-shield. To state the radio-activity level of LLCB TBS components which support the rad-waste and safety assessment, nuclear activation analyses are executed. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.1). The paper describes comprehensive nuclear performance of LLCB TBS in ITER. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J.

    1980-01-01

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

  15. Initial performance of the advanced inventory verification sample system (AVIS)

    International Nuclear Information System (INIS)

    This paper describes the requirements, design and initial performance of the Advanced Inventory Verification Sample System (AVIS) a non-destructive assay (NDA) system to measure small samples of bulk mixed uranium-plutonium oxide (MOX) materials (powders and pellets). The AVIS design has evolved from previously developed conceptual physics and engineering designs for the Inventory Sample Verification System (INVS), a safeguards system for nondestructive assay of small samples. The AVIS is an integrated gamma-neutron system. Jointly designed by the Nuclear Material Control Center (NMCC) and the Los Alamos National Laboratory (LANL), AVIS is intended to meet a performance specification of a total measurement uncertainty of less than 0.5% in the neutron (240Pueffective) measurement. This will allow the AVIS to replace destructive chemical analysis for many samples, with concomitant cost, exposure and waste generation savings for the facility. Data taken to date confirming the performance of the AVIS is presented.

  16. Initial performance of the advanced inventory verification sample system (AVIS)

    Energy Technology Data Exchange (ETDEWEB)

    Marlow, Johnna B [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Rael, Carlos D [Los Alamos National Laboratory

    2009-01-01

    This paper describes the requirements, design and initial performance of the Advanced Inventory Verification Sample System (AVIS) a non-destructive assay (NDA) system to measure small samples of bulk mixed uranium-plutonium oxide (MOX) materials (powders and pellets). The AVIS design has evolved from previously developed conceptual physics and engineering designs for the Inventory Sample Verification System (INVS), a safeguards system for nondestructive assay of small samples. The AVIS is an integrated gamma-neutron system. Jointly designed by the Nuclear Material Control Center (NMCC) and the Los Alamos National Laboratory (LANL), AVIS is intended to meet a performance specification of a total measurement uncertainty of less than 0.5% in the neutron ({sup 240}Pu{sub effective}) measurement. This will allow the AVIS to replace destructive chemical analysis for many samples, with concomitant cost, exposure and waste generation savings for the facility. Data taken to date confirming the performance of the AVIS is presented.

  17. The gas-cooled Li2O moderator/breeder canister blanket for fusion-synfuels

    International Nuclear Information System (INIS)

    A new integrated power and breeding blanket is described. The blanket incorporates features that make it suitable for synthetic fuel production. It is matched to the thermal and electrical requirements of the General Atomic water-splitting process for producing hydrogen. The fusion reaction is the Tandem Mirror Reactor (TMR) using Mirror Advanced Reactor Study (MARS) physics. The canister blanket is a high temperature, pressure balanced, crossflow heat exchanger contained within a low activity, independently cooled, moderate temperature, first wall structural envelope. The canister uses Li2O as the moderator/breeder and helium as the coolant. ''In situ'' tritium control, combined with slip stream processing and self-healing permeation barriers, assures a hydrogen product essentially free of tritium. The blanket is particularly adapted to synfuels production but is equally useful for electricity production or co-generation

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Activation analysis of coolant water in ITER blanket and divertor

    International Nuclear Information System (INIS)

    Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. 16N and 17N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The 16N and 17N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced 17N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 1010 nuclide/cm3 and 1.05 × 1010 nuclide/cm3 of 16N, 8.93 × 106 nuclide/cm3 and 0.33 × 105 nuclide/cm3 of 17N. The decay gamma-rays from 16N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from 17N decay neutrons was much lower than that from fusion neutrons.

  1. LMFBR blanket assembly heat transfer and hydraulic test data evaluation

    International Nuclear Information System (INIS)

    The USA Test Program for characterization of breeder reactor blanket T and H performance is providing a data base for improved confidence in the design tools employed. Pressure drop tests with wire wrapped rod bundles having a 1.08 triangular pitch to diameter ratio and 4 inch (10 cm) wire wrap lead using water, sodium and air have defined a smooth, continuous, single-valued friction factor versus Reynolds number correlation. This eliminates a possible source of flow instability. The rod bundle temperature rise profiles measured in the heat transfer tests using a prototypic blanket rod bundle agrees in magnitude and shape with the predictions of the marching type sub-channel codes currently employed in blanket subchannel analysis. The low flow test data demonstrates increasing buoyancy induced flows in the lower Reynolds number flow regime. This and the remaining test data will supply a base for calibration of the mixing momentum exchange and conduction factors employed in the subchannel analysis codes; which will contribute to the confidence of the blanket design predictions and reduce the uncertainties which are commonly expressed as hot channel/spot factors

  2. The application of advanced rotor (performance) methods for design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Bussel, G.J.W. van [Delft Univ. of Technology, Inst. for Wind Energy, Delft (Netherlands)

    1997-08-01

    The calculation of loads and performance of wind turbine rotors has been a topic for research over the last century. The principles for the calculation of loads on rotor blades with a given specific geometry, as well as the development of optimal shaped rotor blades have been published in the decades that significant aircraft development took place. Nowadays advanced computer codes are used for specific problems regarding modern aircraft, and application to wind turbine rotors has also been performed occasionally. The engineers designing rotor blades for wind turbines still use methods based upon global principles developed in the beginning of the century. The question what to expect in terms of the type of methods to be applied in a design environment for the near future is addressed here. (EG) 14 refs.

  3. Performance of Advanced Light Source particle beam diagnostics

    International Nuclear Information System (INIS)

    The Advanced Light Source (ALS), a third-generation synchrotron radiation facility, is complete. The particle beam diagnostics have been installed and tested. The beam injection systems have been running for two years. We have performance data on beam position monitors, beam intensity monitors, scintillators, beam collimators, a 50 Ω Faraday cup, and broad-band striplines and kickers used in the linac, transport lines, and the booster synchrotron. The single-turn monitoring capability of the booster beam position monitoring system has been particularly useful for studying beam dynamics. Beam diagnostics for the storage ring are being commissioned. In this paper we describe each instrument, show its performance, and outline how the instruments are controlled and their output data displayed

  4. ITER blanket, shield and material data base

    International Nuclear Information System (INIS)

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

  5. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

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

  6. Operation performance and granule characterization of upflow anaerobic sludge blanket (UASB) reactor treating wastewater with starch as the sole carbon source.

    Science.gov (United States)

    Lu, Xueqin; Zhen, Guangyin; Estrada, Adriana Ledezma; Chen, Mo; Ni, Jialing; Hojo, Toshimasa; Kubota, Kengo; Li, Yu-You

    2015-03-01

    Long-term performance of a lab-scale UASB reactor treating starch wastewater was investigated under different hydraulic retention times (HRT). Successful start-up could be achieved after 15days' operation. The optimal HRT was 6h with organic loading rate (OLR) 4g COD/Ld at COD concentration 1000mg/L, attaining 81.1-98.7% total COD removal with methane production rate of 0.33L CH4/g CODremoved. Specific methane activity tests demonstrated that methane formation via H2-CO2 and acetate were the principal degradation pathways. Vertical characterizations revealed that main reactions including starch hydrolysis, acidification and methanogenesis occurred at the lower part of reactor ("main reaction zone"); comparatively, at the up converting acetate into methane predominated ("substrate-shortage zone"). Further reducing HRT to 3h caused volatile fatty acids accumulation, sludge floating and performance deterioration. Sludge floating was ascribed to the excess polysaccharides in extracellular polymeric substances (EPS). More efforts are required to overcome sludge floating-related issues. PMID:25617619

  7. Preliminary Analysis for K-DEMO Water Cooled Breeding Blanket Using MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    the pressure drops in them and transfers the flow rates to each DLL file as inlet flow rate boundary conditions. The methodology was applied for the two blanket module simulation and its performance was successfully validated. The multiple blanket module analysis method will be extended to the entire system analysis following the present work.

  8. Disruption problematics in segmented-blanket concepts

    International Nuclear Information System (INIS)

    In tokamaks, the hostile operating environment originated by plasma disruption events requires that the first-wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence, there is a need to improve the safety features of the segmented-blanket design concepts in order to satisfy the disruption problematics.The present paper describes recent investigations on internal blanket reinforcement systems needed in order to improve the first-wall/blanket/shield structural design for next-step and commercial fusion reactors. Particularly in the context of SEAFP and ITER activities, representative 3-D CAD models of the inboard and outboard blanket regions and the related magnetomechanical simulations are illustrated. (orig.)

  9. Advanced multistage turbine blade aerodynamics, performance, cooling, and heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Fleeter, S.; Lawless, P.B. [Purdue Univ., West Lafayette, IN (United States)

    1995-10-01

    The gas turbine has the potential for power production at the highest possible efficiency. The challenge is to ensure that gas turbines operate at the optimum efficiency so as to use the least fuel and produce minimum emissions. A key component to meeting this challenge is the turbine. Turbine performance, both aerodynamics and heat transfer, is one of the barrier advanced gas turbine development technologies. This is a result of the complex, highly three-dimensional and unsteady flow phenomena in the turbine. Improved turbine aerodynamic performance has been achieved with three-dimensional highly-loaded airfoil designs, accomplished utilizing Euler or Navier-Stokes Computational Fluid Dynamics (CFD) codes. These design codes consider steady flow through isolated blade rows. Thus they do not account for unsteady flow effects. However, unsteady flow effects have a significant impact on performance. Also, CFD codes predict the complete flow field. The experimental verification of these codes has traditionally been accomplished with point data - not corresponding plane field measurements. Thus, although advanced CFD predictions of the highly complex and three-dimensional turbine flow fields are available, corresponding data are not. To improve the design capability for high temperature turbines, a detailed understanding of the highly unsteady and three-dimensional flow through multi-stage turbines is necessary. Thus, unique data are required which quantify the unsteady three-dimensional flow through multi-stage turbine blade rows, including the effect of the film coolant flow. This requires experiments in appropriate research facilities in which complete flow field data, not only point measurements, are obtained and analyzed. Also, as design CFD codes do not account for unsteady flow effects, the next logical challenge and the current thrust in CFD code development is multiple-stage analyses that account for the interactions between neighboring blade rows.

  10. Thermal-hydraulic design and analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and being designed aiming at 50-200 MW fusion power, 30-50% duty time factor, and tritium self-sustained. Three kinds of tritium breeding blanket concepts, including helium-cooled solid blanket, water-cooled solid blanket and liquid metal-cooled liquid blanket, have been considered for CFETR. Compared to ITER test blanket module, the blanket design for CFETR is facing much more challenges due to the compulsive requirements of tritium self-sufficiency, nuclear heat removal and the space limitation for blanket installation. In this paper, a kind of helium cooled solid tritium breeder blanket was designed for CFETR full superconducting tokamak. The thermal-hydraulic designs were carried out based on the blanket structure design and neutronics calculation. The performance evaluation was conducted using ANSYS, and three-dimensional fluid-solid coupled models were modeled for the accuracy results. The results showed that the FW and BU can satisfy the design requirements. (author)

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

    International Nuclear Information System (INIS)

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m2. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ''beryllium-joint'' concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket

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

    International Nuclear Information System (INIS)

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

  13. Development of Human Performance Analysis and Advanced HRA Methodology

    International Nuclear Information System (INIS)

    The purpose of this project is to build a systematic framework that can evaluate the effect of human factors related problems on the safety of nuclear power plants (NPPs) as well as develop a technology that can be used to enhance human performance. The research goal of this project is twofold: (1) the development of a human performance database and a framework to enhance human performance, and (2) the analysis of human error with constructing technical basis for human reliability analysis. There are three kinds of main results of this study. The first result is the development of a human performance database, called OPERA-I/II (Operator Performance and Reliability Analysis, Part I and Part II). In addition, a standard communication protocol was developed based on OPERA to reduce human error caused from communication error in the phase of event diagnosis. Task complexity (TACOM) measure and the methodology of optimizing diagnosis procedures were also finalized during this research phase. The second main result is the development of a software, K-HRA, which is to support the standard HRA method. Finally, an advanced HRA method named as AGAPE-ET was developed by combining methods MDTA (misdiagnosis tree analysis technique) and K-HRA, which can be used to analyze EOC (errors of commission) and EOO (errors of ommission). These research results, such as OPERA-I/II, TACOM, a standard communication protocol, K-HRA and AGAPE-ET methods will be used to improve the quality of HRA and to enhance human performance in nuclear power plants

  14. Biocatalysis conversion of methanol to methane in an upflow anaerobic sludge blanket (UASB) reactor: Long-term performance and inherent deficiencies.

    Science.gov (United States)

    Lu, Xueqin; Zhen, Guangyin; Chen, Mo; Kubota, Kengo; Li, Yu-You

    2015-12-01

    Long-term performance of methanol biocatalysis conversion in a lab-scale UASB reactor was evaluated. Properties of granules were traced to examine the impact of methanol on granulation. Methanolic wastewater could be stably treated during initial 240d with the highest biogas production rate of 18.6 ± 5.7 L/Ld at OLR 48 g-COD/Ld. However, the reactor subsequently showed severe granule disintegration, inducing granule washout and process upsets. Some steps (e.g. increasing influent Ca(2+) concentration, etc.) were taken to prevent rising dispersion, but no clear improvement was observed. Further characterizations in granules revealed that several biotic/abiotic factors all caused the dispersion: (1) depletion of extracellular polymeric substances (EPS) and imbalance of protein/polysaccharide ratio in EPS; (2) restricted formation of hard core and weak Ca-EPS bridge effect due to insufficient calcium supply; and (3) simplification of species with the methanol acclimation. More efforts are required to solve the technical deficiencies observed in methanolic wastewater treatment. PMID:26441026

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

    International Nuclear Information System (INIS)

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

  16. IMPULSE---an advanced, high performance nuclear thermal propulsion system

    International Nuclear Information System (INIS)

    IMPULSE is an advanced nuclear propulsion engine for future space missions based on a novel conical fuel. Fuel assemblies are formed by stacking a series of truncated (U, Zr)C cones with non-fueled lips. Hydrogen flows radially inward between the cones to a central plenum connected to a high performance bell nozzle. The reference IMPULSE engine rated at 75,000 lb thrust and 1800 MWt weighs 1360 kg and is 3.65 meters in height and 81 cm in diameter. Specific impulse is estimated to be 1000 for a 15 minute life at full power. If longer life times are required, the operating temperature can be reduced with a concomitant decrease in specific impulse. Advantages of this concept include: well defined coolant paths without outlet flow restrictions; redundant orificing; very low thermal gradients and hence, thermal stresses, across the fuel elements; and reduced thermal stresses because of the truncated conical shape of the fuel elements

  17. An advanced probabilistic structural analysis method for implicit performance functions

    Science.gov (United States)

    Wu, Y.-T.; Millwater, H. R.; Cruse, T. A.

    1989-01-01

    In probabilistic structural analysis, the performance or response functions usually are implicitly defined and must be solved by numerical analysis methods such as finite element methods. In such cases, the most commonly used probabilistic analysis tool is the mean-based, second-moment method which provides only the first two statistical moments. This paper presents a generalized advanced mean value (AMV) method which is capable of establishing the distributions to provide additional information for reliability design. The method requires slightly more computations than the second-moment method but is highly efficient relative to the other alternative methods. In particular, the examples show that the AMV method can be used to solve problems involving non-monotonic functions that result in truncated distributions.

  18. Initial performance of Los Alamos Advanced Free Electron Laser

    International Nuclear Information System (INIS)

    The Los Alamos compact Advanced FEL has lased at 4.7 and 5.2 μm with a 1-cm period wiggler and a high-brightness electron beam at 16.8 and 15.8 MeV, respectively. The measured electron beam normalized emittance is 1.7 π·mm·mrad at a peak current of 100 A, corresponding to a beam brightness greater than 2 x 1012 A/m2rad2. Initial results indicate that the AFEL small signal gain is ∼8% at 0.3 nC (30 A peak). The maximum output energy is 7 mJ over a 2-μs macropulse. The AFEL performance can be significantly enhanced by improvements in the rf and drive laser stability

  19. Characterization of UMT2013 Performance on Advanced Architectures

    Energy Technology Data Exchange (ETDEWEB)

    Howell, Louis [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-12-31

    This paper presents part of a larger effort to make detailed assessments of several proxy applications on various advanced architectures, with the eventual goal of extending these assessments to codes of programmatic interest running more realistic simulations. The focus here is on UMT2013, a proxy implementation of deterministic transport for unstructured meshes. I present weak and strong MPI scaling results and studies of OpenMP efficiency on the Sequoia BG/Q system at LLNL, with comparison against similar tests on an Intel Sandy Bridge TLCC2 system. The hardware counters on BG/Q provide detailed information on many aspects of on-node performance, while information from the mpiP tool gives insight into the reasons for the differing scaling behavior on these two different architectures. Preliminary tests that exploit NVRAM as extended memory on an Ivy Bridge machine designed for “Big Data” applications are also included.

  20. Advanced Low Conductivity Thermal Barrier Coatings: Performance and Future Directions

    Science.gov (United States)

    Zhu, Dongming; Miller, Robert A.

    2008-01-01

    Thermal barrier coatings will be more aggressively designed to protect gas turbine engine hot-section components in order to meet future engine higher fuel efficiency and lower emission goals. In this presentation, thermal barrier coating development considerations and performance will be emphasized. Advanced thermal barrier coatings have been developed using a multi-component defect clustering approach, and shown to have improved thermal stability and lower conductivity. The coating systems have been demonstrated for high temperature combustor applications. For thermal barrier coatings designed for turbine airfoil applications, further improved erosion and impact resistance are crucial for engine performance and durability. Erosion resistant thermal barrier coatings are being developed, with a current emphasis on the toughness improvements using a combined rare earth- and transition metal-oxide doping approach. The performance of the toughened thermal barrier coatings has been evaluated in burner rig and laser heat-flux rig simulated engine erosion and thermal gradient environments. The results have shown that the coating composition optimizations can effectively improve the erosion and impact resistance of the coating systems, while maintaining low thermal conductivity and cyclic durability. The erosion, impact and high heat-flux damage mechanisms of the thermal barrier coatings will also be described.

  1. Performance of a transmutation advanced device for sustainable energy application

    International Nuclear Information System (INIS)

    Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Science.gov (United States)

    Hillard, G. Barry

    1993-05-01

    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.

  5. Materials issues in the design of the ITER first wall, blanket, and divertor

    International Nuclear Information System (INIS)

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

  6. Materials issues in the design of the ITER first wall, blanket, and divertor

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. [Argonne National Lab., IL (United States); Wu, C.H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team; Koroda, T. [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan); Shatalov, G. [Kurchatov Inst. of Atomic Energy, Moscow (USSR)

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R&D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented.

  7. Materials issues in the design of the ITER first wall, blanket, and divertor

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. (Argonne National Lab., IL (United States)); Wu, C.H. (Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team); Koroda, T. (Japan Atomic Energy Research Inst., Ibaraki-ken (Japan)); Shatalov, G. (Kurchatov Inst. of Atomic Energy, Moscow (USSR))

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented.

  8. Inertial fusion reactors using Compact Fusion Advanced Rankine (CFARII) MHD conversion

    International Nuclear Information System (INIS)

    This study evaluates the potential performance (efficiency and cost) of inertial fusion reactors assumed capable of vaporizing blankets of various working materials to a temperature (10,000-20,000 K) suitable for economical MHD conversion in a Compact Fusion Advanced Rankine II (CFARII) power cycle. Using a conservative model, 1-D neutronics calculations of the fraction of fusion yield captured as a function of the blanket thickness of Flibe, lithium and lead-lithium blankets are used to determine the optimum blanket thickness for each material to minimize CoE for various assumed fusion yields, 'generic' driver costs, and target gains. Lithium-hydride blankets are also evaluated using an extended neutronics model. Generally optimistic ('advanced') combinations of lower driver cost/joule and higher target gain are assumed to allow high enough fusion yields to vaporize and ionize target blankets thick enough to stop most 14 MeV neutrons, and to breed tritium. A novel magnetized, prestressed reactor chamber concept is modeled together with previously developed models for the CFARII Balance-of-Plant (BoP), consisting of a supersonic plasma jet, MHD generator, and 'raindrop' condensor. High fusion yields (20 to 80 GJ) are found necessary to heat and ionize the Flibe, lithium, and lead-lithium blankets for MHD conversion, with initial solid thicknesses sufficient to capture most of the fusion yield. Much smaller fusion yields (1 to 20 GJ) are required for lithium-hybride blankets. For Flibe, lithium, and lead-lithium blankets, improvements in target gain and/or driver cost/joule, characterized by a 'Bang per Buck' figure-of-merit of > or ∼20 joules yield per driver Dollar, would be required for competitive CoE, while a figure-of-merit of > or ∼1 joule yield per driver Dollar would suffice for lithium-hybride blankets. Advances in targets/driver costs would benefit any IFE reactor, but the very low CFARII BoP costs (contributing only 3 mills/kWh for CoE) allows this

  9. Performance of advanced trauma life support procedures in microgravity

    Science.gov (United States)

    Campbell, Mark R.; Billica, Roger D.; Johnston, Smith L 3rd; Muller, Matthew S.

    2002-01-01

    BACKGROUND: Medical operations on the International Space Station will emphasize the stabilization and transport of critically injured personnel and so will need to be capable of advanced trauma life support (ATLS). METHODS: We evaluated the ATLS invasive procedures in the microgravity environment of parabolic flight using a porcine animal model. Included in the procedures evaluated were artificial ventilation, intravenous infusion, laceration closure, tracheostomy, Foley catheter drainage, chest tube insertion, peritoneal lavage, and the use of telemedicine methods for procedural direction. RESULTS: Artificial ventilation was performed and appeared to be unaltered from the 1-G environment. Intravenous infusion, laceration closure, percutaneous dilational tracheostomy, and Foley catheter drainage were achieved without difficulty. Chest tube insertion and drainage were performed with no more difficulty than in the 1-G environment due to the ability to restrain patient, operator and supplies. A Heimlich valve and Sorenson drainage system were both used to provide for chest tube drainage collection with minimal equipment, without the risk of atmospheric contamination, and with the capability to auto-transfuse blood drained from a hemothorax. The use of telemedicine in chest tube insertion was demonstrated to be useful and feasible. Peritoneal lavage using a percutaneous technique, although requiring less training to perform, was found to be dangerous in weightlessness due to the additional pressure of the bowel on the anterior abdominal wall creating a high risk of bowel perforation. CONCLUSIONS: The performance of ATLS procedures in microgravity appears to be feasible with the exception of diagnostic peritoneal lavage. Minor modifications to equipment and techniques are required in microgravity to effect surgical drainage in the presence of altered fluid dynamics, to prevent atmospheric contamination, and to provide for the restraint requirements. A parabolic

  10. Physical investigation for neutron consumption and multiplication in fusion–fission hybrid test blanket module

    International Nuclear Information System (INIS)

    Highlights: • Preliminary design of hybrid test blanket module was done for ITER. • Blanket performance was analyzed for five fuel types with Li and LiPb coolants. • Detailed neutronic analysis was performed by computing neutron production and loss factors. • Inelastic neutron source factor of LiPb caused major change in blanket performance. • TiC reflector improved performance in TRU transmutation and tritium breeding. - Abstract: Inelastic scattering of high energy fusion neutrons does affect the performance of fusion blanket based on the choice of different materials. It will also affect the behavior of source neutrons in a subcritical fusion fission hybrid blanket and consequently the transmutation and tritium breeding performance. A fusion fission hybrid test blanket module (HTBM) is designed which is presumed to be tested in a large sized tokamak and plasma neutron source is similar to ITER. In this preliminary design of HTBM the neutron source and loss factors are computed for the detailed neutronic performance analysis. The neutronic analysis of hybrid blanket module is performed for five different TRU fuel types: TRU-Zr, TRU-Mo, TRU-Oxide, TRU-Carbide and TRU-Nitride. In this module design, it is aimed to burn and transmute the TRU nuclides from high-level radioactive waste of PWR spent fuel. The effect of TiC reflector on transmutation and tritium breeding performance of HTBM is also quantified. MCNPX is used for neutronic computations. Neutron spectrum, capture to fission ratio and waste transmutation ratio of each fuel type are compared to evaluate their waste transmutation performance. Tritium breeding ratio is also compared for two coolant options: Li and LiPb eutectic

  11. Performance modelling for product development of advanced window systems

    Energy Technology Data Exchange (ETDEWEB)

    Appelfeld, D.

    2012-05-15

    The research presented in this doctoral thesis shows how the product development (PD) of Complex Fenestration Systems (CFSs) can be facilitated by computer-based analysis to improve the energy efficiency of fenestration systems as well as to improve the indoor environment. It is complicated to holistically evaluate the performance of a prototyped system, since simulation programs evaluate standardised products such as aluminium venetian blinds. State-of-the-art tools and methods,which can address interrelated performance parameters of CFS, are sought. It is possible to evaluate such systems by measurements, however the high cost and complexity of the measurements are limiting factors. The studies in this thesis confirmed that the results from the performance measurements of CFSs can be interpreted by simulations and hence simulations can be used for the performance analysis of new CFSs. An advanced simulation model must be often developed and needs to be validated by measurements before the model can be reused. The validation of simulations against the measurements proved the reliability of the simulations. The described procedures can be used at the initial stages of the PD to foresee the consequences of the innovation, and aim at the development by an iterative testing to meet the requirements. It was demonstrated that by improving the fenestration system, the overall building energy demand can be reduced by optimizing lighting, heating and cooling. The indoor environment quality can be improved by careful shading strategy and maximizing the use of daylight. The recent developments of the building simulation programmes enabled to perform annual, dynamic and climate based energy evaluation of CFSs. The case study of development of a window frame made of glass fibre reinforced polyester (GFRP) demonstrated that this composite material is suitable for window frames. A window with positive net energy gain (NEG) and a slim window frame was developed, by a combination

  12. Performance evaluation of cognitive radio in advanced metering infrastructure communication

    Science.gov (United States)

    Hiew, Yik-Kuan; Mohd Aripin, Norazizah; Din, Norashidah Md

    2016-03-01

    Smart grid is an intelligent electricity grid system. A reliable two-way communication system is required to transmit both critical and non-critical smart grid data. However, it is difficult to locate a huge chunk of dedicated spectrum for smart grid communications. Hence, cognitive radio based communication is applied. Cognitive radio allows smart grid users to access licensed spectrums opportunistically with the constraint of not causing harmful interference to licensed users. In this paper, a cognitive radio based smart grid communication framework is proposed. Smart grid framework consists of Home Area Network (HAN) and Advanced Metering Infrastructure (AMI), while AMI is made up of Neighborhood Area Network (NAN) and Wide Area Network (WAN). In this paper, the authors only report the findings for AMI communication. AMI is smart grid domain that comprises smart meters, data aggregator unit, and billing center. Meter data are collected by smart meters and transmitted to data aggregator unit by using cognitive 802.11 technique; data aggregator unit then relays the data to billing center using cognitive WiMAX and TV white space. The performance of cognitive radio in AMI communication is investigated using Network Simulator 2. Simulation results show that cognitive radio improves the latency and throughput performances of AMI. Besides, cognitive radio also improves spectrum utilization efficiency of WiMAX band from 5.92% to 9.24% and duty cycle of TV band from 6.6% to 10.77%.

  13. CONCERNING THE ADVANCED SCIENCE IN HIGH PERFORMANCE SPORT

    Directory of Open Access Journals (Sweden)

    Gagea Adrian

    2010-04-01

    Full Text Available The advanced sciences are based on the most recent huge increasing of technology and on interdisciplinary commencement of great interest topics, as top sport is considering. The main problem in top sport seems to be the obtaining high sport’s performance in as short as possible time, having great efficiency and minimum risks.The cell-engineering domain, in which the author of this paper has a modest contribution, is a means of genetic control for human performance, including sport, gene expression, molecular interactions within the cell, intracellular signalling, cell mechanics and motility etc.The domain of Psyche, of controlling feelings and manifestations, is also, on the focus of top sport interest, especially for the reason that, from inside of this domain, is feasible to accede at the biological reserves unavoidable in normal conditions, but avoidable in emergency or surviving situations. The new knowledge about energetic metabolism, about the rotation of ATP molecules, or coming out from scientifically experiments of association of nutrients or of reconsidering the recovery stimulants after effort, are providing, also, very useful information for top sport practitioners.It is not to disregard the contribution of the new information about the human physical limits, biomechanics, tactics of doing and controls the physical effort by means of sensorial biofeedback or theperformance’s advantages coming from new high-minded techniques and materials of sport accessories

  14. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    A near-field performance evaluation of an advanced cold process canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie local-iced corrosion for the steel or copper canisters can be dismissed as a failure mechanism; The evaluation of the effects of processed outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. For completeness also evaluation of post-failure behaviour was carried out. Analyses were focussed on low probability phenomena from faults in canisters. Some items were identified where further research is justified in order to increase knowledge of the phenomena and thus strengthen the confidence of safety margins. However, it can be concluded that the risks of these scenarios can be judged to be acceptable. This is due to the fact that firstly, the probability of occurrence of most of these scenarios can be controlled to a large extent through technical measures. Secondly, these analyses indicated that the consequences would not be severe

  15. UCIMS:Advances in geotechnical construction and performance monitoring

    Institute of Scientific and Technical Information of China (English)

    Raphael Siebenmann a; Hai-Tien Yu b; Robert Bachus a

    2015-01-01

    The Crossrail project currently under construction in Central London has been described as“The Big Dig on Steroids”, obviously referencing the Central Artery/Tunnel project in Boston completed in 2007. To address the multiple demands for timely construction performance monitoring, Crossrail envisioned the underground construction information management system (UCIMS) to monitor construction progress and structural health along the entire route, with a network of geotechnical instruments (i.e. slope in-clinometers, extensometers, piezometers, etc.) and tunnel boring machine (TBM) position information. The UCIMS is a geospatially referenced relational database that was developed using an open source geographic information system (GIS) that allowed all stakeholders near immediate feedback of con-struction performance. The purpose of this article is to provide a brief history of geotechnical and structural monitoring software, to describe the structure and operation of the UCIMS, and to demon-strate how the functionality afforded by this system provided the requisite feedback to the stakeholders. Examples will be given regarding how the data management and visualization concepts incorporated into the UCIMS advanced the geotechnical construction industry.

  16. Performance evaluation of advanced industrial SPECT system with diverging collimator

    International Nuclear Information System (INIS)

    An advanced industrial SPECT system with 12-fold-array diverging collimator was developed for flow visualization in industrial reactors and was discussed in the previous study. The present paper describes performance evaluation of the SPECT system under both static- and dynamic- flow conditions. Under static conditions, the movement of radiotracer inside the test reactor was compared with that of color tracer (blue ink) captured with a high-speed camera. The comparison of the reconstructed images obtained with the radiotracer and the SPECT system showed fairly good agreement with video-frames of the color tracer obtained with the camera. Based on the results of the performance evaluation, it is concluded that the SPECT system is suitable for investigation and visualization of flows in industrial flow reactors. - Highlights: • Industrial SPECT system provides the flow behavior of industrial multiphase processes. • A 12-fold-array industrial SPECT system was constructed using diverging collimators. • The constructed system turned out to be very suitable to examine the fluid behavior

  17. UCIMS: Advances in geotechnical construction and performance monitoring

    Directory of Open Access Journals (Sweden)

    Raphael Siebenmann

    2015-04-01

    Full Text Available The Crossrail project currently under construction in Central London has been described as “The Big Dig on Steroids”, obviously referencing the Central Artery/Tunnel project in Boston completed in 2007. To address the multiple demands for timely construction performance monitoring, Crossrail envisioned the underground construction information management system (UCIMS to monitor construction progress and structural health along the entire route, with a network of geotechnical instruments (i.e. slope inclinometers, extensometers, piezometers, etc. and tunnel boring machine (TBM position information. The UCIMS is a geospatially referenced relational database that was developed using an open source geographic information system (GIS that allowed all stakeholders near immediate feedback of construction performance. The purpose of this article is to provide a brief history of geotechnical and structural monitoring software, to describe the structure and operation of the UCIMS, and to demonstrate how the functionality afforded by this system provided the requisite feedback to the stakeholders. Examples will be given regarding how the data management and visualization concepts incorporated into the UCIMS advanced the geotechnical construction industry.

  18. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    A near-field performance evaluation of an Advanced Cold Process Canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. The canister design was originally proposed by TVO. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. Throughout the analysis, present day underground conditions has been assumed to persist during the service life of the canister. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie localized corrosion for the steel or copper canisters can be dismissed as a failure mechanism. The evaluation of the effects of processes outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. This factor will ensure the safety of the concept. (orig.)

  19. Advanced modeling and simulation to design and manufacture high performance and reliable advanced microelectronics and microsystems.

    Energy Technology Data Exchange (ETDEWEB)

    Nettleship, Ian (University of Pittsburgh, Pittsburgh, PA); Hinklin, Thomas; Holcomb, David Joseph; Tandon, Rajan; Arguello, Jose Guadalupe, Jr. (,; .); Dempsey, James Franklin; Ewsuk, Kevin Gregory; Neilsen, Michael K.; Lanagan, Michael (Pennsylvania State University, University Park, PA)

    2007-07-01

    An interdisciplinary team of scientists and engineers having broad expertise in materials processing and properties, materials characterization, and computational mechanics was assembled to develop science-based modeling/simulation technology to design and reproducibly manufacture high performance and reliable, complex microelectronics and microsystems. The team's efforts focused on defining and developing a science-based infrastructure to enable predictive compaction, sintering, stress, and thermomechanical modeling in ''real systems'', including: (1) developing techniques to and determining materials properties and constitutive behavior required for modeling; (2) developing new, improved/updated models and modeling capabilities, (3) ensuring that models are representative of the physical phenomena being simulated; and (4) assessing existing modeling capabilities to identify advances necessary to facilitate the practical application of Sandia's predictive modeling technology.

  20. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

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

  1. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

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

  2. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Zhongliang; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Chen, Chong; Li, Min; Zhou, Guangming

    2015-06-15

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements.

  3. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    In the frame of the recent CEA studies aiming at the development, evaluation and comparison of solid breeder blanket concepts in view of their adaptation to NET, and the evaluation of specific questions related to the first wall design, the present paper examines first the performances of a helium cooled toroidal blanket design for NET, based on innovative Beryllium/Ceramics breeder rod elements. Neutronic and thermo-mechanical optimisation converges on a concept featured by a breeding capability in excess of 1.2, a reasonnable pumping power of 1 % and a narrow breeder temperature range (470 +- 30 0C of the breeder), the latter being largely independent of the power level. The final section of the paper is devoted to the evaluation of the heat load poloidal distribution and to the irradiation effects on first wall structural materials

  6. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

  7. The advanced scenario analysis for performance assessment of geological disposal

    International Nuclear Information System (INIS)

    First of all, with regard to the FEP information data on the Engineered Barrier System (EBS) developed by JNC, description level and content of the FEPs have been examined from various angles on the basis of the latest research information. Each content of the FEP data has been classified and modified by means of integrating descriptive items, checking detail levels and correlations with other FEPs, collating with the H12 report, and adding technical information after H12 report. Secondly, scenario-modeling process has been studied. The study has been conducted by evaluating representation of the repository system, definition of FEP properties, and process interactions based on the concept of the interaction matrix (RES format) which represents influences between physicochemical characteristics of the repository, followed by an experimental development of the actual RES interaction matrix based on the H12 report as the examination to improve the transparency, traceability and comprehensibility of the scenario analysis process. Lastly, in relation to the geological disposal system, assessment techniques have been examined for more practical scenario analysis on particularly strong perturbations. Possible conceptual models have been proposed for each of these scenarios; seismic, faulting, and dike intrusion. As a result of these researches, a future direction for advanced scenario analysis on performance assessment has been indicated, as well as associated issues to be discussed have been clarified. (author)

  8. Exploratory Study of Blanket Liquid Curtain

    Institute of Scientific and Technical Information of China (English)

    HUGang; HUANGJinhua; FENGKaiming

    2003-01-01

    Blankets and other in-vessel components are easily damaged owing to their circumstance of high radiation and high heat. To protect them, first wall design should be considered. Owing to its high heat removal nd self-refreshing capability, liquid metal first wall has been seen as a potential first wall for a fusion reactor in the future. Blanketliquid curtain is actually a special liquid metal wall to protect blanket.

  9. Adaptation of the HCPB DEMO TBM as breeding blanket for ITER : Neutronic and thermal analyses

    International Nuclear Information System (INIS)

    Two breeding blanket are presently developed in Europe for the DEMO reactor: the first one, the Helium Cooled Lithium Lead (HCLL) uses a liquid breeder while the other , the Helium Cooled Pebble Bed (HCPB), uses a solid breeder in form of pebble bed. The modules of these blankets, called Test Blanket Modules (TBM) will be located in correspondence of the equatorial ports of ITER in order to be tested. ITER FEAT was designed with shielding blankets, therefore in the final stage of the experiment, in the foreseen tritium -deuterium operation phase, the tritium will be supplied to the reactor and not produced inside it. Since the production of tritium is of main importance for the feasibility of a nuclear fusion reactor, perhaps in the ITER final stage, the shielding blanket could be substituted by means of a breeding blanket. The geometry and composition of this breeding blanket would be, of course, similar to that of TBM which demonstrated to have the best performances. This paper illustrates a neutronic and thermal analysis of an hypothetical triziogen blanket for ITER FEAT made similar to a HCPB test module. The main aims of the performed analyses are to determine the Tritium Breeding Ratio (TBR) considering different solid breeders (Li4SiO4 and Li2TiO3) with different enrichment in 6Li and different structural materials (a 9%CRWVTa reduced activation ferritic martensitic steel (EUROFER) or ceramic matrix composites like SiCf/SiC). The breeding blanket design is compared considering the highest value of TBR and the verification of the temperature constraints ( 550 oC for the steel, 950 o C for the breeder and 650 oC for the Beryllium). The neutronic analyses have been performed by means of MCNP-4C code and the thermal analyses using the MSC-MARC code. A TBR about equal 1 was obtained with a SiCf/SiC structural material and a Li4SiO4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give useful

  10. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. High Performance Thrusters for Advanced Green Monopropellants Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The development of an advanced green monopropellant propulsion system could have significant benefits to a wide range of NASA space missions, from deep space...

  13. Electrical connectors for blanket modules in ITER

    International Nuclear Information System (INIS)

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

  14. Activation analysis and waste management for blanket materials of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

    International Nuclear Information System (INIS)

    The preliminary studies of the activation analysis and waste management for blanket materials of the multi-functional experimental fusion–fission hybrid reactor, i.e. Multi-Functional eXperimental Fusion Driven Subcritical system named FDS-MFX, were performed. The neutron flux of the FDS-MFX blanket was calculated using VisualBUS code and Hybrid Evaluated Nuclear Data Library (HENDL) developed by FDS Team. Based on these calculated neutron fluxes, the activation properties of blanket materials were analyzed by the induced radioactivity, the decay heat and the contact dose rate for different regions of the FDS-MFX blanket. The safety and environment assessment of fusion power (SEAFP) strategy, which was developed in Europe, was applied to FDS-MFX blanket for the management of activated materials. Accordingly, the classification and management strategy of activated materials after different cooling time were proposed for FDS-MFX blanket

  15. Advancement of flash hydrogasification: Task VIII. Performance testing

    Energy Technology Data Exchange (ETDEWEB)

    Falk, A.Y.; Schuman, M.D.; Kahn, D.R.

    1986-06-01

    This topical report documents the technical effort required to investigate and verify the reaction chemistry associated with the Rockwell Advanced Flash Hydropyrolysis (AFHP) concept for the production of substitute natural gas (SNG) from coal. The testing phase of the program included 5 preburner performance evaluation tests (14 test conditions) and 11 coal-fed reactor tests (19 test conditions). The reactor test parameters investigated spanned exist temperatures from 1775 to 2050/sup 0/F, residence times from 2 to 8 s, inlet gas-to-coal ratios from 0.15 to 0.27 lb-mole/lb, and inlet-steam-to-H/sub 2/ mole ratios from 0.15 to 0.86. One test was conducted to investigate the effect of CH/sub 4/ addition to the hydrogen feed stream (22 mole % CH/sub 4/), with subsequent partial oxidation of the CH/sub 4/ to CO/sub x/ in the preburner system, on the AFHP reactor chemistry and product gas composition. Overall carbon conversion and total carbon conversion to gases (namely, CH/sub 4/, C/sub 2/H/sub 6/, CO, and CO/sub 2/) ranged from 53 to 68% and 35 to 68%, respectively. The gas produced was primarily CH/sub 4/ (31 to 53% carbon conversion to CH/sub 4/). Carbon conversion to total liquids was strongly dependent on reactor exit temperature and to a lesser extent on residence time, with values ranging from about 20% to 1775/sup 0/F and 2-S residence time to zero at 1975/sup 0/F and residence times greater than 5 s. Carbon conversion to C/sub 6/H/sub 6/ asd high as 11.2% was obtained. Carbon conversion to CO/sub x/ ranged from 3.5 to 29.4%. Methane addition was found not to significantly affect the AFHP reactor chemistry. As a result of this program, Rockwell has expanded its data base and significantly improved its correlation model describing the processes occurring during flash hydropyrolysis. The correlation provides an excellent tool for subsequent process evaluations to determine the economic potential of the Rockwell coal hydrogasification process. 23 refs., 51 figs

  16. Blanket management method for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    A method for reducing thermal striping in liquid metal fast breeder reactors by reducing temperature gradients between adjacent fuel and blanket assemblies by shuffling blanket assemblies at each refueling outage so as to progressively shuffle the blanket assemblies to the core periphery through multiple moves and to generally locate fresh blanket assemblies adjacent to exposed fuel assemblies and exposed blanket assemblies adjacent to fresh fuel. Additionally, assembly orificing is altered to provide less flow to blanket assemblies needing less flow due to an otherwise decreased temperature gradient and providing additional flow to fuel assemblies which need more flow to sufficiently reduce temperature gradients to prevent thermal striping. (author)

  17. Solid breeder blanket design and tritium breeding

    International Nuclear Information System (INIS)

    Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements - to breed one tritium per fusion neutron - are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li2O, LiAlO2, Li2ZrO3, Li4SiO4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the amount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket. (orig.)

  18. Advanced handling-systems with enhanced performance flexibility

    International Nuclear Information System (INIS)

    This report describes the results of a project related to future applications and requirements for advanced handling systems. This report consists of six chapters. Following the description of the aims the tools for setting up the requirements for the handling systems including the experience during the data acquisition process is described. Furthermore some information is given about the current state of the art of robotics and manipulators. Of paramount importance are the descriptions of applications and related concepts in the following chapters leading to specific categories of advanced handling units. The paper closes with the description of the first concepts for realization. (orig./HP)

  19. 2001 GUIDELINES FOR ADVANCEMENT AND EXCEPTIONAL PERFORMANCE AWARDS

    CERN Multimedia

    Human Resources Division

    2001-01-01

    1. General Following the recent Council decision that 1 September 2001 will be the date of entry into force of the new Career Structure conditions, the 2001 annual advancement exercise will be conducted on the same basis as in previous years. The practical details of the implementation of the new Career Structure conditions from 1 September 2001 and for future annual advancement exercises are presently subject to concertation between the Management and the Staff Association and will be communicated to the staff in due course. In the meantime, therefore, this year's exercise will take place within the present Staff Rules and Regulations, and following the procedures of the present Administrative Circular No. 26 (rev. 2). The effective date of decisions will be 1 July 2001. The Guidelines will be as described last year in CERN Bulletin 10/2000, of which the main aspects are summarised below. 2. Advancement guidelines The budget allocation for advancement in 2001 as a percentage of the basic salary cost is the s...

  20. Advanced Modulation Techniques for High-Performance Computing Optical Interconnects

    DEFF Research Database (Denmark)

    Karinou, Fotini; Borkowski, Robert; Zibar, Darko;

    2013-01-01

    optical shared memory supercomputer interconnect system switch fabric. In particular, we investigate the resilience of the aforementioned advanced modulation formats to the nonlinearities of semiconductor optical amplifiers, used as ON/OFF gates in the supercomputer optical switch fabric under study. In...

  1. Neutronics R and D efforts in support of the European breeder blanket development programme

    International Nuclear Information System (INIS)

    The EU fusion technology programme considers two blanket development lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles as breeder material and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy acting both as breeder and neutron multiplier. The long-term strategy aims at providing validated engineering designs of breeder blankets for a fusion power demonstration reactor (DEMO). As an important intermediate step, the breeder blankets need to be tested in a real fusion environment as provided by ITER. HCPB and HCLL Test Blanket Modules (TBM) have been accordingly designed for tests in dedicated ITER blanket ports. The nuclear design and performance of the breeder blanket modules rely on the results provided by neutronics design calculations. Validated computational tools and qualified nuclear data are required for high prediction accuracies including reliable uncertainty assessments. Complementary to the application of established standard tools and data for design analysis, a dedicated neutronics R and D effort is therefore conducted in the EU. This includes the development of dedicated computational tools, the generation of high quality nuclear data and their validation through integral experiments. The recent neutronic design efforts have been devoted to the European DEMO reactor study comprising (i) Monte Carlo based pre-analysis for the dimensioning of the shielding system, (ii) the generation of a generic CAD based Monte Carlo geometry model, and (iii) performance analysis for HCLL and HCPB based DEMO variants. The recent focus of the validation effort is on neutronics TBM mock-up experiments. The first experiment of this kind was performed on a TBM mock-up of the HCPB breeder blanket. The follow-up experiment on a neutronics HCLL TBM mock-up is currently under preparation. Computational pre-analysis were performed to optimise the design of the mock

  2. Neutronics analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1 GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33 000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication.

  3. Neutronics and thermal design analyses of US solid breeder blanket for ITER

    International Nuclear Information System (INIS)

    The US Solid Breeder Blanket is designed to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Safety, low tritium inventory, reliability, flexibility cost, and minimum R ampersand D requirements are the other design criteria. To satisfy these criteria, the produced tritium is recovered continuously during operation and the blanket coolant operates at low pressure. Beryllium multiplier material is used to control the solid-breeder temperature. Neutronics and thermal design analyses were performed in an integrated manner to define the blanket configuration. The reference parameters of ITER including the operating scenarios, the neutron wall loading distribution and the copper stabilizer are included in the design analyses. Several analyses were performed to study the impact of the reactor parameters, blanket dimensions, material characteristics, and heat transfer coefficient at the material interfaces on the blanket performance. The design analyses and the results from the different studies are summarized. 6 refs., 3 figs., 3 tabs

  4. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

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

  5. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

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

  6. Hispanic Student Performance on Advanced Placement Exams: A Multiyear, National Investigation

    Science.gov (United States)

    Jara, Teresa Dianne

    2013-01-01

    Purpose: The purpose of this study was to analyze the Advanced Placement exams that Hispanic students complete and to compare their overall performance with the performance of White students from 2000 to 2012. A second purpose was to determine which Advanced Placement exams were the most difficult exams for Hispanic students and which Advanced…

  7. Nuclear, thermo-mechanical and tritium release analysis of ITER breeding blanket

    International Nuclear Information System (INIS)

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li2TiO3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li2TiO3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328degC in

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, A. [Univ. of California, Berkeley, CA (United States); Fratoni, M. [Univ. of California, Berkeley, CA (United States)

    2015-09-22

    , low electrical conductivity and therefore low MHD pressure drop, low chemical reactivity, and extremely low tritium inventory; the addition of sodium (FLiNaBe) has been considered because it retains the properties of FliBe but also lowers the melting point. Although many of these blanket concepts are promising, challenges still remain. The limited amount of beryllium available poses a problem for ceramic breeders such as the HCPB. FLiBe and FLiNaBe are highly viscous and have a low thermal conductivity. Lithium lead possesses a poor thermal conductivity which can cause problems in both DCLL and LiPb blankets. Additionally, the tritium permeation from these two blankets into plant components can be a problem and must be reduced. Consequently, Lawrence Livermore National Laboratory (LLNL) is attempting to develop a lithium-based alloy—most likely a ternary alloy—which maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) while reducing overall flammability concerns for use in the blanket of an inertial fusion energy (IFE) power plant. The LLNL concept employs inertial confinement fusion (ICF) through the use of lasers aimed at an indirect-driven target composed of deuterium-tritium fuel. The fusion driver/target design implements the same physics currently experimented at the National Ignition Facility (NIF). The plant uses lithium in both the primary coolant and blanket; therefore, lithium-related hazards are of primary concern. Although reducing chemical reactivity is the primary motivation for the development of new lithium alloys, the successful candidates will have to guarantee acceptable performance in all their functions. The scope of this study is to evaluate the neutronics performance of a large number of lithium-based alloys in the blanket of the IFE engine and assess their properties upon activation. This manuscript is organized as follows: Section 12 presents the models and methodologies used for the analysis; Section

  10. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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. (fusion engineering)

  12. A Thermal/Hydraulic Safety Assessment of the Blanket Conceptual Design for the Accelerator Production of Tritium Facility

    International Nuclear Information System (INIS)

    In support of the Accelerator Production of Tritium (APT) project, safety analyses for the blanket system have been performed based on the conceptual design for the Target/Blanket (T/B) Facility. During mitigated event sequences safety engineered features, such as the residual heat removal (RHR) and cavity flood systems, provide sufficient protection for maintaining the structural integrity of the blanket system and its components. During unmitigated (with beam shutdown only) event sequences, passive features such as natural circulation, thermal inertia, and boil-off provide significant time for corrective measures to be taken

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

    Science.gov (United States)

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

    2014-04-01

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

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

    International Nuclear Information System (INIS)

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

  15. Current status of safety design and safety analysis for China ITER helium coolant ceramic breeder test blanket system long

    International Nuclear Information System (INIS)

    Helium Coolant Ceramic Breeder (HCCB) Test Blanket System (TBS) designed by China are planned to be tested in ITER to validate key technologies, including demonstration of nuclear safety, for future fusion reactor breeding blankets. Furthermore, in order to be operated in ITER, a nuclear facility (INB) recognized by French nuclear safety authority, safety design and safety analysis of the TBS are mandatory for the licensing procedures. This paper summarizes the status at current design phase with following main elements: The main radiological source terms in the system are tritium and activation products. Nuclear and tritium analysis are performed to identify their inventories and distributions in system. Multiple confinement barriers are considered to be the most essential safety feature. French regulation for pressure equipment and nuclear equipment (ESP/ESPN regulations) will be followed to ensure the system integrities. ALARA principle is kept in mind during the whole safety design phases. Protective actions including choice of advanced materials, improvement of shielding, optimization of operation and maintenance activities, usage of remote handling operations, zoning and access control have been considered. Passive safety is emphasized in the system design, only minimal active safety functions including call for fusion plasma shutdown and isolation of TBM from ex-vessel ancillary systems. High reliability and redundancies are required for components related to these functions. Several accidents have been identified and analyzed. Consider the limited inventories in the system and the intrinsic safety of fusion device, positive conclusions have been obtained. (author)

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

    International Nuclear Information System (INIS)

    The key functions required for tritium breeding blankets for a fusion power reactor are: (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li2O/H2O/Be, Mo-alloy/Li2O/He/Be, Mo-alloy/LiAlO2/He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. (author)

  17. Phase-IIC experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Neutronics experiments on two types of heterogeneous blankets have been performed as the Phase-IIC experiment of JAERI/USDOE collaborative program on fusion blanket neutronics. The experimental system was used in the same geometry as the previous Phase-IIA series which was a closed geometry using neutron source enclosure of lithium carbonate. The heterogeneous blankets selected here are the beryllium edge-on and the water coolant channel assemblies. In the former the beryllium and lithium-oxide layers are piled up alternately in the front part of test blanket. In the latter, the three simulated water cooling channels are settled in the Li2O blanket. These are producing steep gradient of neutron flux around material boundary. The calculation accuracy and measurement method for these features is a key of interest in the experiments. The measurements were performed for tritium production rate and the other nuclear parameters as well as the previous experiments. This report describes the experimental detail and the results enough to use for the benchmark data for testing the data and method of design calculation of fusion reactors. (author)

  18. Requirements for helium cooled pebble bed blanket and R and D activities

    Energy Technology Data Exchange (ETDEWEB)

    Carloni, D., E-mail: dario.carloni@kit.edu; Boccaccini, L.V.; Franza, F.; Kecskes, S.

    2014-10-15

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.

  19. Experimental lithium-lead module for neutronics studies of fusion blankets

    International Nuclear Information System (INIS)

    In a (D,T) type fusion machine, about 80% of the fusion energy is transported by neutrons outside the reactor's core and deposited in the blanket, an assembly of materials surrounding the machine. Tritium breeders, such as lithium and lithium-lead (LiPb) eutectic alloys, mainly dictate the design of fusion blankets. Neutronics studies, on blanket module assemblies, form an initial step towards real construction of one or another blanket. Within the framework of this dissertation, different blanket elements: first wall/structural material, tritium breeder etc., and leading fusion blanket concepts are briefly reviewed. Lithium-lead eutectic is of particular interest since the neutron multiplication takes place in the breeder, where tritium is produced. Therefore, one-dimensional optimization calculations were performed to study the use of Li17Pb83, with natural lithium abundance. Generally, this breeder is used with very high Li6 enrichment. It was found that it would be difficult to design compact blankets and to achieve reasonable tritium production, with Li17Pb83 eutectic having natural lithium isotopic composition. Either the breeder should be highly enriched in Li6 or another enriched lithium zone should follow. This study, however, led to the design and construction of the Experimental Lithium-Lead Module (EL2M). The EL2M was also used for International Comparison on Measuring Techniques of Tritium Production Rate (TPR) for Fusion Neutronics Experiments, a program initiated by JAERI (Japan Atomic Energy Research Institute) in which eight international research institutions joined. (author) figs., tabs., 124 refs

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

    International Nuclear Information System (INIS)

    This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li2TiO3 and so on, fabrication technology developments and characterization of the Li2TiO3 and Li4SiO4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li2TiO3 and Li4SiO4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

  1. Requirements for helium cooled pebble bed blanket and R and D activities

    International Nuclear Information System (INIS)

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine

  2. Comparison of Performing and Not-Performing the Prophylactic Surgery for Urinary Incontinence in Advanced Pelvic Organ Prolapse

    OpenAIRE

    Parvin Bastani; Neda Shoari; Sakineh Haj Ebrahimi; Fatemeh Mallah; Azadeh Azadi

    2014-01-01

    Objectives: To compare the Performing and not-performing of prophylactic surgery for urinary incontinence in women with pelvic organ prolapse. Materials and Methods: In a randomized controlled clinical trial that performed in department of Gynecology and Obstetrics of Tabriz University of medical sciences on women with advanced pelvic organ prolapse without history of urinary incontinence. The effects of performing and not-performing of prophylactic surgery for urinary incontinence in wome...

  3. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2014-06-15

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

  4. Integral experiment on effects of large opening in fusion reactor blanket on tritium breeding using annular geometry

    International Nuclear Information System (INIS)

    An experiment involving a simulated blanket with an opening has been performed using the line source and annular blanket system developed under the JAERI/USDOE collaborative programme in order to examine the effects of the opening on neutronics parameters such as the tritium-breeding ratio. The annular test assembly was rectangular in shape and consisted of a lithium oxide blanket covered with graphite and SS304 which simulated the graphite armour plate and first wall in a fusion device. A large opening (376mm x 425.5mm) was made in the middle of the test blanket. This opening simulated a neutral beam injector.Tritium production rates and reaction rates were measured inside the blanket. Neutron spectra and reaction rates were also measured on the surfaces of both sides and without the opening of the inner cavity. The opening decreased the number of low energy neutrons contained in the cavity and especially decreased 6Li tritium production by 10% inside the blanket at the opposite side of the opening. The Monte Carlo code GMVP using the JENDL-3 nuclear data library predicted the measured nuclear parameters in the test blankets, such as the tritium production rate, to within 10% accuracy. (orig.)

  5. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%

  6. Implementation of an AMBA Advanced High Performance Bus protocol IP block

    OpenAIRE

    Mahendra Vucha Sandeep Magarde; Miss. Shruti Bhargava; Rishabh Singh Kurmi

    2011-01-01

    The design of an AMBA advanced high performance bus (AHB) protocol IP block is presented in this paper. The AHB (Advanced High-performance Bus) is a high-performance bus in AMBA (Advanced Microcontroller Bus Architecture) family. This AHB can be used in high clock frequency system modules. The AHB acts as the high-performance system backbone bus. AHB supports the efficient connection of processors, on-chip memories and off-chip external memory interfaces with low-power peripheral macro cell f...

  7. The Effect of the Advanced Placement Training and Incentive Program on Increasing Enrollment and Performance on Advanced Placement Science Exams

    Science.gov (United States)

    Ramsey, Susan Brady

    2012-01-01

    The purpose of this study is to examine the effectiveness of the National Math and Science Initiative's Advanced Placement Training and Incentive Program (APTIP) on the number of students taking AP science courses and their performance. The study evaluated 39 schools over a six-year period in six states that participate in the APTIP. The…

  8. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

    A review of recent research activities in liquid metal magnetohydrodynamics (LM-MHDs) is presented in this article. Two major reserach areas are discussed. The first topic involves the thermomechanical design issues in a proposed tokamak fusion reactor. The primary concerns are in the magneto-thermal-hydraulic performance of a self-cooled liquid metal blanket. The second topic involves the application of MHD in material processing in the metallurgical and semiconductor industries. The two representative applications are electromagnetic stirring (EMS) of continuously cast steel and the Czochralski (CZ) method of crystal growth in the presence of a magnetic field. (author) 24 figs., 10 tabs., 136 refs

  9. First wall/blanket/shield design and optimization system

    International Nuclear Information System (INIS)

    First wall/blanket/shield design and optimization system (BSDOS) has been developed to provide a state-of-the-art design tool for fast accurate analysis. In addition, it has been designed to perform several other functions: (1) allowing comparison and evaluation studies for different concepts using the same data bases and ground rules, (2) permitting the use of any figure of merit in the evaluation studies, (3) optimizing the first wall/blanket/shield design parameters for any figure of merit under several design constraints, (4) permitting the use of different reactor parameters in the evaluation and optimization analyses, (5) allowing the use of improved eingineering data bases to study the impact on the design performance for planning future research and development, and (6) evaluating the effect of the data base uncertainties on the design performance. BSDOS is the first design and optimization system to couple the highly interacting neutronics, heat transfer, thermal hydraulics, stress analysis, radioactivity and decay-heat analyses, tritium balance, and capital cost. A brief description of the main features of BSDOS is given in this paper. Also, results from using BSDOS to perform design analysis for several reactor components are presented. 17 refs., 1 fig., 2 tabs

  10. Connecting Performance Analysis and Visualization to Advance Extreme Scale Computing

    Energy Technology Data Exchange (ETDEWEB)

    Bremer, Peer-Timo [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mohr, Bernd [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Schulz, Martin [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Pasccci, Valerio [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Gamblin, Todd [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brunst, Holger [Dresden Univ. of Technology (Germany)

    2015-07-29

    The characterization, modeling, analysis, and tuning of software performance has been a central topic in High Performance Computing (HPC) since its early beginnings. The overall goal is to make HPC software run faster on particular hardware, either through better scheduling, on-node resource utilization, or more efficient distributed communication.

  11. Tritium recovery from ceramic breeder blanket

    International Nuclear Information System (INIS)

    It is known that chemical forms of tritium released from ceramic breeders are T2O and T2. Among issues relevant to the tritium chemical form, tritium inventory is one of the major criteria in the selection of breeder material. The primary purpose of this report is to study the dependence of tritium inventory in a blanket with ceramic solid breeder on the tritium chemical form. In this light, tritium inventory in a Li2O blanket has been evaluated as a function of tritium chemical form under the conditions of the Japanese Fusion Experimental Reactor (FER). It was shown that in a blanket with Li2O as a breeder, which has a strong affinity to water vapor, the inventory due to T2O adsorption becomes quite large. In order to reduce the T2O adsorption inventory, conversion of the tritium chemical form through an isotope exchange reaction with hydrogen added to the sweep gas (T2O + 2 H2 → H2O + 2 HT) has been proposed, and its advantages and problems have been examined. Lithium hydroxide formation and mass transfer, which are considered to be inherent in the Li2O-T2O system and to be critical issues for the feasibility of a Li2O blanket, have been also discussed. (author)

  12. Review: BNL graphite blanket design concepts

    International Nuclear Information System (INIS)

    A review of the Brookhaven National Laboratory (BNL) minimum activity graphite blanket designs is made. Three designs are identified and discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a thick graphite screen (typically 30 cm or greater, depending on type as well as application-experimental power reactor or commercial reactor). Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy. This energy is then either radiated to a secondary blanket with coolant tubes, as in types A and B, or is removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the structural material of the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude by the graphite screen, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma, whatever the degree of radiation damage

  13. Aerogel Blanket Insulation Materials for Cryogenic Applications

    Science.gov (United States)

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

    2009-01-01

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

  14. The Effect of Advanced Management Accounting Practices on the Competitive Strategies and Performance

    OpenAIRE

    Rahmi Yücel; Kayhan Ahmetoğulları

    2015-01-01

    This study is based on a sample of 300 managers from the production industry of West Marmara Region. The goal of this study is to examine the interaction among advanced management accounting practices, competitive strategies and company performance. As a result of study, it is found that advanced accounting management practices have a positive effect on the company performance and competitive strategies. In addition, competitive strategies have an effect on the company performance positively....

  15. Innovation Meets Performance Demands of Advanced Lithium-ion Batteries

    Energy Technology Data Exchange (ETDEWEB)

    2016-06-01

    Advancements in high capacity and low density battery technologies have led to a growing need for battery materials with greater charge capacity and therefore stability. NREL's developments in ALD and molecular layer MLD allow for thin film coatings to battery composite electrodes, which can improve battery lifespan, high charge capacity, and stability. Silicon, one of the best high-energy anode materials for Li-ion batteries, can experience capacity fade from volumetric expansion. Using MLD to examine how surface modification could stabilize silicon anode material in Li-ion batteries, researchers discovered a new reaction precursor that leads to a flexible surface coating that accommodates volumetric expansion of silicon electrodes.

  16. Development of performance criteria for advanced Viking seismic experiments

    Science.gov (United States)

    1972-01-01

    The characteristics and requirements of the seismic instrument for mapping the internal structure of the planet Mars are briefly described. The types of signals expected to exist are microseismic background generated by wind and pressure variations and thermal effects, disturbances of or in the landed vehicle, signals caused by faulting and volcanic activity, and signals due to meteoritic impacts. The advanced instrument package should include a short-period vertical component system, a long-period or wide-band 3-component system, a high frequency vertical component system, and a system for detection and rejection of lander noises. The Viking '75, Surveyor, and Apollo systems are briefly described as potential instruments to be considered for modification. Data processing and control systems are also summarized.

  17. Transport modeling and gyrokinetic analysis of advanced high performance discharges

    International Nuclear Information System (INIS)

    Predictive transport modeling and gyrokinetic stability analyses of demonstration hybrid (HYBRID) and Advanced Tokamak (AT) discharges from the International Tokamak Physics Activity (ITPA) profile database are presented. Both regimes have exhibited enhanced core confinement (above the conventional ITER reference H-mode scenario) but differ in their current density profiles. Recent contributions to the ITPA database have facilitated an effort to study the underlying physics governing confinement in these advanced scenarios. In this paper, we assess the level of commonality of the turbulent transport physics and the relative roles of the transport suppression mechanisms (i.e. ExB shear and Shafranov shift (α) stabilization) using data for select HYBRID and AT discharges from the DIII-D, JET, and AUG tokamaks. GLF23 transport modeling and gyrokinetic stability analysis indicates that ExB shear and Shafranov shift stabilization play essential roles in producing the improved core confinement in both HYBRID and AT discharges. Shafranov shift stabilization is found to be more important in AT discharges than in HYBRID discharges. We have also examined the competition between the stabilizing effects of ExB shear and Shafranov shift stabilization and the destabilizing effects of higher safety factors and parallel velocity shear. Linear and nonlinear gyrokinetic simulations of idealized low and high safety factor cases reveals some interesting consequences. A low safety factor (i.e. HYBRID relevant) is directly beneficial in reducing the transport, and ExB shear stabilization can win out over parallel velocity shear destabilization allowing the turbulence to be quenched. However, at low-q/high current, Shafranov shift stabilization plays less of a role. Higher safety factors (as found in AT discharges), on the other hand, have larger amounts of Shafranov shift stabilization, but parallel velocity shear destabilization can prevent ExB shear quenching of the turbulent

  18. Advancing the use of performance evaluation in health care

    DEFF Research Database (Denmark)

    Traberg, Andreas; Jacobsen, Peter

    2014-01-01

    Purpose – The purpose of this paper is to develop a framework for health care performance evaluation that enables decision makers to identify areas indicative of corrective actions. The framework should provide information on strategic pro-/regress in an operational context that justifies the need......, and two internal and five external databases are used for a quantitative data collection. Findings – By aggregating performance outcomes, collective measures of performance are achieved. This enables easy and intuitive identification of areas not strategically aligned. In general, the framework has...... proven helpful in an MRI unit, where operational decision makers have been struggling with extensive amounts of performance information. Research limitations/implications – The implementation of the framework in a single case in a public and highly political environment restricts the generalizing...

  19. Design and analysis of breeding blanket with helium cooled solid breeder for ITER-TBM

    International Nuclear Information System (INIS)

    Test blanket module (TBM) is one of important components in ITER. Some of related blanket technologies of future fusion, such as tritium self-sufficiency, the exaction of high-grade heat, design criteria and safety requirements and environmental impacts, will be demonstrated in ITER-TBM. In ITER device, the three equatorial ports have allocated for TBM testing. China had proposed to develop independently the ITER-TBM with helium cooled solid breeder in 12th meeting of test blanket workgroup (TBWG-12). In this work, the preliminary design and analysis for Chinese HCSB TBM will be carried out. The TBM must be contains the function of the first wall, breeding blanket, shield and structure. Finally, in the period of preliminary investigation, HCSB TBM design adopt modularization concept which is helium as coolant and tritium purge gas, ferritic/martensitic steel as structural material, Lithium orthosilicate (Li4SiO4) as tritium breeder, beryllium pebble as neutron multiplier. TBM is allocated in standard vertical frame port. HCSB TBM consist of first wall, backplate, breeding sub-modules, caps, grid and support plate, and breeding sub-modules is arranged by layout of 2 x 6 in blanket box. In this paper, main components of HCSB TBM will be described in detail, also performance analysis of main components have been completed. (authors)

  20. Nuclear analysis of DEMO water-cooled blanket based on sub-critical water condition

    International Nuclear Information System (INIS)

    Highlights: ► For sub-critical water condition, the size of cooling loop would be more longer, for example, 2 m. ► Local TBR is related to the material fraction of breeders and multipliers, the beryllium is the dominant. ► Front area of blanket is dominant for blanket design and it would contribute the most of TBR comparing to the backside zones. - Abstract: For the water-cooled solid blanket of DEMO, the nuclear analysis was performed based on present cooling piping system. Especially, distributions of neutron load and temperature were calculated with Pn is 5 MW/m2. Furthermore, the local TBR was optimized by changing the material proportion for each Pn level (1–5 MW/m2). It was confirmed that the size of cooling loop for sub-critical water could be used as about 2000 × 450 mm and the cooling pipe diameter of D is 12 mm, d is 9 mm at v is 5.36 m/s. The pipe pitches would vary with Pn level which is related to the blanket structure design. Nuclear heat distribution is the base to decide the distribution of cooling pipe positions. It was found that the local TBR of blanket would be dropped down along with the Pn level rising which was mainly depended on the thickness of beryllium variation. Finally, the layout of cooling pipes for each level was obtained.

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

    International Nuclear Information System (INIS)

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

  2. Performance Evaluation of INACT - INDECT Advanced Image Cataloguing Tool

    Directory of Open Access Journals (Sweden)

    Libor Michalek

    2012-01-01

    Full Text Available In this article, we describe the performance evaluation of INACT tool which is developed for cataloguing of high-level and low-level metadata of the evidence material. INACT tool can be used by police forces in the cases of prosecution of such crimes as as possession and distribution of child pornography (CP. In live forensic cases, the time to first hit (time when the first image containing e.g. CP is found is important, as then further legal actions are justified (such as arrest of the suspect and his hardware. The performance evaluation of first hit was performed on real data with the cooperation of Czech Police, Department of Internet Crime.

  3. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    International Nuclear Information System (INIS)

    This paper presents results of conceptual design activities and associated R and D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R and D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  4. EVA – Advanced method for performance evaluation in banks

    OpenAIRE

    Popa, Gabriela; MIHAILESCU Laurentiu; CARAGEA Codin

    2009-01-01

    EVA (Economic Valued Added) is a modern financial measurement tool that determines if a business is earning more than its true cost of capital. Including a cost for the use of equity capital sets EVA apart from more popular measures of bank performance, such as return on assets (ROA), return on equity (ROE), net banking income and the efficiency ratio, which do not consider the cost of equity capital employed. As a result, these measures may suggest a bank is performing well, when in fact it ...

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

    International Nuclear Information System (INIS)

    The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li2O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

  6. Performance of a high efficiency advanced coal combustor

    Energy Technology Data Exchange (ETDEWEB)

    Toqan, M.A.; Paloposki, T.; Yu, T.; Teare, J.D.; Beer, J.M. (Massachusetts Inst. of Tech., Cambridge, MA (United States))

    1989-12-01

    Under contract from DOE-PETC, Combustion Engineering, Inc. undertook the lead-role in a multi-task R D program aimed at development of a new burner system for coal-based fuels; the goal was that this burner system should be capable of being retrofitted in oil- or gas-fired industrial boilers, or usable in new units. In the first phase of this program a high efficiency advanced coal combustor was designed jointly by CE and MIT. Its burner is of the multiannular design with a fixed shrouded swirler in the center immediately surrounding the atomizer gun to provide the primary act,'' and three further annuli for the supply of the secondary air.'' The degree of rotation (swirl) in the secondary air is variable. The split of the combustion air into primary and secondary air flows serves the purpose of flame stabilization and combustion staging, the latter to reduce NO{sub x} formation.

  7. Development of Human Performance Analysis and Advanced HRA Methodology

    International Nuclear Information System (INIS)

    The purpose of this project is to build a systematic framework that can evaluate the effect of human factors related problems on the safety of nuclear power plants (NPPs) as well as develop a technology that can be used to enhance human performance. The research goal of this project is twofold: the development of a domestic human performance database, and the analysis of human error with constructing technical basis for human reliability analysis. There are three kinds of main results of this study. The first result is the development of a domestic human performance database, called OPERA-I (Operator Performance and Reliability Analysis, Part I). In addition, based on OPERA-I, task complexity (TACOM) measure and the methodology of optimizing diagnosis procedures were developed. The second main result is the standardization of HRA method. Finally, Misdiagnosis Tree Analysis (MDTA) technique that can analyze EOC has been developed. These results, such as OPERA-I, TACOM, standardized HRA method and MDTA technique, will be used to improve the quality of domestic PSA

  8. Performance of advanced high-temperature fuels for nuclear propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Stark, W.A.; Butt, D.P.; Storms, E.K.; Wallace, T.C. [Los Alamos National Lab., NM (United States)

    1994-12-31

    Nuclear propulsion using hydrogen has been demonstrated to operate at nearly twice the performance level of today`s chemical rockets. However, higher temperatures lead to a variety of degradations that compromise safety and longevity. Foremost among these is the melting of the propulsion reactor fuel. The melting behaviour of the U-Zr-C and U-Nb-C systems have been evaluated.

  9. Measurement-Based Performance Evaluation of Advanced MIMO Transceiver Designs

    Directory of Open Access Journals (Sweden)

    Schneider Christian

    2005-01-01

    Full Text Available This paper describes the methodology and the results of performance investigations on a multiple-input multiple-output (MIMO transceiver scheme for frequency-selective radio channels. The method relies on offline simulations and employs real-time MIMO channel sounder measurement data to ensure a realistic channel modeling. Thus it can be classified in between the performance evaluation using some predefined channel models and the evaluation of a prototype hardware in field experiments. New aspects for the simulation setup are discussed, which are frequently ignored when using simpler model-based evaluations. Example simulations are provided for an iterative ("turbo" MIMO equalizer concept. The dependency of the achievable bit error rate performance on the propagation characteristics and on the variation in some system design parameters is shown, whereas the antenna constellation is of particular concern for MIMO systems. Although in many of the considered constellations turbo MIMO equalization appears feasible in real field scenarios, there exist cases with poor performance as well, indicating that in practical applications link adaptation of the transmitter and receiver processing to the environment is necessary.

  10. Design and performance verification of advanced multistage depressed collectors

    Science.gov (United States)

    Kosmahl, H. G.; Ramins, P.

    1975-01-01

    Design and performance of a small size, 4 stage depressed collector are discussed. The collector and a spent beam refocusing section preceding it are intended for efficiency enhancement of octave bandwidth, high CW power traveling wave tubes for use in ECM.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-17

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

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

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

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

  13. Thermal cycle test of elemental mockups of ITER breeding blanket

    International Nuclear Information System (INIS)

    Thermal cycle tests for mockups of breeder pebble beds of ITER breeding blanket have been carried out to investigate their thermo-mechanical behavior with the interaction between a pebble bed and a breeder rod containing the breeder pebbles. The mockups have been designed to demonstrate a part of the Breeder Inside Tube (BIT)' structure of ITER breeding blanket. Candidate material pebbles of Li2TiO3 was applied as breeder specimen, and Al pebbles were applied for simulating the neutron multiplier of Be pebbles. These pebbles have been packed in test tubes by using a vibration machine. Tested configurations were single layer mockups with Li2TiO3 single diameter packing and binary packing beds, and double layer mockups with Li2TiO3/Al single diameter packing and binary packing beds. In order to clarify the deformation performance of breeder tube, two different thickness of the breeder rod were also tested: one for nominal condition and another for acceleration test. Pebble bed of Li2TiO3 is heated with an electric heater, which is equipped at the center of the breeder rod, simulating the temperature profile by volumetric heating of breeder pebbles. The outside of a breeder rod in a single layer mockup and the outside of the outer tube in case of double layer mockup is cooled by water. Temperature of the breeder beds has been controlled by a power input of the heater. After the thermal cycle tests, the internal dimensions and local packing fraction of mockups have been examined by using an X-ray CT device. As the result, no significant change of packing fraction was observed after five thermal cycles with maximum heater temperature of 600degC. Any bulging of the breeder rod or any cracking of the pebble has not been observed. A soundness of the typical structure and breeder pebble bed of ITER breeding blanket against thermal cycles was confirmed. (author)

  14. Fabrication techniques development of test blanket module based on CLAM

    International Nuclear Information System (INIS)

    The Reduced Activation Ferritic/Martensitic steels (RAFMs) are considered as the primary candidate structural material for the DEMO fusion reactor and the first fusion power plant. China Low Activation Martensitic (CLAM) steel, a version of RAFMs, is being developed in ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences), under wide collaboration with many institutes and universities in China and overseas. The designs of FDS (Fusion Design Study) series liquid LiPb blankets for fusion reactors and corresponding Dual Functional Lithium Lead (DFLL) Test Blanket Module (TBM) in International Thermonuclear Experimental Reactor (ITER) are being carried out in ASIPP. And CLAM steel is chosen as the primary candidate structural material in these designs. So the fabrication techniques for DFLL TBM with CLAM are or urgently needed to be studied in detail. The fabrication of DFLL TBM mainly includes the manufacturing of the First Wall (FW), the Cooling Plates (CP) and the joining of the FW and CPs. Currently, solid Hot Isostatic Pressing (HIP) bonding and uniaxial diffusion bonding method are the most promising candidate fabrication method for the FW and CP. Experiments of HIP and unixial diffusion bonding of CLAM/CLAM were carried out and good joints were obtained. As for the joining technique of FW and CPs, the fusion welding techniques such as Tungsten Inert Gas welding, Laser welding and Electron Beam welding are candidates. Preliminary experiments on these welding techniques were performed. The simulation of thermal process by Gleeble 2000 was also carried out. Results of these experiments are summarized and further R and D plan on blanket fabrication techniques is also stated. (authors)

  15. Production enhancement and quality degradation of Pu produced in FBR blankets

    International Nuclear Information System (INIS)

    Full text: For smooth deployment of FBR in near future, securing economy and non-proliferation are pivotal factors. This study has two main objectives related to these two key factors. One is to enhance Pu production efficiency in blanket region of fast breeder reactor core. The other is Pu composition degradation to improve the proliferation resistance of FBR fuel cycle. The former contributes to reduce amount of blanket fuels to be fabricated and reprocessed to gain the same quantity of Pu, consequently it is expected to improve the fuel cycle economy. The composition of Pu generated in FBR blanket generally has very high quality, namely, equal to weapon or super grade. The latter objective is to deteriorate the quality of generated Pu by adjusting neutron spectrum in the blanket region and the initial composition of blanket fuels. A conceptual core design of MOX fueled, sodium cooled FBR with 1500MWe rating is performed by using SLAROM and CIRATION code with nuclear data set prepared for fast reactors of JFS-3-J3.2. The initial fuels in the active core region contains 5% of minor actinides produced in UO2-fueled PWR with enrichment of 4.1% and burnup of 43GWd/t after 10 years cooling. A index FIR (fissile inventory ratio at EOC and BOC) is used to measure the net fissile balance that is suited for fuel cycle mass balance analysis in addition to the BR (breeding ratio) in the standard definition. The moderator material used to tailor the neutron spectrum in the blanket region is ZrH. The depleted uranium pins in the radial blanket are replaced by ZrH pins in the volume fraction range from 0% to 30%. By increasing ZrH pins, the Pu production per unit mass of UO2 in blanket increased from 0.08kg-Pu/kg-UO2 to more than 0.1kg-Pu/kg-UO2. The FIR also slightly improved by replacing small fraction of UO2 by ZrH pins however it turned for the worse in the higher range more than 5% of ZrH. Loading ZrH into blanket fuel assembly drastically affects the Pu composition at

  16. The Effect of Advanced Management Accounting Practices on the Competitive Strategies and Performance

    Directory of Open Access Journals (Sweden)

    Rahmi Yücel

    2015-10-01

    Full Text Available This study is based on a sample of 300 managers from the production industry of West Marmara Region. The goal of this study is to examine the interaction among advanced management accounting practices, competitive strategies and company performance. As a result of study, it is found that advanced accounting management practices have a positive effect on the company performance and competitive strategies. In addition, competitive strategies have an effect on the company performance positively. Furthermore, there is a positive and significant relation between long-term strategy based on management accounting practices and non-financial company performance and between activity based on management accounting practices and financial performance.

  17. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    International Nuclear Information System (INIS)

    Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

  18. Advances in exercise, fitness, and performance genomics in 2012.

    Science.gov (United States)

    Pérusse, Louis; Rankinen, Tuomo; Hagberg, James M; Loos, Ruth J F; Roth, Stephen M; Sarzynski, Mark A; Wolfarth, Bernd; Bouchard, Claude

    2013-05-01

    A small number of excellent articles on exercise genomics issues were published in 2012. A new PYGM knock-in mouse model will provide opportunities to investigate the exercise intolerance and very low activity level of people with McArdle disease. New reports on variants in ACTN3 and ACE have increased the level of uncertainty regarding their true role in skeletal muscle metabolism and strength traits. The evidence continues to accumulate on the positive effects of regular physical activity on body mass index or adiposity in individuals at risk of obesity as assessed by their FTO genotype or by the number of risk alleles they carry at multiple obesity-susceptibility loci. The serum levels of triglycerides and the risk of hypertriglyceridemia were shown to be influenced by the interactions between a single nucleotide polymorphism (SNP) in the NOS3 gene and physical activity level. Allelic variation at nine SNPs was shown to account for the heritable component of the changes in submaximal exercise heart rate induced by the HERITAGE Family Study exercise program. SNPs at the RBPMS, YWHAQ, and CREB1 loci were found to be particularly strong predictors of the changes in submaximal exercise heart rate. The 2012 review ends with comments on the importance of relying more on experimental data, the urgency of identifying panels of genomic predictors of the response to regular exercise and particularly of adverse responses, and the exciting opportunities offered by recent advances in our understanding of the global architecture of the human genome as reported by the Encyclopedia of DNA Elements project. PMID:23470294

  19. Performance modelling for product development of advanced window systems

    DEFF Research Database (Denmark)

    Appelfeld, David

    The research presented in this doctoral thesis shows how the product development (PD) of Complex Fenestration Systems (CFSs) can be facilitated by computer-based analysis to improve the energy efficiency of fenestration systems as well as to improve the indoor environment. The first chapter defines...... the hypothesis and objectives of the thesis, which is followed by an extended introduction and background. The third chapter briefly suggests the PD framework which is suitable for CFSs. The fourth and fifth chapter refer to the detailed performance modelling of thermal properties (chapter 4) and...... optical properties (chapter 5) of CFSs. The last chapter concludes the thesis and the individual investigations. It is complicated to holistically evaluate the performance of a prototyped system, since simulation programs evaluate standardised products such as aluminium venetian blinds. State...

  20. Advanced sensor technologies for high performance infrared detectors

    Science.gov (United States)

    Ziegler, J.; Bruder, M.; Finck, M.; Krüger, R.; Menger, P.; Simon, Th.; Wollrab, R.

    2002-06-01

    For high performance IR imaging and seeker systems AIM has established a high yield and reproducible HgCdTe detector technology. For continuous improvement of detector performance, yield and reliability, key processes have been optimized and new approaches have been developed. By a superior CdZnTe Bridgman growth process, dislocation densities crosstalk and improves modulation transfer function. For long linear arrays, a multichip-module-technique has been developed, which meets the demands for high temperature-cycle-reliability. In addition, a cycle-to-failure model has been established by cooldown tests on AIM-FPA's to predict cycle-to-failure at existing FPA approach or maximum allowable strain at demanded cycles-to-failure specification.

  1. Enhancing performance and reducing stress in sports technological advances

    CERN Document Server

    Ivancevic, Tijana; Greenberg, Ronald

    2015-01-01

    This book is designed to help athletes and individuals interested in high sports performance in their journey towards the perfection of human sports abilities and achievements. It has two main goals: accelerating the acquisition of motor skills and preparing and vigilantly reducing the recovery time after training and competition. The Diamond Sports Protocol (DSP) presents state-of-the-art techniques for current sport and health technologies, particularly neuromuscular electrical stimulation (Sports Wave), oxygen infusion (Oxy Sports), infrared (Sports Infrared Dome) and lactic acid cleaning (Turbo Sports). The book suggest DSP as an essential part of every future athlete's training, competition and health maintenance. The book is for everyone interested in superior sports performance, fast and effective rehabilitation from training and competition and sports injury prevention.

  2. Neutronics design and analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Neutronics calculations were performed to analysis the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS-EM (Energy Multiplier) blanket. The most significant and main goal of the water-cooled FDS-EM blanket is to achieve the energy gain of about 1 GW with self-sustaining tritium, which can operate for as long as possible without fuel unloading and reloading. The preliminarily designed neutronics parameters for FDS-EM were presented, which show that the blanket loaded with the Nuclear Waste (transuranic from 33 000 MWD/MTU PWR and depleted uranium) for energy multiplication (M≅90) with tritium self-sufficiency can operate for at least 10 years without fuel unloading and reloading. (authors)

  3. Use of advanced simulations in fuel performance codes

    International Nuclear Information System (INIS)

    The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

  4. Advancing Purchasing Performance in Medium Sized Manufacturing Companies

    OpenAIRE

    Lind, Linnea

    2009-01-01

      Background and research problem: In more recent years, the purchasing function has been given more attention as it has come to play a bigger role in improving and maintaining company performance. One of the underlying reasons for this could be the trend of increased outsourcing which implies that costs that used to be internal in form of wages and production costs are now external in the form of purchased items. This development therefore enhances the strategic importance of purchasing and...

  5. UCIMS: Advances in geotechnical construction and performance monitoring

    OpenAIRE

    Raphael Siebenmann; Hai-Tien Yu; Robert Bachus

    2015-01-01

    The Crossrail project currently under construction in Central London has been described as “The Big Dig on Steroids”, obviously referencing the Central Artery/Tunnel project in Boston completed in 2007. To address the multiple demands for timely construction performance monitoring, Crossrail envisioned the underground construction information management system (UCIMS) to monitor construction progress and structural health along the entire route, with a network of geotechnical instruments (i.e...

  6. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  7. Advanced multiphysics coupling for LWR fuel performance analysis

    International Nuclear Information System (INIS)

    Highlights: • Overviews the BISON nuclear fuel performance analysis code. • Discusses loose and tight coupling approaches applied to thermomechanics. • Outlines coupling to the DeCART neutronics code. • Demonstrates multiscale coupling using MARMOT. - Abstract: Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution

  8. Water-cooled blanket concepts for the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    The primary goal of the Blanket Comparison and Selection Study (BCSS) was to select a limited number of blanket concepts for fusion power reactors, to serve as the focus for the U.S. Department of Energy blanket research and development program. The concepts considered most seriously by the BCSS can be grouped for discussion purposes by coolant: liquid metals and alloys, pressurized water, helium, and nitrate salts. Concepts using pressurized water as the coolant are discussed. Water-cooled concepts using liquid breeders-lithium and 17Li-83Pb (LiPb)-have severe fundamental safety problems. The use of lithium and water in the blanket was considered unacceptable. Initial results of tests at Hanford Engineering Development Laboratory using steam injected into molten LiPb indicate that use of LiPb and water together in a blanket is a very serious concern from the safety standpoint. Key issues for water-cooled blankets with solid tritium breeders (Li2O, or a ternary oxide such as LiAlO2) were identified and examined: reliability against leaks, control of tritium permeation into the coolant, retention of breeder physical integrity, breeder temperature predictability, determination of allowable temperature limits for breeders, and 6Li burnup effects (for LiAlO2). The BCSS's final rankings and associated rationale for all water-cooled concepts are examined. Key issues and factors for tokamak and tandem mirror reactor versions of water-cooled solid breeder concepts are discussed. The reference design for the top-ranked concept-LiAlO2 breeder, ferritic steel structure, and beryllium neutron multiplier-is presented. Finally, some general conclusions for water-cooled blanket concepts are drawn based on the study's results

  9. Flibe blanket concept for transmuting transuranic elements and long lived fission products

    International Nuclear Information System (INIS)

    A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

  10. Advanced stellar compass - Onboard autonomous orbit determination, preliminary performance

    DEFF Research Database (Denmark)

    Betto, Maurizio; Jørgensen, John Leif; Jørgensen, Peter Siegbjørn;

    2004-01-01

    of this instrument, the authors have devised a method to determine the orbit of a spacecraft autonomously, onboard, and without a priori knowledge of any kind. The solution is robust and fast. This paper presents the preliminary performance obtained during the ground testing in August 2002 at the Mauna Kea...... Observatories. The main goals were: (1) to assess the robustness of the method in solving autonomously, onboard, the position lost-in-space problem; (2) to assess the preliminary accuracy achievable with a single planet and a single observation; (3) to verify the autonomous navigation (AutoNav) module could...

  11. Study of MHD Corrosion and Transport of Corrosion Products of Ferritic/Martensitic Steels in the Flowing PbLi and its Application to Fusion Blanket

    OpenAIRE

    Saeidi, Sheida

    2014-01-01

    Two important components of a liquid breeder blanket of a fusion power reactor are the liquid breeder/coolant and the steel structure that the liquid is enclosed in. One candidate combination for such components is Lead-Lithium (PbLi) eutectic alloy and advanced Reduced Activation Ferritic/Martensitic (RAFM) steel. Implementation of RAFM steel and PbLi in blanket applications still requires material compatibility studies as many questions related to physical/chemical interactions in the RAFM...

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

    International Nuclear Information System (INIS)

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

  13. Swiss fusion blanket experiments: Final report, November 1, 1985-October 31, 1987

    International Nuclear Information System (INIS)

    The major thrust of this project related to the effort to transfer the Lithium Blanket Module (LBM) to the Nuclear Engineering Laboratory of the Swiss Institute of Technology at Lausanne, and to the subsequent support with analytical calculations of a variety of experiments performed with the LBM. 12 refs

  14. Mechanics of advanced materials analysis of properties and performance

    CERN Document Server

    Matveenko, Valery

    2015-01-01

    The last decades have seen a large extension of types of materials employed in various applications. In many cases these materials demonstrate mechanical properties and performance that vary significantly from those of their traditional counterparts. Such uniqueness is sought – or even specially manufactured – to meet increased requirements on modern components and structures related to their specific use. As a result, mechanical behaviors of these materials under different loading and environmental conditions are outside the boundaries of traditional mechanics of materials, presupposing development of new characterization techniques, theoretical descriptions and numerical tools. The book presents interesting examples of recent developments in this area. Among the studied materials are bulk metallic glasses, metamaterials, special composites, piezoelectric smart structures, nonwovens, etc.

  15. Performance Evaluation of Complex Multiplier Using Advance Algorithm

    Directory of Open Access Journals (Sweden)

    Gopichand D. Khandale

    2013-06-01

    Full Text Available In this paper VHDL implementation of complex number multiplier using ancient Vedic mathematics and conventional modified Booth algorithm is presented and compared. The idea for designing the multiplier unit is adopted from ancient Indian mathematics "Vedas". The Urdhva Tiryakbhyam sutra (method was selected for implementation since it is applicable to all cases of multiplication. Multiplication using Urdhva Tiryakbhyam sutra is performed by vertically and crosswise. The feature of this method is any multi-bit multiplication can be reduced down to single bit multiplication and addition. On account of these formulas, the partial products and sums are generated in one step which reduces the carry propagation from LSB to MSB. The implementation of the Vedic mathematics and their application to the complex multiplier ensure substantial reduction of propagation delay. The simulation results for 4 bit multiplication using Booth’s algorithm and using Vedic sutra are illustrated.

  16. Advanced wet--dry cooling tower concept performance prediction

    Energy Technology Data Exchange (ETDEWEB)

    Snyder, T.; Bentley, J.; Giebler, M.; Glicksman, L.R.; Rohsenow, W.M.

    1977-01-01

    The purpose of this year's work has been to test and analyze the new dry cooling tower surface previously developed. The model heat transfer test apparatus built last year has been instrumented for temperature, humidity and flow measurement and performance has been measured under a variety of operating conditions. Tower Tests showed approximately 40 to 50% of the total energy transfer as taking place due to evaporation. This can be compared to approximately 80 to 85% for a conventional wet cooling tower. Comparison of the model tower test results with those of a computer simulation has demonstrated the validity of that simulation and its use as a design tool. Computer predictions have been made for a full-size tower system operating at several locations. Experience with this counterflow model tower has suggested that several design problems may be avoided by blowing the cooling air horizontally through the packing section. This crossflow concept was built from the previous counterflow apparatus and included the design and fabrication of new packing plates. Instrumentation and testing of the counterflow model produced data with an average experimental error of 10%. These results were compared to the predictions of a computer model written for the crossflow configuration. In 14 test runs the predicted total heat transfer differed from the measured total heat transfer by no more than 8% with most runs coming well within 5%. With the computer analogy's validity established, it may now be used to help predict the performance of fullscale wet-dry towers.

  17. Water-cooled lithium-lead blanket

    International Nuclear Information System (INIS)

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The present study examines whether the water-cooled lithium-lead blanket designed for NET can be directly extrapolated to a demonstration (DEMO) reactor. A fundamental requirement of the exercise is that the DEMO design should have a tritium breeding ratio which is higher than that in NET. The water-cooled lithium-lead blanket is discussed with respect to: neutronics design, design parameter survey and thermohydraulics, and engineering design. Results are reported of three-dimensional calculations using the Monte Carlo code MORSE-H to investigate possible neutron leakage between the poloidally disposed breeder tubes, and to determine the global tritium breeding ratio for the final double null machine design. (U.K.)

  18. Diffusive heat blanketing envelopes of neutron stars

    CERN Document Server

    Beznogov, M V; Yakovlev, D G

    2016-01-01

    We construct new models of outer heat blanketing envelopes of neutron stars composed of binary ion mixtures (H - He, He - C, C - Fe) in and out of diffusive equilibrium. To this aim, we generalize our previous work on diffusion of ions in isothermal gaseous or Coulomb liquid plasmas to handle non-isothermal systems. We calculate the relations between the effective surface temperature Ts and the temperature Tb at the bottom of heat blanketing envelopes (at a density rhob= 1e8 -- 1e10 g/cc) for diffusively equilibrated and non-equilibrated distributions of ion species at different masses DeltaM of lighter ions in the envelope. Our principal result is that the Ts - Tb relations are fairly insensitive to detailed distribution of ion fractions over the envelope (diffusively equilibrated or not) and depend almost solely on DeltaM. The obtained relations are approximated by analytic expressions which are convenient for modeling the evolution of neutron stars.

  19. Novel method for sludge blanket measurements.

    Science.gov (United States)

    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

  20. Stellar model atmospheres with magnetic line blanketing

    CERN Document Server

    Kochukhov, O; Shulyak, D

    2004-01-01

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

  1. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

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

  2. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Balaraman Kumar

    2010-06-01

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

  3. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Lalitha Jayaraman

    2010-01-01

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

  4. Fabrication of advanced military radiation detector sensor and performance evaluation

    International Nuclear Information System (INIS)

    shows a performance comparable to commercial ones. The beta ray sensor was manufactured using a BC-408 organic plastic scintillator. The main goal of this study was the measurement of beta - ray responses of the PMMA light guide attached to the sensor and the determination of window materials. Finally, the CsI(Tl) scintillator gamma ray sensor manufactured as engineering model and the shielding case was made of duralumin (side plane) and aluminium (window). The performance of the gamma - ray sensor was comparable to commercial high - sensitivity industrial sensors. These three types of scintillation sensors can be used directly as military equipment. And this study helps improving the protection of our country from nuclear threats

  5. Advanced Computational and Experimental Techniques for Nacelle Liner Performance Evaluation

    Science.gov (United States)

    Gerhold, Carl H.; Jones, Michael G.; Brown, Martha C.; Nark, Douglas

    2009-01-01

    The Curved Duct Test Rig (CDTR) has been developed to investigate sound propagation through a duct of size comparable to the aft bypass duct of typical aircraft engines. The axial dimension of the bypass duct is often curved and this geometric characteristic is captured in the CDTR. The semiannular bypass duct is simulated by a rectangular test section in which the height corresponds to the circumferential dimension and the width corresponds to the radial dimension. The liner samples are perforate over honeycomb core and are installed on the side walls of the test section. The top and bottom surfaces of the test section are acoustically rigid to simulate a hard wall bifurcation or pylon. A unique feature of the CDTR is the control system that generates sound incident on the liner test section in specific modes. Uniform air flow, at ambient temperature and flow speed Mach 0.275, is introduced through the duct. Experiments to investigate configuration effects such as curvature along the flow path on the acoustic performance of a sample liner are performed in the CDTR and reported in this paper. Combinations of treated and acoustically rigid side walls are investigated. The scattering of modes of the incident wave, both by the curvature and by the asymmetry of wall treatment, is demonstrated in the experimental results. The effect that mode scattering has on total acoustic effectiveness of the liner treatment is also shown. Comparisons of measured liner attenuation with numerical results predicted by an analytic model based on the parabolic approximation to the convected Helmholtz equation are reported. The spectra of attenuation produced by the analytic model are similar to experimental results for both walls treated, straight and curved flow path, with plane wave and higher order modes incident. The numerical model is used to define the optimized resistance and reactance of a liner that significantly improves liner attenuation in the frequency range 1900-2400 Hz. A

  6. Fissile fuel breeding in DT fusion reactor blankets

    International Nuclear Information System (INIS)

    Results of neutronic evaluations of fissile fuel breeding in a variety of DT fusion hybrid-reactor blankets are presented. The blankets are of the fast-fission or fission-suppressed rather than fission-enhanced designs, i.e. in the blankets considered emphasis is on fissile fuel rather than power production. For 233U breeding, when Li metal is the coolant for the first wall and the graphite moderator and the tritium breeding constituent of the blanket, the number of atoms of 233U produced per fusion in blankets that could be of practical interest is in the range 0.5 - 0.68, with the lower value applying to water-cooled ThO2 fertile fuel, the upper to gas-cooled Th-metal fuel located next to the reactor first wall. Neutron multipliers like Pb or Be can increase the production to about 0.74. For 239Pu breeding, the production ratio in practical blankets is 0.6 - 1.64, with the best results being for gas, Na- or Li-metal-cooled U-metal fuels located adjacent to the first wall (the U is depleted uranium). Gas-cooled U-Th-metal blankets, optimized for 233U breeding, yield 0.76 atoms of 233U and 0.38 atoms of 239Pu. The blanket energy multiplication factors are in the range 1.6 - 2.5 for Th blankets, 2.5 - 9.0 for U blankets and approximately 5.5 for the U-Th-metal blanket. The tritium breeding ratio in all blankets is 1.075. Blankets with other first wall, coolant and tritium breeding constituents are also considered. The fusion power requirements of hybrids that could supply the fuel needs of thorium-burning CANDU power reactors, and the allowed costs for building the hybrids are indicated

  7. Study on compact design of remote handling equipment for ITER blanket maintenance

    International Nuclear Information System (INIS)

    In the ITER, the neutrons created by D-T reactions activate structural materials, and thereby, the circumstance in the vacuum vessel is under intense gamma radiation field. Thus, the in-vessel components such as blanket are handled and replaced by remote handling equipment. The objective of this report is to study the compactness of the remote handling equipment (a vehicle/manipulator) for the ITER blanket maintenance. In order to avoid the interferences between the blanket and the equipment during blanket replacement in the restricted vacuum vessel, a compact design of the equipment is required. Therefore, the compact design is performed, including kinematic analyses aiming at the reduction of the sizes of the vehicle equipped with a manipulator handling the blanket and the rail for the vehicle traveling in the vacuum vessel. Major results are as follows: 1. The compact vehicle/manipulator is designed concentration on the reduction of the rail size and simplification of the guide roller mechanism as well as the reduction of the gear diameter for vehicle rotation around the rail. Height of the rail is reduced from 500 mm to 400 mm by a parameter survey for weight, stiffness and stress of the rail. The roller mechanism is divided into two simple functional mechanisms composed of rollers and a pad, that is, the rollers support relatively light loads during rail deployment and vehicle traveling while a pad supports heavy loads during blanket replacement. Regarding the rotation mechanism, the double helical gear is adopted, because it has higher contact ratio than the normal spur gear and consequently can transfer higher force. The smaller double helical gear, 996 mm in diameter, can achieve 26% higher output torque, 123.5 kN·m, than that of the original spur gear of 1,460 mm in diameter, 98 kN·m. As a result, the manipulator becomes about 30% lighter, 8 tons, than the original weight, 11.2 tons. 2. Based on the compact design of the vehicle/manipulator, the

  8. A study of performance evaluation of advanced CVD system

    Energy Technology Data Exchange (ETDEWEB)

    Kuribara, Masayuki [Central Research Inst. of Electric Power Industry, Tokyo (Japan); Yoshida, Kenji; Takeyama, Kaneyoshi

    1995-12-31

    UV-II type ICVD(Improved Cerenkov Viewing Device) is used to identify LWR spent fuel assemblies stored in a pond. However, it is difficult or impossible to apply ICVD to low burnup or long cooling time assemblies because of their feeble Cerenkov glow images. Then, ACVD system has been developed. ACVD uses newly developed proximity focused type UV-II, a CCD camera and an image processor to improve Cerenkov glow image quality. In this paper, an observation range of ACVD is discussed based on the performance comparisons between ICVD and ACVD system and the characteristics of human eyes. The observation range of ACVD is extended more than 2 times compared to that of ICVD with the real time operation mode of CCD camera (30 frames/second). When an image is integrated 1 second in the CCD camera, the observation range of ACVD system is extended more than 4.5 times. Moreover, when an image processor is used with the CCD camera, it is extended more than 10 times compared to that of ICVD. As the result, this system can detect Cerenkov glow image radiated from spent fuel assembly which burnup and cooling time range more than 4,000(MWD/MTU) and less than 30 years. (author)

  9. Detection of Breeding Blankets Using Antineutrinos

    Science.gov (United States)

    Cogswell, Bernadette; Huber, Patrick

    2016-03-01

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

  10. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

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

  11. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  12. Fabrication of prototype mockups of ITER shielding blanket with separable first wall

    International Nuclear Information System (INIS)

    Design of shielding blanket for ITER-FEAT applies the first wall which has the separable structure from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. This report summarizes the demonstrative fabrication of the ITER shielding blanket with separable first wall performed for the shielding blanket fabrication technology development, under the task agreement of G 16 TT 108 FJ (T420-2) in ITER Engineering Design Activity Extension Period. The objectives of the demonstrative fabrication are: to demonstrate the comprehensive fabrication technique in a large scale component (e.g the joining techniques for the beryllium armor/copper alloy and copper alloy/SS, and the slotting method of the FW and shield block); to develop an improved fabrication method for the shielding blanket based on the ITER-FEAT updated design. In this work, the fabrication technique of full scale separable first wall shield blanket was confirmed by fabricating full width Be armored first wall panel, full scale of 1/2 shield block with poloidal cooling channels. As the R and D for updated cooling channel configuration, the fabrication technique of the radial channel shield block was also demonstrated. Concluding to the all R and D results, it was demonstrated successfully that the fabrication technique and optimized conditions in the results obtained under the task agreement of G 16 TT 95 FJ (T420-1) was applicable to the prototype of the separable first wall blanket module. Additionally, basic echo data of ultra-sonic test method (UT) was obtained to show the applicability of UT method for in tube access detection of defect on the Cu alloy/SS tube interface. (author)

  13. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 2. Detailed version

    International Nuclear Information System (INIS)

    A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary. Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated RandD-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required RandD-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.)

  14. Damage Identification of Continuum Structures Using Blanketing Effect

    International Nuclear Information System (INIS)

    A damage identification method for the continuum structures with making use of the blanketing effect is proposed in this paper. The conventional damage indicator methods such as flexibility matrix method are difficult to deal with the damage identification of complex continuum structure such as multi-span beam. The presented method takes the change rates according to certain diagonal elements of the damaged structure flexibility matrix as identification indicator function. The 'blanketing effect' of the indicator function with respect to the damage factors makes it easy to identify the damage of multi-span beam and plate structures. Some useful formulas are derived and the sensitivity matrix of the identification indicator function is studied. The numerical simulation is performed and the identification results of several damage status are compared. Both beam and plate structures are used for numerical simulation. The results show that the presented indicator function method is very brevity and effective. It is especially suitable for the damage identification of the multi-span beam and plate structures such as bridges and floor

  15. Conceptual study of a helium cooled ceramics/Beryllium blanket for a power reactor

    International Nuclear Information System (INIS)

    In the frame of recent CEA studies aiming at the evaluation and the comparison of various candidate blanket concepts in view of their possible extrapolation to anticipated power reactor operating conditions (p/subNgreater than or equal to2 MW/m2), the present work examines the performances of a design which combines the attractive thermal performances of helium cooling in the radial direction, which minimizes the breeder temperature gradient along a cooling channel, with the promising breeding capability of composite Beryllium/LiA102 (85/15 %) breeder elements. The optimization of the neutronic and thermomechanical performances converges on a canister blanket concept, featured by a breeding capability in excess of 1.45, a pumping power of only 1 % of the thermal power and a breeder temperature distribution quasi uniform throughout the blanket (500 +- 200C) and largely independent of the power level. This unique feature provides a natural adaptation to ceramic breeders assigned to very strict and narrow working conditions and provides a valuable margin for any change in the thermal and heat transfer characteristics over the blanket lifetime

  16. Advanced Coatings Enabling High Performance Instruments for Astrophysics Missions

    Science.gov (United States)

    Nikzad, Shouleh

    We propose a three-year effort to develop techniques for far-ultraviolet (FUV) and ultraviolet coatings both as reflective optics coatings and as out-of-band-rejection (solar-blind) filters that will have a dramatic effect on the throughput and efficiency of instruments. This is an ideal time to address this problem. On the one hand, exciting new science questions posed in UV and optical realm place exacting demands on instrument capabilities far beyond HST-COS, FUSE, and GALEX with large focal plane arrays and high efficiency requirements. And on the other hand, the development of techniques and process such as atomic layer deposition with its atomically precise capability and nano-engineered materials approach enables us to address the challenging materials issues in the UV where interaction of photons and matter happen in the first few nanometers of the material surface. Aluminum substrates with protective overlayers (typically XFy, where X = Li, Mg, or Ca) have been the workhorse of reflective coatings for ultraviolet and visible instruments; however, they have demonstrated severe limitations. The formation of oxide at the Al-XFy interface and thick protective layers both affect the overall optical performance, leading to diminished reflection at shorter wavelengths. To address these long-standing shortcomings of coatings, we will use our newly developed processes and equipment to produce high-quality single- and multi-layer films of a variety of dielectrics and metals deposited with nano-scale control. JPL s new ALD system affords high uniformity, low oxygen background, good plasma processes, and precise temperature control, which are vital to achieving the large scale, uniform, and ultrathin films that are free of oxygen at interfaces. For example, ALD-grown aluminum can be protected using our newly developed chemistry for ALD magnesium fluoride. Our work will verify that the ALD technique reliably prevents the oxidation of aluminum, and will subsequently be

  17. The ITER Blanket System Design Challenge

    International Nuclear Information System (INIS)

    Full text: The blanket system is one of the most technically challenging components of the ITER machine, having to accommodate high heat fluxes from the plasma, large electromagnetic loads during off-normal events and demanding interfaces with many key components (in particular the vacuum vessel and in-vessel coils) and the plasma. Plasma scenarios impose demanding requirements on the blanket in terms of heat fluxes on various areas of the first wall during different phases of operation (inboard and outboard midplane for start-up/shut-down scenarios and the top region close to the secondary X-point during flat top) as well as large electro-magnetic (EM) loads and transient energy deposition during off-normal plasma events (such as disruptions and vertical displacement events (VDE)). The high heat fluxes resulting in some areas have necessitated the use of “enhanced heat flux” panels capable of accommodating an incident heat flux of up to 5 MW/m2 in steady state. The other regions utilize “normal heat flux” panels, which have been developed and tested for a heat flux of the order of 1 — 2 MW/m2. The FW shaping design requires a compromise between the conflicting requirements for accommodation of steady state and transient loads (energy deposition during off-normal events). A shaped surface increases the heat loads which are due to plasma particles following the field lines compared to a perfectly toroidal surface. The blanket provides a major contribution to the shielding of the vacuum vessel and coils. A challenging criterion is the need to limit the integrated heating in the toroidal field coil (TFC) to ∼ 14 kW. This is particularly severe on the inboard leg where approximately 80% of the total nuclear heat on the TFC is deposited. Several design modifications were considered and analyzed to help achieve this, including increasing the inboard blanket radial thickness and reducing the assembly gaps. This paper summarizes the latest progress in the

  18. How to measure human performance in main control room of an advanced NPP?

    International Nuclear Information System (INIS)

    As CRT-based display and advanced information technology were applied to advanced reactor such as APR-1400 (Advanced Power Reactor-1400), human operators' tasks became more cognitive works. Human Factors Engineering (HFE) became more important in designing the MCR (Main Control Room) of an advanced reactor. In order to support the advanced reactor design certification reviews, the Human Factors Engineering Program Review Model (HFE PRM) was developed with the support of U.S. NRC. The HFE PRM describes the HFE program elements that are necessary and sufficient to develop an acceptable detailed design specification and an acceptable implemented design and provides the review criteria for their evaluation. One of the review elements is human factors verification and validation (V and V). The role of V and V evaluations in the HFE PRM is to comprehensively determine that the design conforms to HFE design principles and it enables plant personnel to successfully perform their tasks to achieve plant safety and other operational goals. Integrated System Validation (ISV) is part of this review activity. An integrated system design is evaluated through performance-based tests to determine whether it acceptably supports safe operation of the plant. The performance- based tests are based on several human (operator) performance measures such as plant performance, personnel task, situation awareness, workload, team work, and anthropometric/physiological factors. In this work, some techniques already developed in nuclear or other industry and new techniques are incorporated into a methodology for the human performance evaluation

  19. Fast-Breeder-Blanket Project: FBBF. Final report

    International Nuclear Information System (INIS)

    This report is the final report for DOE contract DE-AC02-76ET37237 with the Purdue Fast Breeder Blanket Project. The Project was initiated to investigate the uncertainties in Fast Breeder Reactor blanket calculations. Absolute measurements of key neutron reaction rates, neutron spectra, and gamma-ray energy depositions were made in simulated FBF blankets in the Fast Breeder Blanket Facility (FBBF), a Cf-252 driven subcritical facility. Calculation of the spectra and integral reaction rates were made using methods, computer codes, and cross section data typical of those currently used in the design of FBR's. Comparisons of calculated to experimental integral neutron reaction rates give good agreement at the inner portions of the blanket by diverge to C/E ratios of about 0.65 at the outer edge of the blanket for reactions sensitive to the neutron density

  20. The frontiers of research on fusion blanket technology

    International Nuclear Information System (INIS)

    Current topics concerning blanket technology are reviewed. In the chemical engineering/chemistry area, the qualitative and quantitative effects of mass transfer steps of tritium is important in the understanding of the behavior of bred tritium in the solid breeder blanket system. Such phenomena as adsorption, isotope exchange reactions, and water formation reaction at the grain surface produce profound effects on the behavior of the bred tritium in the blanket. Regarding the liquid system, the physical or chemical properties of Li, Li17Pb83 and Flibe as liquid blanket materials were compared. Some recent studies were introduced regarding tritium recovery from the liquid blanket materials, impurity removal from salts, ceramic coating of structural materials, and the vapor pressure of mixtures of metals or salts. Thermal hydraulic topics in relation to several candidate power reactor concepts are summarized. Emphasis is laid on the simultaneous removal of heat and tritium from the blanket and some aspects of forming effective power cycles are developed. (author)

  1. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R-and-D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  2. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R(and)D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  3. A standard data set for performance analysis of advanced IR image processing techniques

    NARCIS (Netherlands)

    Weiss, A.R.; Adomeit, U.; Chevalier, P.; Landeau, S.; Bijl, P.; Champagnat, F.; Dijk, J.; Göhler, B.; Landini, S.; Reynolds, J.P.; Smith, L.N.

    2012-01-01

    Modern IR cameras are increasingly equipped with built-in advanced (often non-linear) image and signal processing algorithms (like fusion, super-resolution, dynamic range compression etc.) which can tremendously influence performance characteristics. Traditional approaches to range performance model

  4. Preliminary thermo-mechanical analysis of ITER breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

    Thermo-mechanical analysis has been conducted on ITER breeding blanket taking into account thermo-mechanical characteristics peculiar to pebble beds. The features of the analysis are to adopt an elasto-plastic constitutive model for pebble beds and to take into account spatially varying thermal conductivity and heat transfer coefficient, especially in the Be pebble bed, depending on the stress. ABAQUS code and COUPLED TEMPERATURE-DISPLACEMENT procedure of the code are selected so that thermal conductivity is automatically calculated in each calculation point depending on the stress. The modified DRUCKER-PRAGER/Cap plasticity model for granular materials of the code is selected so as to deal with such mechanical features of pebble bed as shear failure flow and hydrostatic plastic compression, and capability of the model is studied. The thermal property-stress correlation used in the analysis is obtained based on the experimental results at FZK and the results of additional thermo-mechanical analysis performed here. The thermo-mechanical analysis of an ITER breeding blanket module has been performed for four conditions: case A; nominal case with spatial distribution of thermal conductivity and heat transfer coefficient in Be pebble bed depending on the stress, case B; constant thermal conductivity, case C; thermal conductivity = -20% of nominal case, and case D; thermal conductivity = +20% of nominal case. In the nominal case the temperature of breeding material (Li{sub 2}ZrO{sub 3}) ranges from 317degC to 554degC and the maximum temperature of Be pebble bed is 446degC. It is concluded that the temperature distribution is within the current design limits. Though the analyses performed here are preliminary, the results exhibit well the qualitative features of the pebble bed mechanical behaviors observed in experiments. For more detail quantitative estimates of the blanket performance, further investigation on mechanical properties of pebble beds by experiment

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

    International Nuclear Information System (INIS)

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

  6. Neutronic analysis of alternative structural materials for fusion reactor blankets

    International Nuclear Information System (INIS)

    The neutronic performance of the International Tokamak Reactor (INTOR) blanket was studied when several alternative structural materials were used instead of the INTOR reference structural material, type 316 stainless steel. The alternative structural materials included: ferritic-, vanadium-, titanium-, long range ordered-, manganese austenitic-, and nimonic-alloys. All were treated both with and without a first-wall coating of beryllium or graphite. The tritium breeding ratio, the nuclear heating, and the gas (hydrogen and helium) production rates in the structural materials were calclated for the possible combinations of structural material and first-wall coating. These parameters were compared with those obtained by using SS-316. The nimonic alloy was the only one with worse neutronic performance than the SS-316. (orig.)

  7. Performance Assessment using ACM in DVB-S2/Advanced DVB-RCS Satellite Systems

    OpenAIRE

    Boussemart, Vincent; Brandt, Hartmut

    2009-01-01

    This paper presents the benefits of using adaptive coding and modulation in DVB-S2/advanced DVB-RCS satellite systems. The results are based on a satellite system simulator which was developed during an ESA study. Such a simulator aims at making performance assessments before implementing a system, at developing and evaluating new algorithms like schedulers, encapsulators or resource managers. The simulator is capable of reproducing DVB-S2 and advanced DVB-RCS standards and techniques so that...

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

    International Nuclear Information System (INIS)

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li2TiO3, tritium release behavior of Li2TiO3 and Li2ZrO3 including tritium diffusion, modeling of tritium release from Li2ZrO3 in ITER condition, helium release behavior from Li2O, results of tritium release irradiation tests of Li4SiO4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  9. Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

    International Nuclear Information System (INIS)

    Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R and D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R and D progress on these areas are introduced here

  10. Oxidation behavior of SiC/SiC composites for helium cooled solid breeder blanket

    International Nuclear Information System (INIS)

    In order to evaluate the oxidation behavior and mechanism of SiC/SiC composites with conventional pyrolitic graphite interface (PyC-SiC/SiC) and advanced multilayer interface (ML-SiC/SiC) in a HCSB blanket environment, a thermal gravimetric analysis (TGA) in He + O2 environment at 1000 deg. C and 1200 deg. C was performed. The PyC-SiC/SiC at 1200 deg. C and the ML-SiC/SiC at 1000 deg. C and 1200 deg. C showed relatively smaller weight change during oxidation because SiO2 formed on the SiC-matrix and SiC-fiber sealed the specimen surface before the PyC interface recession by gasification of graphite due to relatively high SiO2 formation rate. While the PyC-SiC/SiC at 1000 deg. C showed significant weight loss because the specimen surface was not sealed by SiO2 and significant PyC interface recession occurred due to relatively slow SiO2 formation

  11. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 3: External HR Break at Pump Outlet without Pump Trip

    International Nuclear Information System (INIS)

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal (HR) system. These simulations were performed for the Preliminary Safety Analysis Report

  12. APT Blanket System Loss-of-Flow Accident (LOFA) Analysis Based on Initial Conceptual Design - Case 1: with Beam Shutdown and Active RHR

    International Nuclear Information System (INIS)

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report

  13. Oxidation behavior of SIC/SIC composites for helium cooled solid breeder blanket

    International Nuclear Information System (INIS)

    SiC/SiC composite is one of the candidate structural materials for a fusion reactor blanket because of its low induced radioactivity, excellent high temperature mechanical properties and excellent radiation resistance. Helium (He) gas cooled blanket (HCSB) has been considered as one of the blanket design concepts using the SiC/SiC composite for relatively high temperature plant operation. Chemical stability, especially an oxidation resistance, is a key issue to be solved for the HCSB structural material because He gas in the HCSB might include partial oxygen. The desired strength of SiC/SiC composite can be given by an optimized interface layer between the fiber and matrix (F/M interface). In order to improve its mechanical properties, several advanced F/M interfaces such as an SiC/C multilayer (ML) and a porous SiC have been developed. However, SiC/SiC composites have a possibility of F/M interface degradation by oxidation at fusion reactor operating condition, for example by the reaction of C+O2→CO2. The purpose of this study is to evaluate the oxidation behavior of SiC/SiC composites with conventional pyrolitic carbon interface (PyC-SiC/SiC) and advanced multilayer interface (ML-SiC/SiC) in He+O2 environments at 1273 K. The SiC/SiC composites used in this work were fabricated at ORNL. Reinforced SiC fiber was 1D Hi-Nicalon Type-S fiber. SiC matrix was beta-SiC fabricated by a forced chemical vapor infiltration (FCVI) process. The average thickness of F/M interface (pyrolitic carbon and SiC/C multilayer) was 1000nm. Samples were machined into 2 mm x 1.5 mm x 2 mm blocks and the surface of them was mechanically polished. Oxidation tests were carried out using a thermal gravimetric analysis (TGA) equipment. Mixtures of He with 1500ppm O2 were used. Samples were heated from room temperature to the test temperature (1273 K) at 40 K/min. and then held at the test temperature for 100 h. Experimental conditions included 100 sccm flow rate and system pressure of 1 atm

  14. Development of a qualitative evaluation framework for performance shaping factors (PSFs) in advanced MCR HRA

    International Nuclear Information System (INIS)

    Highlights: → Context changes in advanced MCR have impact on PSFs. → PSFs in the 1st and 2nd generation HRA methods are reviewed. → We made a qualitative evaluation framework for PSF based on human factor issues. - Abstract: Human reliability analysis (HRA) is performed as part of the probabilistic risk assessment to identify and quantify human actions and the associated impacts on structures, systems, and components of complex facilities. In performing HRA, conditions that influence human performance have been analyzed in terms of several context factors. These context factors, which are called performance shaping factors (PSFs) are used to adjust the basic human error probability (BHEP), and PSFs have been derived in various ways depending on the HRA methods used. As the design of instrumentation and control (I and C) systems for nuclear power plants (NPPs) is rapidly moving toward fully digital I and C, and modern computer techniques have been gradually introduced into the design of advanced main control room (MCR), computer-based human-system interfaces (HSIs), such as CRT-based displays, large display panels (LDPs), advanced information systems, soft control, and computerized procedure system (CPS) will be applied in advanced MCR. Environmental changes in MCR have some implications for PSFs, and they have an influence on when PSFs should be applied in HRA because different situations might induce different internal or external factors which can lead to human errors. In this study, PSFs for advanced MCR HRA are derived, and a new qualitative evaluation framework for these PSFs is suggested. First, PSFs from various HRA methods are collected, and these PSFs are further grouped into PSFs categories to be used in advanced MCR HRA. Second, human factor (HF) issues in advanced MCR are analyzed and derived to be used as an evaluation framework for PSFs.

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

    International Nuclear Information System (INIS)

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

  16. The ARIES-AT advanced tokamak, Advanced technology fusion power plant

    International Nuclear Information System (INIS)

    The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R and D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ x = 2.2) which is the result of a 'thinner' blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β N. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb-17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 deg. C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb-17Li to about 1000 deg. C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES

  17. The EU power plant conceptual study - neutronic design analyses for near term and advanced reactor models

    International Nuclear Information System (INIS)

    A power plant conceptual study (PPCS) has been conducted in the framework of the European fusion programme with the main objective to demonstrate the safety and environmental advantages and the economic viability of fusion power. Power plant models with limited (''near term concepts'') and advanced plasma physics and technological extrapolations (''advanced concepts'') were considered. Two near term plant models were selected, one employing a water cooled lithium-lead (WCLL), and the other one a helium cooled pebble bed (HCPB) blanket. Two variants were also considered for the advanced power plant models, one adopting a liquid metal blanket with a self-cooled lithium-lead breeder zone and a helium cooled steel structure (''dual coolant lithium lead'', DCLL), and the other one a self-cooled lithium-lead (SCLL) blanket with SiCf/SiC composite as structural material. This report provides a detailed documentation of the neutronics design analyses performed as part of the PPCS study for both the near term and advanced power plant models. Main issues are the assessment of the tritium breeding capability, the evaluation of the nuclear power generation and its spatial distribution, and the assessment and optimisation of the shielding performance. The analyses were based on three-dimensional Monte Carlo calculations with the MCNP code using suitable torus sector models developed for the different PPCS plant variants. (orig.)

  18. Transient electromagnetic and dynamic structural analyses of a blanket structure with coupling effects

    International Nuclear Information System (INIS)

    Transient electromagnetic and dynamic structural analyses of a blanket structure in the fusion experimental reactor (FER) under a plasma disruption event and a vertical displacement event (VDE) have been performed to investigate the dynamic structural characteristics and the feasibility of the structure. Coupling effects between eddy currents and dynamic deflections have also been taken into account in these analyses. In this study, the inboard blanket was employed because of our computer memory limitation. A 1/192 segment model of a full torus was analyzed using the analytical code, EDDYCUFF. In the plasma disruption event, the maximum magnetic pressure caused by eddy currents and poloidal fields was 1.2MPa. The maximum stress intensity by this magnetic pressure was 114MPa. In the VDE, the maximum magnetic pressure was 2.4MPa and the maximum stress intensity was 253MPa. This stress was somewhat beyond the allowable stress limit. Therefore, the blanket structure and support design should be reviewed to reduce the stress to a suitable value. In summary, the dynamic structural characteristics and design issues of the blanket structure have been identified. (orig.)

  19. Welding techniques development of CLAM steel for Test Blanket Module

    International Nuclear Information System (INIS)

    Fabrication techniques for Test Blanket Module (TBM) with CLAM are being under development. Effect of surface preparation on the HIP diffusion bonding joints was studied and good joints with Charpy impact absorbed energy close to that of base metal have been obtained. The mechanical properties test showed that effect of HIP process on the mechanical properties of base metal was little. Uniaxial diffusion bonding experiments were carried out to study the effect of temperature on microstructure and mechanical properties. And preliminary experiments on Electron Beam Welding (EBW), Tungsten Inert Gas (TIG) Welding and Laser Beam Welding (LBW) were performed to find proper welding techniques to assemble the TBM. In addition, the thermal processes assessed with a Gleeble thermal-mechanical machine were carried out as well to assist the fusion welding research.

  20. Blanket and vacuum vessel design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li2O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

  1. Neutronics experiments for DEMO blanket at JAERI/FNS

    Science.gov (United States)

    Sato, S.; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

    2003-07-01

    In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross-sections needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  5. Mirror Advanced Reactor Study engineering overview

    International Nuclear Information System (INIS)

    The Mirror Advanced Reactor Study (MARS) was the first comprehensive conceptual design of a commercial tandem mirror reactor with thermal barriers. The design exploited the inherent attractive features of a tandem mirror: steady state operation, linear central cell, simple high performance blankets, low first wall heat fluxes, natural impurity diversion by the halo plasma, no driven plasma currents or associated disruptions, and direct conversion of the charged particle power lost out the ends. The study introduced new design concepts in high field magnets, neutral beams, ECRH systems, drift pumping, direct conversion, lithium-lead blankets and plant safety. The MARS design would produce 1200 MWsub(e) net and more than 1500 MWsub(e) gross from only 2600 MW of fusion power. This high efficiency is achieved through a combination of blanket design and direct conversion. Special emphasis was placed on fusion's potential for inherent safety, lower activation and simpler disposal of radioactive waste as compared with fission. The blanket has a very low tritium inventory, cannot melt in loss-of-coolant and/or loss-of-flow accidents and can be disposed of as low level waste subject to near-surface burial. MARS would produce busbar electricity at about 7 cents per kilowatthour (constant 1983 dollars). This value is near the upper end of the cost range for new generation capability being installed in the late 1980's. Significant cost reductions can be gained by further improvements in the engineering designs combined with a simplified end cell. The largest cost reductions from engineering can be attained through redesigned magnets, heat transport system and electrical system. The combination of engineering and physics improvements are projected to lower the cost of electricity by about 40% without sacrificing the environmental, safety and maintainability attributes of MARS. This work is now being pursued in the MINIMARS study. (orig.)

  6. Developing a Comprehensive Software Suite for Advanced Reactor Performance and Safety Analysis

    International Nuclear Information System (INIS)

    This paper provides an introduction to the reactor analysis capabilities of the nuclear power reactor simulation tools that are being developed as part of the US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Toolkit. The NEAMS Toolkit is an integrated suite of multiphysics simulation tools that leverage high performance computing to reduce uncertainty in the prediction of the performance and safety of advanced reactor and fuel designs. The toolkit effort is composed of two major components, the fuels product line, which provides tools for fuel performance analysis, and the reactor product line, which provides tools for reactor performance and safety analysis. This paper presents an overview of the NEAMS reactor product line development effort. (author)

  7. Comparative Evaluation of the Loss and Thermal Performance of Advanced Three Level Inverter Topologies

    DEFF Research Database (Denmark)

    Anthon, Alexander; Zhang, Zhe; Andersen, Michael A. E.; Holmes, Grahame; McGrath, Brendan; Teixeira, Carlos

    This paper presents a comparative evaluation of the loss and thermal performance of two advanced three-level inverter topologies, namely the SiC based T-Type and the Hybrid-NPC, both of which are aimed at reducing the high switching losses associated with a conventional Si based T-Type inverter. ...

  8. Validating Student Satisfaction Related to Persistence, Academic Performance, Retention and Career Advancement within ODL Perspectives

    Science.gov (United States)

    Sembiring, Maximus Gorky

    2015-01-01

    Student satisfaction associated with persistence, academic performance, retention, and its relations to career advancement were examined. It was aimed at measuring service quality (Servqual) dimensions as a foundation of satisfaction and how, in what comportments, they were interrelated. The study was conducted under explanatory-design. Data was…

  9. Towards high performance and multi-functional structural membranes using advanced fibrous and textile materials

    OpenAIRE

    Fangueiro, Raúl; Rana, Sohel

    2015-01-01

    Scientific and technological advancements in the area of fibrous and textile materials have greatly enhanced their application potential in several high-end technical and industrial sectors including construction, transportation, medical, sports, aerospace engineering, electronics and so on. Excellent performance accompanied by light-weight, mechanical flexibility, tailor-ability, design flexibility, easy fabrication and relatively lower cost are the driving force...

  10. National Skill Standards for Advanced High Performance Manufacturing. Version 2.1.

    Science.gov (United States)

    National Coalition for Advanced Manufacturing, Washington, DC.

    This document presents and discusses the national skill standards for advanced high-performance manufacturing that were developed during a project that was commissioned by the U.S. Department of Education. The introduction explains the need for national skill standards. Discussed in the next three sections are the following: benefits of national…

  11. Advanced control rooms and crew performance issues: Implications for human reliability

    International Nuclear Information System (INIS)

    Recent trends in advanced control room (ACR) design are considered with respect to their impact on human performance. It is concluded that potentially negative influences exist, however, a variety of factors make it difficult to model, analyze, and quantify these effects for human reliability analyses (HRAs)

  12. MIT LMFBR blanket research project. Final summary report

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J.

    1983-08-01

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

  13. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

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

  14. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

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

  15. Fuel performance analysis of the Korea advanced nuclear fuel using ESCORE

    International Nuclear Information System (INIS)

    The Lead Use Assembly (LUA) of the Korea advanced nuclear fuel will be loaded in YGN4 cycle 7. The core cycle length is 16,248 MWD/MTU and 57,996 MWD/MTU is the maximum pin peak burnup. Fuel thermal and mechanical performance (i.e. maximum temperature, rod internal pressure, and cladding corrosion) evaluation is performed with ESCORE code which is developed by EPRI

  16. Research on the Performance Evaluation and Advancing Path of Henan Agricultural Products E-marketing

    OpenAIRE

    Pan, Jingqiang

    2013-01-01

    Given that the traditional marketing management of agricultural products has been unable to meet the diversified demands of market, and the network has been applied to the marketing field of agricultural products, the research on the performance evaluation and advancing path of Henan agricultural products E-marketing is more important than before. Based on current problems of Henan agricultural products E-marketing, a system of performance evaluation is established, which covers website build...

  17. Neutronics integral experiments of lithium-oxide fusion blanket with heterogeneous configurations using deuterium-tritium neutrons

    International Nuclear Information System (INIS)

    Neutronics experiment for two types of heterogeneous blankets are performed in the Phase-IIC experiments of the Japan Atomic Energy Research Institute/U.S. Department of Energy collaborative program on fusion blanket neutronics. The experimental system uses the same geometry as the previous Phase-IIA series, which was a closed geometry that used a neutron source enclosure of lithium carbonate. The heterogeneities selected for testing are the beryllium edge-on and the water coolant channel assemblies that appear in typical blankets. In the former, the beryllium and the lithium-oxide (Li2O) layers are piled up alternately in the front part of the test blanket. In the latter, the two simulated water cooling channels are emplaced vertically in the Li2O blanket. These channels produce a steep gradient of neutron flux and a significant spectrum change around the material boundary. The calculation accuracy and measurement method for these transient regions are key areas of interest in the experiments. The measurements are performed for the tritium production rate and the other nuclear parameters as well as the previous experiments. The void effect is found to not be negligible around the heterogeneous region for the detector with a low-energy response. At the same time, enhancements of tritium production are seen near the beryllium and hydrogenous material. However, the current Monte Carlo calculation shows good agreement with the experiment even in such a boundary. 22 refs., 20 figs., 7 tabs

  18. Progress in the integration of Test Blanket Systems in ITER equatorial port cells and in the interfaces definition

    International Nuclear Information System (INIS)

    Highlights: ► The design integration of two test blanket systems in ITER port cell is addressed. ► Definition of interfaces of TBSs with building and other ITER systems is done. ► Designs of pipe forest, bioshield plug and ancillary equipment unit are described. ► The maintenance of the two test blanket systems in ITER port cell is considered. ► The management of the heat and tritium releases in the TBM port cell is described. - Abstract: In the framework of the TBM Program, three ITER vacuum vessel equatorial ports (no. 16, no. 18 and no. 02) have been allocated for the testing of up to six mock-ups of six different DEMO tritium breeding blankets. Each one is called a Test Blanket System (TBS). A TBS consists mainly of the Test Blanket Module (TBM), the in-vessel component facing the plasma, and several ancillary systems, in particular the cooling system and the tritium extraction system. Each port accommodates two TBMs and therefore the two TBSs have to share the corresponding port cell. This paper deals with the design integration aspects of the two TBSs in each port cell performed at ITER Organization (IO) with the corresponding definition of interfaces with other ITER systems. The performed activities have raised several issues that are discussed in the paper and for which design solutions are proposed.

  19. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  20. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  1. Lifetime performance analysis of HCPB power plant in-vessel components using HERCULES

    International Nuclear Information System (INIS)

    The PPCS plant models explore a range of physics and technology assumptions spanning from near-term, ITER-like assumptions to very advanced scenarios. Plant model B (PMB) employs a helium-cooled, pebble bed (HCPB) blanket concept. Creep, both thermal and irradiation, is the primary limiting mechanism for in-vessel component lifetime, with the bulk of cumulative damage due to the power-on period. Lifetime improvement is possible if the time-to-failure at the full power-on stress level can be extended, which can be achieved through geometrical shape optimization. Redesign of the in-vessel build could rely on a slimmer tritium generating blanket (∼40 cm) instead of ∼50 cm, with the first wall acting as a 'shielding' blanket ∼10 cm thick. The concept relies on extending the first wall and breeding blanket lifetime significantly, by limiting creep and fatigue, without losing the breeding capability of tritium breeding ratio (TBR) > 1. HERCULES was used for a neutronics analysis and the extraction of engineering parameters. A comparison with previous creep-fatigue analyses of helium-cooled blankets and first walls show that this new concept is capable of substantially increased lifetime, with TBR ∼ 1.0 (original TBR ∼ 1.14). The concept needs further refinement and optimization, but it is promising to deliver not only engineering parameters but also economic performance

  2. Comparison of Performing and Not-Performing the Prophylactic Surgery for Urinary Incontinence in Advanced Pelvic Organ Prolapse

    Directory of Open Access Journals (Sweden)

    Parvin Bastani,

    2014-07-01

    Full Text Available Objectives: To compare the Performing and not-performing of prophylactic surgery for urinary incontinence in women with pelvic organ prolapse. Materials and Methods: In a randomized controlled clinical trial that performed in department of Gynecology and Obstetrics of Tabriz University of medical sciences on women with advanced pelvic organ prolapse without history of urinary incontinence. The effects of performing and not-performing of prophylactic surgery for urinary incontinence in women with advance pelvic organ prolapse evaluated. Results: Mean age of patients was 53.83±5.52 year in the range of 41-66 year. Mean ICIQ-SF questionnaire scores at the first three questions in patients with pelvic organ prolapse with prophylactic surgery was 0.50±1.54 and in patients with pelvic organ prolapse without prophylactic surgery was 0.86±2.02. Significant difference was not found at the answers to the first question (P=0.507, the second question (P=0.415 and to the third question (P=0.445 ICIQ-SF questionnaire of in the patients advance pelvic organ prolapse underwent surgery for pelvic organ prolapse with and without prophylactic surgery. Conclusion: As regards, a significant difference was not found in the prevention of urinary incontinence between groups with and without prophylactic surgery. Therefore, prophylactic surgery is not recommended routinely for all patients and only we will attempt to do this surgery in symptomatic patients and reduced the additional costs and complications associated with prophylactic surgery in other patients who actually did not need to do this surgery.

  3. Exploring Advanced Technology Gas Turbine Engine Design and Performance for the Large Civil Tiltrotor (LCTR)

    Science.gov (United States)

    Snyder, Christopher A.

    2014-01-01

    A Large Civil Tiltrotor (LCTR) conceptual design was developed as part of the NASA Heavy Lift Rotorcraft Systems Investigation in order to establish a consistent basis for evaluating the benefits of advanced technology for large tiltrotors. The concept has since evolved into the second-generation LCTR2, designed to carry 90 passengers for 1,000 nautical miles at 300 knots, with vertical takeoff and landing capability. This paper explores gas turbine component performance and cycle parameters to quantify performance gains possible for additional improvements in component and material performance beyond those identified in previous LCTR2 propulsion studies and to identify additional research areas. The vehicle-level characteristics from this advanced technology generation 2 propulsion architecture will help set performance levels as additional propulsion and power systems are conceived to meet ever-increasing requirements for mobility and comfort, while reducing energy use, cost, noise and emissions. The Large Civil Tiltrotor vehicle and mission will be discussed as a starting point for this effort. A few, relevant engine and component technology studies, including previous LCTR2 engine study results will be summarized to help orient the reader on gas turbine engine architecture, performance and limitations. Study assumptions and methodology used to explore engine design and performance, as well as assess vehicle sizing and mission performance will then be discussed. Individual performance for present and advanced engines, as well as engine performance effects on overall vehicle size and mission fuel usage, will be given. All results will be summarized to facilitate understanding the importance and interaction of various component and system performance on overall vehicle characteristics.

  4. Validating Student Satisfaction Related to Persistence, Academic Performance, Retention and Career Advancement within ODL Perspectives

    Directory of Open Access Journals (Sweden)

    Maximus Gorky Sembiring

    2015-11-01

    Full Text Available Student satisfaction associated with persistence, academic performance, retention, and its relations to career advancement were examined. It was aimed at measuring service quality (Servqual dimensions as a foundation of satisfaction and how, in what comportments, they were interrelated. The study was conducted under explanatory-design. Data was collected proportionally and purposively followed by congregating them through unified interviews. Population was 1,814 Universitas Terbuka students domiciled overseas; 350 questionnaires were dispersed, 169 completed. Satisfaction was assessed by examining Servqual dimensions. Importance-performance analysis (IPA and customer-satisfaction index (CSI were applied to measure satisfaction and the level of its importance. Structural equation model (SEM was then employed to examine influencing variables. Nine hypotheses developed were all validated by the analysis. Responsiveness, assurance, tangible, reliability, and empathy were in harmony to satisfaction. Career advancement, retention, academic performance, and persistence were influenced by satisfaction. Qualitative inquiry implemented afterwards was basically coherent with the quantitative findings.

  5. Nation-State Size, Ethnic Diversity and Economic Performance in the Advanced Capitalist Countries

    DEFF Research Database (Denmark)

    Patsiurkoa, Natalka; Campbell, John L.; Hall, John A.

    2013-01-01

    This paper examines the proposition that the economic performance of advanced capitalist countries depends on their size and ethnic composition. As such it blends insights from two important literatures in comparative political economy. One is exemplified by the work of Peter Katzenstein, who wrote...... the classic treatise on the relationship between nation-state size and economic performance. Another is illustrated by the work of Ernest Gellner, whose work suggested that economic performance depends on the ethnic composition of the nation-state. The argument is tested on pooled data from 30...... advanced capitalist countries for the 1985 through 2007 period. Regression analysis confirms that ethnically homogenous countries tend to have stronger rates of economic growth during this period than ethnically heterogeneous countries but that neither the size of countries nor the interaction of size and...

  6. Advances in multi-physics and high performance computing in support of nuclear reactor power systems modeling and simulation

    International Nuclear Information System (INIS)

    Significant advances in computational performance have occurred over the past two decades, achieved not only by the introduction of more powerful processors but the incorporation of parallelism in computer hardware at all levels. Simultaneous with these hardware and associated system software advances have been advances in modeling physical phenomena and the numerical algorithms to allow their usage in simulation. This paper presents a review of the advances in computer performance, discusses the modeling and simulation capabilities required to address the multi-physics and multi-scale phenomena applicable to a nuclear reactor core simulator, and present examples of relevant physics simulation codes' performances on high performance computers.

  7. Seismic isolation of Advanced LIGO gravitational waves detectors: Review of strategy, instrumentation, and performance

    CERN Document Server

    Matichard, F; Mittleman, R; Mason, K; Kissel, J; McIver, J; Abbott, B; Abbott, R; Abbott, S; Allwine, E; Barnum, S; Birch, J; Biscans, S; Celerier, C; Clark, D; Coyne, D; DeBra, D; DeRosa, R; Evans, M; Foley, S; Fritschel, P; Giaime, J A; Gray, C; Grabeel, G; Hanson, J; Hardham, C; Hillard, M; Hua, W; Kucharczyk, C; Landry, M; Roux, A Le; Lhuillier, V; Macleod, D; Macinnis, M; Mitchell, R; Reilly, B O; Ottaway, D; Paris, H; Pele, A; Puma, M; Radkins, H; Ramet, C; Robinson, M; Ruet, L; Sarin, P; Shoemaker, D; Stein, A; Thomas, J; Vargas, M; Venkateswara, K; Warner, J; Wen, S

    2015-01-01

    Isolating ground-based interferometric gravitational wave observatories from environmental disturbances is one of the great challenges of the advanced detector era. In order to directly observe gravitational waves, the detector components and test masses must be highly inertially decoupled from the ground motion not only to sense the faint strain of space-time induced by gravitational waves, but also to maintain the resonance of the very sensitive 4 km interferometers. This article presents the seismic isolation instrumentation and strategy developed for Advanced LIGO interferometers. It reviews over a decade of research on active isolation in the context of gravitational wave detection, and presents the performance recently achieved with the Advanced LIGO observatory. Lastly, it discusses prospects for future developments in active seismic isolation and the anticipated benefits to astrophysical gravitational wave searches. Beyond gravitational wave research, the goal of this article is to provide detailed is...

  8. The Impact of Time Between Cognitive Tasks on Performance: Evidence from Advanced Placement Exams

    OpenAIRE

    Ian Fillmore; Devin G. Pope

    2012-01-01

    In many education and work environments, economic agents must perform several mental tasks in a short period of time. As with physical fatigue, it is likely that cognitive fatigue can occur and affect performance if a series of mental tasks are scheduled close together. In this paper, we identify the impact of time between cognitive tasks on performance in a particular context: the taking of Advanced Placement (AP) exams by high-school students. We exploit the fact that AP exam dates change f...

  9. Avaliação do desempenho do reator anaeróbio de manta de lodo (uasb em escala laboratorial na remoção da carga orgânica de águas residuárias da suinocultura Performance evaluation of a lab-scale upflow anaerobic sludge blanket reactor (UASB removing organic loading rate from swine manure

    Directory of Open Access Journals (Sweden)

    Cláudio Milton Montenegro Campos

    2005-04-01

    Full Text Available Objetivou-se com o presente trabalho avaliar o desempenho do reator anaeróbio de manta de lodo (UASB-Upflow Anaerobic Sludge Blanket construído em escala laboratorial na redução da carga orgânica poluidora dos despejos suinícolas brutos. O sistema completo foi composto de um tanque de acidificação e equalização, reator UASB e lagoa aerada facultativa. O tempo de detenção hidráulica (TDH e temperatura adotada para o reator UASB foram de 30 horas e 30 ºC, respectivamente. Os valores médios afluentes de DQO T, ST e SVT foram de 1806, 1810 e 1240 mg.L-1. As eficiências de remoção de DQO T, ST e SVT foram de 84, 58 e 73%, respectivamente. O sistema se apresentou-se estável, com boas condições de tamponamento, retenção e digestibilidade de sólidos, demonstrando que os critérios adotados foram adequados, principalmente aqueles referentes ao TDH, carga orgânica volumétrica (COV e temperatura.The present work was carried out in order to evaluate the performance of a lab scale Upflow Anaerobic Sludge Blanket reactor (UASB treating liquid effluent from swine manure without solids separation. The treatment system consisted of one acidification tank, which also equalized the substrate, an UASB reactor, and an aerated facultative pound. The hydraulic retention time (HRT and temperature adopted for the UASB reactor were 30h and 30ºC, respectively. The influent average values of Chemical Oxygen Demand (COD, Total Solids (TS and Total Volatile Solids (TVS were 1806, 1810 and 1240 mg.L-1. The removal efficiencies were 84, 58 and 73 %, respectively. The system presented good stability and buffering conditions, and also a good solids digestibility, showing that the research criteria adopted was adequate, mainly those parameters referred to the HRT, Volumetric Organic Loading Rate (VOLR and temperature.

  10. Advanced SFR Concept Design Studies at KAERI

    International Nuclear Information System (INIS)

    Advanced SFR design concepts have been developed which satisfy the Gen IV technology goals at KAERI. Two types of reactor core were developed for breakeven and TRU burner and both cores do not have blankets to enhance proliferation resistance. The Advanced SFR is a pool-type reactor that improves system safety through slow system transients. The heat transport system adopts two double wall tube Steam Generators and a passive Residual Heat Removal System PDRC. To secure the economic competitiveness of an SFR, the diameter of the reactor vessel of the Advanced SFR is designed to be 14.5 m, which is a very compact size compared to other designs. Also, various R and D activities have been performed in order to prepare some analysis tools and to support the development of design concepts. (author)

  11. Manufacture of blanket shield modules for ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-11-15

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

  12. Efficient COD removal and nitrification in an upflow microaerobic sludge blanket reactor for domestic wastewater.

    Science.gov (United States)

    Zheng, Shaokui; Cui, Cancan

    2012-03-01

    The treatment performance of an upflow microaerobic sludge blanket reactor (UMSB) for synthetic domestic wastewater was investigated at two dissolved oxygen (DO) levels, 0.3-0.5 and 0.7-0.9 mg l(-1), focusing on nitrification performance. The higher DO level induced complete nitrification of ammonia nitrogen (NH(3)-N), achieving chemical oxygen demand and NH(3)-N removals of 97 and 92%, respectively. There were consistently significantly higher nitrate nitrogen (NO(3)-N) and nitrite nitrogen (NO(2)-N) levels in the effluent, with ~66% of newly-produced oxidised nitrogen as NO(2)-N. Despite the high nitrification efficiency, only about 23% of the removed NH(3)-N amount from the influent was ultimately transformed into oxidised nitrogen due to the simultaneous nitrification-denitrification. Sludge blanket development and granulation occurred simultaneously in the UMSB. PMID:22105554

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

    Science.gov (United States)

    Scialdone, John J.

    1990-01-01

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

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

    Science.gov (United States)

    Dillard, P. A.

    1983-01-01

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

  15. MHD pressure drop at bare welding positions in pipes of DCLL blankets (KIT Scientific Reports ; 7636)

    OpenAIRE

    Bühler, Leo

    2013-01-01

    A systematic parametric analysis has been performed using asymptotic numerical methods for determination of MHD flows near gaps of electrically insulating inserts in well conducting pipes. Such gaps could be present at several positions in fusion blankets, where cutting and rewelding by remotely controlled tools is foreseen. Gaps in the insulation provide additional current paths which leads to increased current density and braking electromagnetic Lorentz forces.

  16. Tidal Volume Estimation Using the Blanket Fractal Dimension of the Tracheal Sounds Acquired by Smartphone

    OpenAIRE

    Natasa Reljin; Reyes, Bersain A.; Chon, Ki H

    2015-01-01

    In this paper, we propose the use of blanket fractal dimension (BFD) to estimate the tidal volume from smartphone-acquired tracheal sounds. We collected tracheal sounds with a Samsung Galaxy S4 smartphone, from five (N = 5) healthy volunteers. Each volunteer performed the experiment six times; first to obtain linear and exponential fitting models, and then to fit new data onto the existing models. Thus, the total number of recordings was 30. The estimated volumes were compared to the true val...

  17. Preliminary design of test facilities for tritium breeding blanket development, (1)

    International Nuclear Information System (INIS)

    This report describes the results of the preliminary design of outpile test facilities which are used for development of tritium breeding blanket with ceramic breeding material. The facilities which were designed are as follows; High heat flux test facility, Thermal-hydraulic test facility, Integrity test facility, Fabrication Technology Development Facility. This design study was performed by Kawasaki Heavy Industries, Ltd. under the contract to Fusion Research System Laboratory. (author)

  18. Advanced water processing system (AWPS), including advanced filtration system (AFS) and advanced ion selective system (AISS) for improved utility (PWR/BWR) water processing performance

    International Nuclear Information System (INIS)

    The advanced water processing system (AWPS) has the potential for wide spread success on a worldwide scale in both PWR and BWRs. The AWPS incorporates the advanced features (patent pending) of advanced filtration and advanced ion selective technologies (patented). Typical problems encountered in current filtration systems include: (1) poor effluent quality, (2) short run lengths on filters, (3) frequent filter change-outs/backwashes, (4) large waste volumes, and (5) failed filter cartridges. The advanced filtration system (AFS) features reduced waste production per million gallons of water processed, cleaner water for recycle or release to the environment, filter element volume 100 times less than that of competitive filters, and a far lower capital cost compared to systems with similar performance. The AWPS should be of interest to plants that are upgrading, or to new plants to lower both their capital and operating costs, as well as total curie discharge levels. In addition, the AWPS will function in non-nuclear, as well as nuclear, applications of water purification, specially where pre coat filtration/ion exchange or reverse osmosis (RO) is being applied to process water with high concentrations of colloidal contaminants. Pilot testing has been successfully completed in the U. S. at the Byron (PWR), LaSalle (BWR), and Dresden(BWR) nuclear plants for Commonwealth Edison, and the Bruce several spent filters in a High Integrated Container these bench- and pilot-scale demonstrations will be presented herein. Full-scale designs or systems have been shipped to these locations. In all cases, the testing demonstrated: (1) longer run lengths (300,000 gallons between backwashes--a 100 fold improvement), (2) recoverability of cartridge filters after backwash (cartridge lives of approximately 6 months to a year--a 5 to 10 fold improvement in filter life), (3) large removal efficiencies for colloidal particles (reduced discharge curies), and (4) reduced waste volumes

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

    Science.gov (United States)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Performance and economics of advanced energy conversion systems for coal and coal-derived fuels

    Science.gov (United States)

    Corman, J. C.; Fox, G. R.

    1978-01-01

    The desire to establish an efficient Energy Conversion System to utilize the fossil fuel of the future - coal - has produced many candidate systems. A comparative technical/economic evaluation was performed on the seven most attractive advanced energy conversion systems. The evaluation maintains a cycle-to-cycle consistency in both performance and economic projections. The technical information base can be employed to make program decisions regarding the most attractive concept. A reference steam power plant was analyzed to the same detail and, under the same ground rules, was used as a comparison base. The power plants were all designed to utilize coal or coal-derived fuels and were targeted to meet an environmental standard. The systems evaluated were two advanced steam systems, a potassium topping cycle, a closed cycle helium system, two open cycle gas turbine combined cycles, and an open cycle MHD system.

  3. Research on the Performance Evaluation and Advancing Path of Henan Agricultural Products E-marketing

    Institute of Scientific and Technical Information of China (English)

    Jingqiang; PAN

    2013-01-01

    Given that the traditional marketing management of agricultural products has been unable to meet the diversified demands of market, and the network has been applied to the marketing field of agricultural products,the research on the performance evaluation and advancing path of Henan agricultural products E-marketing is more important than before. Based on current problems of Henan agricultural products E-marketing,a system of performance evaluation is established,which covers website building and maintenance and other four first-class indexes as well as profit ratio of sales and other thirteen second-class indexes,and then the system is evaluated by comprehensive evaluation method. Meanwhile,it is explored to discuss how to further advance Henan agricultural products E-marketing.

  4. Hybrid reactor blankets for constant energy multiplication and flat power distribution

    International Nuclear Information System (INIS)

    Two blanket design difficulties are usually attributed to the blanket neutronic properties: high peak-to-average power density ratio distribution and the variation of the energy multiplication with burnup. This work shows that blankets can be designed to have a constant energy multiplication and a flat power distribution. These features are illustrated for light water hydrid reactor blankets

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

    Science.gov (United States)

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

    2015-09-01

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

  6. Development and Performance of an Advanced Ejector Cooling System for a Sustainable Built Environment

    OpenAIRE

    Pereira, Paulo R.; Varga, Szabolcs; Oliveira, Armando Carlos; Soares, João

    2015-01-01

    Ejector refrigeration is a promising technology for the integration into solar driven cooling systems because of its relative simplicity and low initial cost. The major drawback of such a system is associated with its relatively low coefficient of performance (COP) under variable operating conditions. In order to overcome this problem, an advanced ejector was developed that changes its geometrical features depending on the upstream and downstream conditions. This paper provides a short overvi...

  7. Advanced Control Scenario of High-Performance Steady-State Operation for JT-60 Superconducting Tokamak

    Institute of Scientific and Technical Information of China (English)

    H. Tamai; Y. Kamada; A. Sakasai; S. Ishida; G. Kurita; M. Matsukawa; K. Urata; S. Sakurai; K. Tsuchiya; A. Morioka; Y. M. Miura; K. Kizu

    2004-01-01

    Plasma control on high-βN steady-state operation for JT-60 superconducting modification is discussed. Accessibility to high-βN exceeding the free-boundary limit is investigated with the stabilising wall of reduced-activated ferritic steel and the active feedback control of the in-vessel non-axisymmetric field coils. Taking the merit of superconducting magnet, advanced plasma control for steady-state high performance operation could be expected.

  8. A Wake Model for the Prediction of Propeller Performance at Low Advance Ratios

    Directory of Open Access Journals (Sweden)

    Ye Tian

    2012-01-01

    Full Text Available A low order panel method is used to predict the performance of propellers. A wake alignment model based on a pseudounsteady scheme is proposed and implemented. The results from this full wake alignment (FWA model are correlated with available experimental data, and results from RANS for some propellers at design and low advance ratios. Significant improvements have been found in the predicted integrated forces and pressure distributions.

  9. Open Innovation at NASA: A New Business Model for Advancing Human Health and Performance Innovations

    Science.gov (United States)

    Davis, Jeffrey R.; Richard, Elizabeth E.; Keeton, Kathryn E.

    2014-01-01

    This paper describes a new business model for advancing NASA human health and performance innovations and demonstrates how open innovation shaped its development. A 45 percent research and technology development budget reduction drove formulation of a strategic plan grounded in collaboration. We describe the strategy execution, including adoption and results of open innovation initiatives, the challenges of cultural change, and the development of virtual centers and a knowledge management tool to educate and engage the workforce and promote cultural change.

  10. Experience of neutronic evaluation for in-pile tests of fusion blanket with the JMTR. Influence of impurities in Beryllium

    International Nuclear Information System (INIS)

    Impurities in beryllium for fusion blanket application such as armor of the first wall and neutron multiplier is one of important issues to perform its properties. In this study, tritium production rate evaluation was carried out on the base of JMTR irradiation test that simulated the breeding blanket and the impact of impurity on tritium production rate was discussed. Influence of impurities in S200F beryllium was therefore less than 1%, concerning tritium production rate. Accordingly, it is reasonable to estimate extremely small impact on tritium production rate in case of fusion application that uses high purity beryllium such as S65C. (author)

  11. Thermal-hydraulics deisgn and analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    A conceptual design of fusion-fission hybrid reactor for energy production, named FDS-EM (Energy Multiplier), was proposed. It was preliminary designed to generate about an electricity power of about 1.0 GW with self-sustaining tritium cycle. This contribution performed the thermal-hydraulics design and analyses for FDS-EM water-cooled blanket. The typical thermal-hydraulics parameters were designed by using mature technologies of PWR, and temperature and stress analyses were carried out, according to typical parameters of the blanket. The results preliminarily demonstrated the engineering feasibility of the design. (authors)

  12. Convertible liquid metal blankets for ITER with Pb-17Li as breeding material

    International Nuclear Information System (INIS)

    A convertible blanket concept is proposed for ITER, where, without replacement of the blanket structure, a non-breeding Pb alloy is used during the basic performance phase and the eutectic Pb-17Li during the enhanced performance phase. The concept is based on austenitic steel as structural material, an average neutron wall load of 1MWm-2 and either helium or water as coolant. The same design concept was used for both coolant options with respect to a stiff blanket segment box, direct cooling of the first wall using toroidal ducts, poloidal hairpin tubes to cool the quasi-stagnant liquid metal and tritium removal outside the vacuum vessel.Various design options were considered for the first-wall and pool cooling and corresponding headers. Owing to the different coolant properties, different combinations were selected for the two versions. The performance of the two versions was assessed among other things with respect to tritium breeding and control, reliability and R and D needs. (orig.)

  13. Preliminary three-dimensional neutronics design and analysis of helium-cooled blanket for a multi-functional experimental fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    A multi-functional experimental fusion-fission hybrid reactor concept named FDS-MFX, which is based on viable fusion and fission technologies, has been proposed. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this paper, the design optimization for the layout and the size of high enriched uranium modules in later stage of uranium-fueled blanket has been performed. Finally, proposing a preliminary three-dimension neutronics design with maximum average Power Density (PDmax) 100 MW/ m3, loaded mass of the 235U 1000 kg and TBR (Tritium Breeding Ratio) 1.05. (authors)

  14. Front-End Electronics for Verification Measurements: Performance Evaluation and Viability of Advanced Tamper Indicating Measures

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) continues to expand its use of unattended, remotely monitored measurement systems. An increasing number of systems and an expanding family of instruments create challenges in terms of deployment efficiency and the implementation of data authentication measures. A collaboration between Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL), and Los Alamos National Laboratory (LANL) is working to advance the IAEA's capabilities in these areas. The first objective of the project is to perform a comprehensive evaluation of a prototype front-end electronics package, as specified by the IAEA and procured from a commercial vendor. This evaluation begins with an assessment against the IAEA's original technical specifications and expands to consider the strengths and limitations over a broad range of important parameters that include: sensor types, cable types, and the spectrum of industrial electromagnetic noise that can degrade signals from remotely located detectors. A second objective of the collaboration is to explore advanced tamper-indicating (TI) measures that could help to address some of the long-standing data authentication challenges with IAEA's unattended systems. The collaboration has defined high-priority tampering scenarios to consider (e.g., replacement of sensor, intrusion into cable), and drafted preliminary requirements for advanced TI measures. The collaborators are performing independent TI investigations of different candidate approaches: active time-domain reflectometry (PNNL), passive noise analysis (INL), and pulse-by-pulse analysis and correction (LANL). The initial investigations focus on scenarios where new TI measures are retrofitted into existing IAEA UMS deployments; subsequent work will consider the integration of advanced TI methods into new IAEA UMS deployments where the detector is separated from the front-end electronics. In this paper, project progress

  15. Neutronics Experiment on A HCPB Breeder Blanket Mock-Up

    International Nuclear Information System (INIS)

    A neutronics experiment has been performed in the frame of European Fusion Technology Program on a mock-up of the EU Test Blanket Module (TBM), Helium Cooled Pebble Bed (HCPB) concept, with the objective to validate the capability of nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. The experiment has been carried out at the FNG 14-MeV neutron source in collaboration between ENEA, Technische Universitaet Dresden, Forschungszentrum Karlsruhe, J. Stefan Institute Ljubljana and with the participation of JAEA. The mock-up, designed in such a way to replicate all relevant nuclear features of the TBM-HCPB, consisted of a steel box containing beryllium block and two intermediate steel cassettes, filled with of Li2CO3 powder, replicating the breeder insert main characteristics: radial thickness, distance between ceramic layers, thickness of ceramic layers and of steel walls. In the experiment, the TPR has been measured using Li2CO3 pellets at various depths at two symmetrical positions at each depth, one in the upper and one in the lower cassette. Twelve pellets were used at each position to determine the TPR profile through the cassette. Three independent measurements were performed by ENEA, TUD/VKTA and JAEA. The neutron flux in the beryllium layer was measured as well using activation foils. The measured tritium production in the TBM (E) was compared with the same quantity (C) calculated by the MCNP.4c using a very detailed model of the experimental set up, and using neutron cross sections from the European Fusion File (EFF ver.3.1) and from the Fusion Evaluated Nuclear Data Library (FENDL ver. 2.1, ITER reference neutron library). C/E ratios were obtained with a total uncertainty on the C/E comparison less than 9% (2 s). A sensitivity and uncertainty analysis has also been performed to evaluate the calculation uncertainty due to the uncertainty on neutron cross sections. The results of such analysis

  16. US DCLL test blanket module design and relevance to DEMO

    International Nuclear Information System (INIS)

    Full text: In the design of Test Blanket Modules (TBMs) for ITER, it is required to provide a design concept that is demonstration power reactor (DEMO) relevant. It should be noted that in the US, DEMO is defined to be a good representation of the first generation fusion power reactor. In order to evaluate the potential of the US TBM design for DEMO, a system evaluation of DEMO design was performed with an improved GA system code, and the physics results were benchmarked to ITER. With the selection of ferritic steel as the structural material, the maximum neutron wall loading is limited to 3 MW/m2. When designed to a 3 GW fusion device the optimum aspect ratio is found to be in the range of 2.5 to 3. Results show that the US dual coolant lead-lithium (DCLL) blanket can satisfy all the DEMO design requirements. On the chamber wall material, for the ITER-TBM design, the design guidance is to apply a 2 mm Be layer onto the plasma facing surface. When extrapolated to the DEMO design, the Be layer will not be suitable due to radiation damage. Similarly, a carbon surface will not be suitable due to high physical and chemical sputtering rates, radiation damage of the material and potential large retention of tritium. Unfortunately, the remaining commonly proposed material, tungsten (W), would suffer radiation damage from alpha particle implantation and, with blistering, W transport to the plasma core could severely limit the core performance. To resolve this potential impasse, different innovative options were evaluated. All high performance tokamak experiments presently use boron or silicon to condition the first wall. To use boron in DEMO, it is found that in-situ boronization will be required in order to maintain a boronized layer on the chamber wall. This boronized layer could also protect the W substrate, while retaining low-Z wall characteristics. Further innovative ideas are being evaluated to handle transient events like ELMs and disruptions. TOPICS: (PPCA) Power

  17. 18 CFR 284.402 - Blanket marketing certificates.

    Science.gov (United States)

    2010-04-01

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

  18. Safety and personnel access aspects of low activation fusion blankets

    International Nuclear Information System (INIS)

    The use of silicon carbide and carbon materials for structural applications in fusion reactor first wall and blanket regions has been proposed and a continuing effort spent on the development of the ceramics technology. The advantages identified are an extremely low induced radioactivity inventory, a high temperature operating capability, abundant raw material resource availability, and minimized plasma impurity effects. One of the unique features of the applications of these materials to fusion reactor blanket designs is that no alloying element is needed in order to assure the specified mechanical properties such as occurs in metal alloys. The major source of long term radioactivity in these materials is impurities. The impurity elements and their concentrations carried over to the blanket structure during fabrication can be minimized by proper fabrication procedures and techniques. The safety and personnel access aspects of such fusion blankets in conjunction with the impurity element concentration are the main subjects of this paper

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

    Data.gov (United States)

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

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

    International Nuclear Information System (INIS)

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

  1. An analysis of electron beam welds in a dual coolant liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Numerical simulation of electron beam welding of blanket segments was performed using non-linear finite element code ABAQUS. The thermal and stress fields were assumed uncoupled, while preserving the temperature dependency of all material parameters. The martensite-austenite and austenite-martensite transformations were taken into account through volume shrinking/expansion effects, which is consistent with available data. The distributions of post welding residual stress in a complex geometry of the first wall are obtained. Also, the effects of preheating and post-welding heat treatment were addressed. Time dependent temperature and stress-strain fields obtained provide good insight into the welding process. They may be used directly to support reliability and life-time studies of blanket structures. On the other hand, they provide useful hints about the feasibility of the geometrical configurations as proposed by different design concepts. (orig.)

  2. Benchmark experiment on the model of fusion reactor blanket with uranium neutron multiplier

    International Nuclear Information System (INIS)

    Benchmark experiment on the model of thermonuclear reactor blanket with 14 MeV neutron source is described. The model design corresponds to the known concept of the fast hybrid blanket with 238U neutron multiplier and main tritium production on 6Li. Detailed measurements of the following process velocities were carried out: tritium production on lithium isotopes; reactions modelling tritium production; (n, γ) and (n, 2n) processes for 238U; fission reactions for 235,238U, 239Pu, 237Np. Neutron flux integral measurements were performed by a set of threshold detectors on the basic of the 115In(n, n'), 204Pb(n, n'), 64Zn(n, p), 27Al(n, p), 56Fe(n, p), 107Ag(n, 2n), 63Cu(n, 2n) and 64(n, 2n) reactions

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

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

    International Nuclear Information System (INIS)

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li2TiO3 or Li2O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for the energy

  7. Four loss-of-flow accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    This report presents the thermal-hydraulic analysis of four Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the alternative SEAFP reactor design. The LOFAs considered result from a loss of electrical power for the recirculation pump in the primary cooling circuit. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analyses, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall and blanket. For the LOFA without plasma shutdown, significant loss of heat removal due to dryout occurs at the midplane of the outboard first wall cooling pipes about 41 s after pump trip. For the three LOFA cases with emergency plasma shutdown that have been studied, the temperature increase in the Be-coating at the midplane of the outboard first wall is limited to about 30 K. (orig.)

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

    International Nuclear Information System (INIS)

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

  9. Numerical Analyses of Electromagnetic Forces on the ITER Blanket Module Shield Block During Major Disruptions

    International Nuclear Information System (INIS)

    Electromagnetic (EM) load is one of the key design drivers for the blanket shield block (SB) and other in-vessel components. In this article, an EM analysis method was developed to address the EM force on the SB. The plasma currents, which vary spatially and temporally, are loaded as a filament at each time point. The standard blanket module No.04 (BM04) under major disruption (MD) is selected to perform the analyses. The analyses results are validated by comparing currents on the passive structure. To better understand the effects of cooling channels and slits on the EM force, the case of SB without cooling channel and the case without slits are calculated to make comparisons. The results show that the slits play an important role in controlling the EM load on SB. (fusion engineering)

  10. Performance Analysis of enhanced Inter-cell Interference Coordination in LTE-Advanced Heterogeneous Networks

    DEFF Research Database (Denmark)

    Wang, Yuanye; Pedersen, Klaus I.

    2012-01-01

    The performance of enhanced Inter-Cell Interference Coordination (eICIC) for Long Term Evolution (LTE)- Advanced with co-channel deployment of both macro and pico is analyzed. The use of pico-cell Range Extension (RE) and time domain eICIC (TDM muting) is combined. The performance is evaluated in...... the downlink by means of extensive system level simulations that follow the 3GPP guidelines. The overall network performance is analyzed for different number of pico-eNBs, transmit power levels, User Equipment (UE) distributions, and packet schedulers. Recommended settings of the RE offset and TDM...... muting ratio in different scenarios are identified. The presented performance results and findings can serve as input to guidelines for co-channel deployment of macro and pico-eNBs with eICIC....

  11. Performance and life evaluation of advanced battery technologies for electric vehicle applications

    Science.gov (United States)

    Deluca, W. H.; Gillie, K. R.; Kulaga, J. E.; Smaga, J. A.; Tummillo, A. F.; Webster, C. E.

    Advanced battery technology evaluations are performed under simulated electric vehicle (EV) operating conditions at the Argonne Analysis and Diagnostic Laboratory (ADL). The ADL provides a common basis for both performance characterization and life evaluation with unbiased application of tests and analyses. This paper summarizes the performance characterizations and life evaluations conducted in 1990 on nine single cells and fifteen 3- to 360-cell modules that encompass six technologies: (Na/S, Zn/Br, Ni/Fe, Ni/Cd, Ni-metal hydride, and lead-acid). These evaluations were performed for the Department of Energy and Electric Power Research Institute. The results provide battery users, developers, and program managers an interim measure of the progress being made in battery R and D programs, a comparison of battery technologies, and a source of basic data for modelling and continuing R and D.

  12. Importance of performance status for treatment outcome in advanced pancreatic cancer

    Institute of Scientific and Technical Information of China (English)

    Stefan Boeck; Axel Hinke; Ralf Wilkowski; Volker Heinemann

    2007-01-01

    Despite progress in the treatment of advanced and metastatic pancreatic cancer (PC), the outcome of this disease remains dismal for the majority of patients.Given the moderate efficacy of treatment, prognostic factors may help to guide treatment decisions. Several trials identified baseline performance status as an important prognostic factor for survival. Unfit patients with a Karnofsky performance status (KPS) below 70% only have a marginal benefit from chemotherapy with gemcitabine (Gem) and may often benefit more from optimal supportive care. Once, however, the decision is taken to apply chemotherapy, KPS may be used to select either mono- or combination chemotherapy. Patients with a good performance status (KPS = 90%-100%) may have a significant and clinically relevant survival benefit from combination chemotherapy. By contrast, patients with a poor performance status (KPS ≤ 80%) have no advantage from intensified therapy and should rather receive single-agent treatment.

  13. Interim report on the performance test results for the Advanced Verification for Inventory sample System (AVIS)

    International Nuclear Information System (INIS)

    The AVIS (Advanced Verification for Inventory sample System) is a high accuracy nondestructive assay (NDA) system for Safeguards to measure the plutonium mass for plutonium uranium mixed oxide (MOX) powder and pellet for large scale MOX fuel fabrication plant (J-MOX). It is intended that the AVIS measurement will be substituted for a fraction of the DA samples to reduce the number of DA from J-MOX. Therefore, the AVIS has a crucial role to attain effective safeguards for J-MOX. JAEA, which has experience/knowledge to develop the NDA system and plutonium handling field, was entrusted with performance testing for the AVIS from NMCC. JAEA has conducted performance testing on the AVIS using standard radiation source (neutron and gamma). As the results of test, it was confirmed that the AVIS has the designed performance. This paper reports the performance test results and evaluation results of measurement accuracy of the AVIS. (author)

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

    International Nuclear Information System (INIS)

    of fabrication steps. Based on the results of study for optimization of the simultaneous HIP bonding conditions, the HIP conditions were 1050degC, 150 MPa and holding time of 2 hours. Before this assembly for the HIP process, a deep drilling was performed for the coolant channels of the shield block from both sides of the block, then the shield block was bent by 10000-ton press machine to provide the specified curvature. During the bending, iced water was inserted into the drilled holes to prevent excessive deformation of the holes. Iced water was applied as the inserted material in this study because it was easy to remove the inserted material from the drilled holes and chemical reaction could be prevented during removal of the inserted materials. After the HIP process, the first wall surface was finally machined. The back part of the module was also machined to provide coolant manifolds, then cover plates of the manifolds were welded by TIG welding. A series of measurements and inspections was performed in the course of fabrication to make sure the dimensional accuracy and integrity of pressure boundaries. A destructive inspection was also performed with a cut specimen from the edge of the fabricated module to examine the bondability of HIPed interfaces. As a result of this fabrication experience, sufficient bonding by the single step solid HIP process has been demonstrated, and sufficient technical data base on the fabrication of the ITER shielding blanket module has been obtained. (author)

  15. Energy production from distillery wastewater using single and double-phase upflow anaerobic sludge blanket (UASB) reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muyodi, F.J.; Rubindamayugi, M.S.T. [Univ. of Dar es Salaam, Applied Microbiology Unit (Tanzania, United Republic of)

    1997-12-31

    A Single-phase (SP) and Double-phase (DP) Upflow Anaerobic Sludge Blanket (UASB) reactors treating distillery wastewater were operated in parallel. The DP UASB reactor showed better performance than the SP UASB reactor in terms of maximum methane production rate, methane content and Chemical Oxygen Demand (COD) removal efficiency. (au) 20 refs.

  16. Treatment of domestic wastewater in an up-flow anaerobic sludge blanket reactor followed by moving bed biofilm reactor

    NARCIS (Netherlands)

    Tawfik, A.; El-Gohary, F.; Temmink, B.G.

    2010-01-01

    The performance of a laboratory-scale sewage treatment system composed of an up-flow anaerobic sludge blanket (UASB) reactor and a moving bed biofilm reactor (MBBR) at a temperature of (22-35 A degrees C) was evaluated. The entire treatment system was operated at different hydraulic retention times

  17. Impact of prescribed burning on blanket peat hydrology

    OpenAIRE

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

    2015-01-01

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

  18. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. FIRST STEP blanket structure and fuel assembly design

    International Nuclear Information System (INIS)

    FIRST STEP (Fusion, Inertial, Reduced Requirement Systems Test for Special Nuclear Material, Tritium, and Energy Production) is an Inertial Confinement Fusion (ICF) plant designed to produce tritium, SNM, and energy using near-term technology. It is an integrated facility that will serve as a test bed for fusion power plant technology. The design of the blanket structure and blanket fuel assembly for wetted-wall FIRST STEP reactors is presented here

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

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji [ed.

    1998-03-01

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

  3. Requirements for a helium-cooled blanket heat removal system development facility for fusion reactor research

    International Nuclear Information System (INIS)

    Existing and potential design problems associated with the helium-cooled blanket assemblies of experimental, demonstration and hybrid reactor designs considered in the Magnetic Fusion Energy (MFE) Program were assessed. It was observed that a balanced program of design, analysis and experimentation would be required to develop, verify and qualify these designs and those of related hardware and equipment. To respond to the potential experimental requirements of the first-generation reactors (the EPRs and possibly the hybrid concept), the need for a helium test facility was identified. It was determined that this facility should have the capacity for recirculating 100,000 kg/hr of helium at 70 atm and 6000C and should have 3 MW of electrical power available for simulating neutron heating. No radioactive material or processes should be used to facilitate ''hands-on'' experimentation and development. The general types of testing anticipated in this facility would include: (1) thermal and coolant flow performance of the blanket and other components in the primary cooling circuit; (2) structural adequacy of the blanket and first wall including vibration considerations; (3) capability for accommodating safety/off-normal conditions. Existing facilities worldwide were surveyed. It was determined that a number of facilities exist in foreign nations for performing the anticipated experiments. However, no large helium gas flow loop exists within the USA. Consequently, it is recommended that a helium thermal-hydraulic blanket test facility be planned and build on a schedule that will meet the unique design development and verification needs of the fusion program. This report provides the rationale and preliminary scoping of the operational characteristics and requirements for such a facility

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

    International Nuclear Information System (INIS)

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

  5. Thermal hydraulics and mechanics research on fusion blanket system

    International Nuclear Information System (INIS)

    In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding / neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding / n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R and Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics. (author)

  6. Carotid artery stenosis: Performance of advanced vessel analysis software in evaluating CTA

    International Nuclear Information System (INIS)

    Objectives: The aim of this study was to evaluate time efficiency and diagnostic reproducibility of an advanced vessel analysis software for diagnosis of carotid artery stenosis. Material and methods: 40 patients with suspected carotid artery stenosis received head and neck DE-CTA as part of their pre-interventional workup. Acquired data were evaluated by 2 independent radiologists. Stenosis grading was performed by MPR eyeballing with freely adjustable MPRs and with a preliminary prototype of the meanwhile available client-server and advanced visualization software syngo.via CT Vascular (Siemens Healthcare, Erlangen, Germany). Stenoses were graded according to the following 5 categories: I: 0%, II: 1–50%, III: 51–69%, IV: 70–99% and V: total occlusion. Furthermore, time to diagnosis for each carotid artery was recorded. Results: Both readers achieved very good specificity values and good respectively very good sensitivity values without significant differences between both reading methods. Furthermore, there was a very good correlation between both readers for both reading methods without significant differences (kappa value: standard image interpretation k = 0.809; advanced vessel analysis software k = 0.863). Using advanced vessel analysis software resulted in a significant time saving (p < 0.0001) for both readers. Time to diagnosis could be decreased by approximately 55%. Conclusions: Advanced vessel analysis application CT Vascular of the new imaging software syngo.via (Siemens Healthcare, Forchheim, Germany) provides a high rate of reproducibility in assessment of carotid artery stenosis. Furthermore a significant time saving in comparison to standard image interpretation is achievable

  7. Application and Analysis of Performance of DQPSK Advanced Modulation Format in Spectral Amplitude Coding OCDMA

    Directory of Open Access Journals (Sweden)

    Abdul Latif Memon

    2014-04-01

    Full Text Available SAC (Spectral Amplitude Coding is a technique of OCDMA (Optical Code Division Multiple Access to encode and decode data bits by utilizing spectral components of the broadband source. Usually OOK (ON-Off-Keying modulation format is used in this encoding scheme. To make SAC OCDMA network spectrally efficient, advanced modulation format of DQPSK (Differential Quaternary Phase Shift Keying is applied, simulated and analyzed. m-sequence code is encoded in the simulated setup. Performance regarding various lengths of m-sequence code is also analyzed and displayed in the pictorial form. The results of the simulation are evaluated with the help of electrical constellation diagram, eye diagram and bit error rate graph. All the graphs indicate better transmission quality in case of advanced modulation format of DQPSK used in SAC OCDMA network as compared with OOK

  8. Application and analysis of performance of dqpsk advanced modulation format in spectral amplitude coding ocdma

    International Nuclear Information System (INIS)

    SAC (Spectral Amplitude Coding) is a technique of OCDMA (Optical Code Division Multiple Access) to encode and decode data bits by utilizing spectral components of the broadband source. Usually OOK (ON-Off-Keying) modulation format is used in this encoding scheme. To make SAC OCDMA network spectrally efficient, advanced modulation format of DQPSK (Differential Quaternary Phase Shift Keying) is applied, simulated and analyzed, m-sequence code is encoded in the simulated setup. Performance regarding various lengths of m-sequence code is also analyzed and displayed in the pictorial form. The results of the simulation are evaluated with the help of electrical constellation diagram, eye diagram and bit error rate graph. All the graphs indicate better transmission quality in case of advanced modulation format of DQPSK used in SAC OCDMA network as compared with OOK. (author)

  9. Assessment of advanced technologies for high performance single-engine business airplanes

    Science.gov (United States)

    Kohlman, D. L.; Holmes, B. J.

    1982-01-01

    The prospects for significantly increasing the fuel efficiency and mission capability of single engine business aircraft through the incorporation of advanced propulsion, aerodynamics and materials technologies are explored. It is found that turbine engines cannot match the fuel economy of the heavier rotary, diesel and advanced spark reciprocating engines. The rotary engine yields the lightest and smallest aircraft for a given mission requirement, and also offers greater simplicity and a multifuel capability. Great promise is also seen in the use of composite material primary structures in conjunction with laminar flow wing surfaces, a pusher propeller and conventional wing-tail configuration. This study was conducted with the General Aviation Synthesis Program, which can furnish the most accurate mission performance calculations yet obtained.

  10. Recent Advances in Iron Metabolism: Relevance for Health, Exercise, and Performance.

    Science.gov (United States)

    Buratti, Paolo; Gammella, Elena; Rybinska, Ilona; Cairo, Gaetano; Recalcati, Stefania

    2015-08-01

    Iron is necessary for physiological processes essential for athletic performance, such as oxygen transport, energy production, and cell division. However, an excess of "free" iron is toxic because it produces reactive hydroxyl radicals that damage biological molecules, thus leading to cell and tissue injury. Therefore, iron homeostasis is strictly regulated; and in recent years, there have been important advancements in our knowledge of the underlying processes. Hepcidin is the central regulator of systemic iron homeostasis and exerts its function by controlling the presence of the iron exporter ferroportin on the cell membrane. Hepcidin binding induces ferroportin degradation, thus leading to cellular iron retention and decreased levels of circulating iron. As iron is required for hemoglobin synthesis, the tight link between erythropoiesis and iron metabolism is particularly relevant to sports physiology. The iron needed for hemoglobin synthesis is ensured by inhibiting hepcidin to increase ferroportin activity and iron availability and hence to make certain that efficient blood oxygen transport occurs for aerobic exercise. However, hepcidin expression is also affected by exercise-associated conditions, such as iron deficiency, anemia or hypoxia, and, particularly, inflammation, which can play a role in the pathogenesis of sports anemia. Here, we review recent advances showing the relevance of iron for physical exercise and athletic performance. Low body iron levels can cause anemia and thus limit the delivery of oxygen to exercising muscle, but tissue iron deficiency may also affect performance by, for example, hampering muscle oxidative metabolism. Accordingly, a hemoglobin-independent effect of iron on exercise capacity has been demonstrated in animal models and humans. Here, we review recent advances showing the relevance of iron for physical exercise and athletic performance. PMID:25494391

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

    International Nuclear Information System (INIS)

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

  12. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  13. Safety Evaluation of the EVOLVE Blanket Concept

    International Nuclear Information System (INIS)

    This article summarizes the results of the safety evaluation of the Evaporation of Lithium and Vapor Extraction (EVOLVE) W-alloy first wall (FW) and blanket concept. We have analyzed the EVOLVE design response during a confinement bypass accident. A confinement bypass accident was chosen because, based on previous safety studies, this accident can produce environmental releases by breaching the primary radioactive confinement boundary of EVOLVE, which is the EVOLVE vacuum vessel (VV). As a consequence of a bypass accident, air from a room adjoining the reactor enters the plasma chamber by way of a failed VV port. This air reacts with the high temperature metals inside of the VV to release energy in the case of a lithium spill, or to mobilize radioactive material by oxidation, and then transport this material to the environment by natural convection airflow through the failed VV port. We use the MELCOR code to analyze the response of EVOLVE during this accident. Based on these results, the EVOLVE concept can meet the no-evacuation dose goal set by the DOE Fusion Safety Standard if the EVOLVE confinement building ventilation system is closed within two hours of the onset of this accident

  14. Current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

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

  15. Flow characteristics of the Cascade granular blanket

    International Nuclear Information System (INIS)

    Analysis of a single granule on a rotating cone shows that for the 350 half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-15

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

  18. Structural design and preliminary analysis of liquid lead–lithium blanket for China Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    China Fusion Engineering Test Reactor (CFETR) has been proposed as an option in China to bridge the gaps between ITER and fusion power plant. Since one major goal of CFETR is to demonstrate long pulse or steady-state operation with duty cycle time ≥0.3–0.5, easier maintenance of the in-vessel components is emphasized in the design process. In this contribution, a kind of liquid lead–lithium tritium breeder blanket concept focus on the remote maintenance has been designed for CFETR. To make the pipes and mechanical connections at the rear of the blanket accessible from vacuum vessel, two kinds of guide tubes were adopted to provide passageways for remote handling tools. In order to evaluate the effects of the guide tube installation on the structural performance of the blanket, as a preliminary stage, thermal-hydraulic analysis of first wall was carried out based on the heat load obtained from 3D modeled neutronics calculations. In addition, thermal stress analysis of the first wall under normal condition was performed to evaluate the thermomechanical behavior. The preliminary analysis results validated the performance of current blanket design

  19. Structural design and preliminary analysis of liquid lead–lithium blanket for China Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ni, Muyi; Lian, Chao; Zhang, Shichao; Nie, Baojie; Jiang, Jieqiong, E-mail: jieqiong.jiang@fds.org.cn

    2015-05-15

    China Fusion Engineering Test Reactor (CFETR) has been proposed as an option in China to bridge the gaps between ITER and fusion power plant. Since one major goal of CFETR is to demonstrate long pulse or steady-state operation with duty cycle time ≥0.3–0.5, easier maintenance of the in-vessel components is emphasized in the design process. In this contribution, a kind of liquid lead–lithium tritium breeder blanket concept focus on the remote maintenance has been designed for CFETR. To make the pipes and mechanical connections at the rear of the blanket accessible from vacuum vessel, two kinds of guide tubes were adopted to provide passageways for remote handling tools. In order to evaluate the effects of the guide tube installation on the structural performance of the blanket, as a preliminary stage, thermal-hydraulic analysis of first wall was carried out based on the heat load obtained from 3D modeled neutronics calculations. In addition, thermal stress analysis of the first wall under normal condition was performed to evaluate the thermomechanical behavior. The preliminary analysis results validated the performance of current blanket design.

  20. Advancing Replicable Solutions for High-Performance Homes in the Southeast

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, S. G. [Partnership for Home Innovation, Upper Marlboro, MD (United States). Southface Energy Inst.; Sweet, M. L. [Partnership for Home Innovation, Upper Marlboro, MD (United States). Southface Energy Inst.; Francisco, A. [Partnership for Home Innovation, Upper Marlboro, MD (United States). Southface Energy Inst.

    2016-03-01

    The work presented in this report advances the goals of the U.S. Department of Energy Building America program by improving the energy performance of affordable and market-rate housing. Southface Energy Institute (Southface), part of the U.S. Department of Energy Building America research team Partnership for Home Innovation, worked with owners and builders with various market constraints and ultimate goals for three projects in three climate zones (CZs): Savannah Gardens in Savannah, Georgia (CZ 2); JMC Patrick Square in Clemson, South Carolina (CZ 3); and LaFayette in LaFayette, Georgia (CZ 4). This report documents the design process, computational energy modeling, construction, envelope performance metrics, long-term monitoring results, and successes and failures of the design and execution of these high-performance homes.

  1. Basic Concepts of DEMO and a Design of a Helium Cooled Molten Lithium Blanket

    International Nuclear Information System (INIS)

    Demonstration fusion power plant, DEMO is regarded as the last step before the development of a commercial fusion reactor in Korea National Basic Plan for the Development of Fusion Energy. The DEMO should demonstrate a net electric power generation, a tritium self sufficiency, and the safety aspect of a power plant. With a limited extension of the improved plasma physics and technology from the 2nd phase of the ITER operation (EPP phase), we developed the basic concepts of DEMO and identified the design parameters by considering the dependence of DEMO on performance objectives, design features and physical and technical constraints. Extensive system analyses have been performed to find device variables which optimize figures of merit such as major radius, ignition margin, neutron wall load, etc. The He Cooled Molten Lithium/FS (HCML) blanket is one of options for DEMO blanket and its tritium breeding capability and heat removal capability will be tested in ITER as a test blanket module (TBM). HCML blanket uses He as a coolant and Li as a tritium breeder. From a sensitivity study, 6Li enrichment was optimized in terms of tritium breeding ratio (TBR). An optimum was found for a natural enrichment in DEMO blanket but it was 12 wt% in TBM since the amount of Li is limited in ITER. Two layers of a graphite reflector were inserted as a reflector in the breeder zone to increase the TBR and the shielding performances. The graphite reflector thickness was optimized to maximize TBR without any special neutron multiplier and to minimize the neutron leakage. For TBM, a 3-D Monte Carlo neutronic analysis was performed with the MCCARD code and the total power was founded to be a 0.739 MW at normal heat flux 0.3 MW/m2 from plasma side. From the thermal-hydraulic analysis using CFX-10, the He cooling path was optimized and it was found that the maximum temperature of FW is below 550 oC at structural materials and the coolant velocities are 45 m/sec and 8.2 m/sec at FW and breeding

  2. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    International Nuclear Information System (INIS)

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials (1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF4 or ThF4 or some combination thereof. Future systems could look at using PuF3 or PuF4 as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory. Preliminary design studies

  3. Advancing the use of checklists for evaluating performance in health care.

    Science.gov (United States)

    Rosen, Michael A; Pronovost, Peter J

    2014-07-01

    Patients frequently do not receive recommended therapies because performance expectations are often unclear. Clinical guidelines provide exhaustive details and recommendations, but this information is not formatted in a way that supports decision making or bedside translation of therapies. When performance expectations are unclear, it is difficult for clinicians to assess their own or others' competence. Checklists offer hope because they codify interventions, remove ambiguity, and increase reliability of care processes. Schmutz and colleagues developed a robust methodology to create a checklist for evaluating clinical performance, which is described in this issue of Academic Medicine.In this commentary, the authors offer several points to consider as checklists become more prevalent in medical education and clinical practice. First, culture is a much larger part of the equation than the checklist; understanding what all stakeholders expect to gain will help engage checklist use. Second, the construction, validation, and maintenance of checklist evaluation tools is labor intensive, requiring innovative dissemination approaches to ensure maximum access and use of checklists. Third, integrated systems that evaluate technically specified and adaptive performance are needed because some aspects of clinical performance cannot be captured on a checklist. Fourth, checklists provide an opportunity to evaluate and improve an individual's performance concurrently with the context in which it is delivered. A tighter connection between education and training activities and process improvement strategies will accelerate improvements in safety and quality. Schmutz and colleagues have provided advancements in performance evaluation that will help health care achieve higher-quality and safer care. PMID:24826861

  4. Environmental performance of advanced hybrid energy storage systems for electric vehicle applications

    International Nuclear Information System (INIS)

    Highlights: • The environmental impact of advanced energy storage systems is assessed. • The methodology used is Life Cycle Assessment following the ISO 14040 and 14044. • Twelve impact categories are assessed to avoid burden shifting. • Increasing the efficiency and extending the lifetime benefits the environmental performance. • The results show that there are hot spots where to act and reduce the overall impact. - Abstract: In this paper the environmental performance of an advanced hybrid energy storage system, comprising high power and high energy lithium iron phosphate cells, is compared with a stand alone battery concept composed of lithium manganese oxide cells. The methodology used to analyse the environmental impacts is Life Cycle Assessment (LCA). The manufacturing, use phase and end-of-life of the battery packs are assessed for twelve impact categories. The functional unit is 1 km driven under European average conditions. The present study assesses the environmental performance of the two battery packs for two scenarios: scenario 1 with a vehicle total drive range of 150,000 km and scenario 2 with total driving range of the car of 300,000 km. The results of scenario 1 show that the increased efficiency of the hybrid system reduces, in general, the environmental impact during the use stage, although the manufacturing stage has higher impact than the benchmark. Scenario 2 shows how the extended lifetime of the hybrid system benefits the emissions per km driven

  5. Commercial scale performance predictions for high-temperature electrolysis plants coupled to three advanced reactor types

    International Nuclear Information System (INIS)

    This paper presents results of system analyses that have been developed to assess the hydrogen-production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor - power-cycle combinations: a high-temperature helium-cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to-hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable hydrogen production rates with the high-temperature helium-cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor. (authors)

  6. Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

    International Nuclear Information System (INIS)

    Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

  7. A standard data set for performance analysis of advanced IR image processing techniques

    Science.gov (United States)

    Weiß, A. Robert; Adomeit, Uwe; Chevalier, Philippe; Landeau, Stéphane; Bijl, Piet; Champagnat, Frédéric; Dijk, Judith; Göhler, Benjamin; Landini, Stefano; Reynolds, Joseph P.; Smith, Leslie N.

    2012-06-01

    Modern IR cameras are increasingly equipped with built-in advanced (often non-linear) image and signal processing algorithms (like fusion, super-resolution, dynamic range compression etc.) which can tremendously influence performance characteristics. Traditional approaches to range performance modeling are of limited use for these types of equipment. Several groups have tried to overcome this problem by producing a variety of imagery to assess the impact of advanced signal and image processing. Mostly, this data was taken from classified targets and/ or using classified imager and is thus not suitable for comparison studies between different groups from government, industry and universities. To ameliorate this situation, NATO SET-140 has undertaken a systematic measurement campaign at the DGA technical proving ground in Angers, France, to produce an openly distributable data set suitable for the assessment of fusion, super-resolution, local contrast enhancement, dynamic range compression and image-based NUC algorithm performance. The imagery was recorded for different target / background settings, camera and/or object movements and temperature contrasts. MWIR, LWIR and Dual-band cameras were used for recording and were also thoroughly characterized in the lab. We present a selection of the data set together with examples of their use in the assessment of super-resolution and contrast enhancement algorithms.

  8. Design and Performance Evaluation of Sensors and Actuators for Advanced Optical Systems

    Science.gov (United States)

    Clark, Natalie

    2011-01-01

    Current state-of-the-art commercial sensors and actuators do not meet many of NASA s next generation spacecraft and instrument needs. Nor do they satisfy the DoD needs for satellite missions, especially micro/nano satellite missions. In an effort to develop advanced optical devices and instruments that meet mission requirements, NASA Langley recently completed construction of a new cleanroom housing equipment capable of fabricating high performance active optic and adaptive optic technologies including deformable mirrors, reconfigurable lenses (both refractive and diffractive), spectrometers, spectro-polarimeters, tunable filters and many other active optic devices. In addition to performance, these advanced optic technologies offer advantages in speed, size, weight, power consumption, and radiation tolerance. The active optic devices described in this paper rely on birefringent liquid crystal materials to alter either the phase or the polarization of the incoming light. Design considerations and performance evaluation results for various NASA applications are presented. Applications presented will include large space telescopes, optical communications, spacecraft windows, coronagraphs, and star trackers. Keywords: Photonics, Adaptive Optics, Tunable Filters, MEMs., MOEMs, Coronagraph, Star Tracker

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

    Science.gov (United States)

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

    2016-02-01

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

  10. Remote-handling concept for target/blanket modules in the accelerator production of tritium

    International Nuclear Information System (INIS)

    The accelerator production of tritium (APT) has been proposed as the source of tritium for the United States in the next century. The APT will accelerate protons that will strike replaceable tungsten target modules. The tungsten target modules generate neutrons that pass through blanket modules and other modules where He gas is turned into tritium. The target and blanket modules are predicted to require replacement every 1 to 10 yr, depending on their location. The target modules may weigh as much as 78.8 tonnes (85 t) each. All of the modules will be contained in a target/blanket vessel, which is in a shielded facility. The spent modules will be very radioactive so that remote replacement is required. A proposed concept is to use a remotely operated bridge crane and a remotely operated, bridge-mounted manipulator to perform the entire replacement operation. This will require removing/replacing the vessel lid, installing/removing temporary water cooling, closing/opening valves on manifolds and modules, draining of jumpers, removing/replacing jumpers, removing/replacing shielding keys, and removing/replacing the modules. This application is unique because of the size and weight of the modules, the precision required, the type of connectors required, and the complexity of the entire operation. A three-dimensional simulation of the entire module replacement operation has been developed to help understand, communicate, and refine the concepts

  11. Neutronic and thermal calculation of blanket for high power operating condition of fusion reactor

    International Nuclear Information System (INIS)

    Internal (breeding region) structures of ceramic breeder blanket to accommodate high power operating conditions such as a DEMO reactor have been investigated. The conditions considered here are the maximum neutron wall load of 2.8 MW/m2 at outboard midplane corresponding to a fusion power of 3.0 GW and the coolant temperature of 200 degrees C. Structure of a blanket is based on the layered pebble bed concept, which has been proposed by Japan since the ITER CDA. Lithium oxide with 50% enriched 6Li is used in a shape of small spherical pebbles which are filled in a 316SS can avoid its compatibility issue with Be. Beryllium around the breeder can is filled also in a shape of spherical pebbles which works not only as a neutron multiplier but also as a thermal resistant layer to maintain breeder temperature for effective in-situ tritium recovery. Diameters and packing fractions of both pebbles are ≤ 1 mm and 65%, respectively. A layer of block Be between cooling panels is introduced as a neutron multiplier (not as the thermal resistant layer) to enhance tritium breeding performance. Inlet temperature of water coolant is 200 degrees C to meet the high temperature conditioning requirement to the first wall which is one of walls of the blanket vessel. Neutronics calculations have been carried out by one-dimensional transport code, and thermal calculations have also been carried out by one-dimensional slab code

  12. Development of electrically insulating coatings on vanadium alloys for lithium-cooled blankets

    International Nuclear Information System (INIS)

    The self-cooled lithium blanket concept with a vanadium structure offers a potential for high performance with attractive safety and environmental features. Based on blanket design studies, it became apparent that electrically insulating duct walls would be required to reduce the magnetohydrodynamic (MHD) pressure drop for liquid metal-cooled blankets for high magnetic field fusion devices. As a result, development of insulator coatings was recommended as the most appropriate approach for resolving this issue. Oxides such as CaO, Y2O3, BeO, MgO, MgAl2O4, and Y3Al2O12 and nitrides such as AlN, BN and Si3N2 were initially considered potential candidate coating materials. Based on results of scoping studies, CaO and AlN have been selected as primary candidates for further development. Progress on the development of CaO and AlN coatings, including in-situ formation and electrical properties measurements, are summarized in this paper

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    R Saeedi

    2010-06-01

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

  15. A spallation neutron source based on Pb target surrounded by U blanket

    International Nuclear Information System (INIS)

    A new spallation source, efficient for transmutation experiments, was designed and constructed at the Dubna High Energy Laboratory (LHE). The spallation source has a cylindrical shape Pb target surrounded by natU rods. Experiments with protons of 0.7-2 GeV were performed and neutron spatial distribution on the surface of U blanket was studied. Total neutron fluences and estimates of their energy distributions were determined using solid state nuclear track detectors. Slow and fast neutron components were studied as a function of the proton beam energy. The experimental results were fitted and compared with calculations derived from empirical relations based on physics near high-energy accelerators. The results show that neutron spatial distribution along the U blanket surface (parallel to the beam direction) has the same shape independent of the proton beam energy. The neutron fluence spatial distribution is characterized by an increase at the beginning of the target and after reaching a maximum drops as a function of the target thickness. The maximum is reached at about one mean free path of protons in the Pb target. The total number of neutrons produced, as was measured on U blanket surface, is an increasing function of the proton beam energy

  16. Design requirements for SiC/SiC composites structural material in fusion power reactor blankets

    International Nuclear Information System (INIS)

    This paper recalls the main features of the TAURO blanket, a self-cooled Pb-17Li concept using SiC/SiC composites as structural material, developed for FPR. The objective of this design activity is to compare the characteristics of present-day industrial SiC-SiC composites with those required for a fusion power reactor blanket (FPR) and to evaluate the main needs of further R and D. The performed analyses indicated that the TAURO blanket would need the availability of SiC/SiC composites approximately 10 mm thick with a thermal conductivity through the thickness of approximately 15 Wm-1K-1 at 1000 C and a low electrical conductivity. A preliminary MHD analysis has indicated that the electrical conductivity should not be greater than 500 Ω-1m-1. Irradiation effects should be included in these figures. Under these conditions, the calculated pressure drop due to the high Pb-17Li velocity (approximately 1 m s-1) is much lower then 0.1 MPa. The characteristics and data base of the recently developed 3D-SiC/SiC composite, Cerasep trademark N3-1, are reported and discussed in relation to the identified blanket design requirements. The progress on joining techniques is briefly reported. For the time being, the best results have been obtained using Si-based brazing systems initially developed for SiC ceramics and whose major issue is the higher porosity of the SiC/SiC composites. (orig.)

  17. Seismic isolation of Advanced LIGO: Review of strategy, instrumentation and performance

    Science.gov (United States)

    Matichard, F.; Lantz, B.; Mittleman, R.; Mason, K.; Kissel, J.; Abbott, B.; Biscans, S.; McIver, J.; Abbott, R.; Abbott, S.; Allwine, E.; Barnum, S.; Birch, J.; Celerier, C.; Clark, D.; Coyne, D.; DeBra, D.; DeRosa, R.; Evans, M.; Foley, S.; Fritschel, P.; Giaime, J. A.; Gray, C.; Grabeel, G.; Hanson, J.; Hardham, C.; Hillard, M.; Hua, W.; Kucharczyk, C.; Landry, M.; Le Roux, A.; Lhuillier, V.; Macleod, D.; Macinnis, M.; Mitchell, R.; O'Reilly, B.; Ottaway, D.; Paris, H.; Pele, A.; Puma, M.; Radkins, H.; Ramet, C.; Robinson, M.; Ruet, L.; Sarin, P.; Shoemaker, D.; Stein, A.; Thomas, J.; Vargas, M.; Venkateswara, K.; Warner, J.; Wen, S.

    2015-09-01

    The new generation of gravitational waves detectors require unprecedented levels of isolation from seismic noise. This article reviews the seismic isolation strategy and instrumentation developed for the Advanced LIGO observatories. It summarizes over a decade of research on active inertial isolation and shows the performance recently achieved at the Advanced LIGO observatories. The paper emphasizes the scientific and technical challenges of this endeavor and how they have been addressed. An overview of the isolation strategy is given. It combines multiple layers of passive and active inertial isolation to provide suitable rejection of seismic noise at all frequencies. A detailed presentation of the three active platforms that have been developed is given. They are the hydraulic pre-isolator, the single-stage internal isolator and the two-stage internal isolator. The architecture, instrumentation, control scheme and isolation results are presented for each of the three systems. Results show that the seismic isolation sub-system meets Advanced LIGO’s stringent requirements and robustly supports the operation of the two detectors.

  18. Effect of advanced injection timing on the performence of natural gas in diesel engines

    Indian Academy of Sciences (India)

    O M I Nwafor

    2000-02-01

    Concern over the environment and/or the increasing demand for conventional fossil fuel has promoted interest in the development of alternative sources of fuel energy for internal combustion (IC) engines. The effect ofadvanced injection timing on the performance of natural gas used as primary fuel in dual-fuel combustion has been examined. Satisfactory diesel engine combustion demands self-ignition of the fuel as it is injected near the top dead centre (TDC) into the hot swirling compressed cylinder gas. Longer delays between injection and ignition lead to unacceptable rates of pressure rise (diesel knock) because too much fuel is ready to burn when combustion eventually occurs. Natural gas has been noted to exhibit longer ignition delays and slower burning rates especially at low load levels hence resulting in late combustion in the expansion stroke. Advanced injection timing is expected to compensate for these effects. The engine has standard injection timing of 30° before TDC (BTDC). The injection was first advanced by 5.5° given injection timing of 35.5° BTDC. The engine ran for about 5 minutes at this timing and stopped. The engine failed to start upon subsequent attempts. The injection was then advanced by 3.5° (i.e. 33.5° BTDC). The engine ran smoothly on this timing butseemed to incur penalty on fuel consumption especially at high load levels.

  19. Modified-open fuel cycle performance with breed-and-burn advanced reactor concepts

    International Nuclear Information System (INIS)

    Recent advances in fast reactor designs enable significant increase in the uranium utilization in an advanced fuel cycle. The category of fast reactors, collectively termed breed-and-burn reactor concepts, can use a large amount of depleted uranium as fuel without requiring enrichment with the exception of the initial core critical loading. Among those advanced concepts, some are foreseen to operate within a once-through fuel cycle such as the Traveling Wave Reactor, CANDLE reactor or Ultra-Long Life Fast Reactor, while others are intended to operate within a modified-open fuel cycle, such as the Breed-and-Burn reactor and the Energy Multiplier Module. This study assesses and compares the performance of the latter category of breed-and-burn reactors at equilibrium state. It is found that the two reactor concepts operating within a modified-open fuel cycle can significantly improve the sustainability and security of the nuclear fuel cycle by decreasing the uranium resources and enrichment requirements even further than the breed-and-burn core concepts operating within the once-through fuel cycle. Their waste characteristics per unit of energy are also found to be favorable, compared to that of currently operating PWRs. However, a number of feasibility issues need to be addressed in order to enable deployment of these breed-and-burn reactor concepts. (author)

  20. Fabrication of the full scale separable first wall of ITER shielding blanket

    International Nuclear Information System (INIS)

    Shielding blanket for ITER-FEAT applies the unique first wall structure which is separable from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. Such unique features of blanket structure required technological clarification from the technical base of the previous achievement of the blanket module fabrication development. Previously, within the EDA Task T216+, a prototype for the no.4 Primary Wall Module of the ITER Shield Blanket with integrated first wall has been manufactured by forging and drilling and the first wall has been manufactured and joined to the shield block by Hot Isostatic Pressing (HIP) in one step process. This work has been performed to clarify the remaining R and D issues which have not been covered in the previous R and D. This report summarizes the demonstrative fabrication of the real scale separable first wall for ITER shielding blanket designed for ITER-FEAT, together with the essential technology developments such as, the slit grooving of the first wall with beryllium armor and SS shield block and fabrication of a partial mockup of beryllium armored first wall panel with built-in cooling channels. This work has been performed under the task agreement of G 16 TT 95 FJ (T420-1) in ITER Engineering Design Activity Extension Period. By the demonstration of the Be armor joining to the first wall panel, the joining technique of Be and DSCu developed previously, was shown to be applicable to the realistic structure of first wall panel. Also, the slit grooving by an end-mill method and an electron discharge machining method have been applied to the first wall mockup with Be armor tiles and demonstrated the applicability within the design tolerance. As the slit grooving technique