WorldWideScience

Sample records for protective thermal blankets

  1. Synthesis and Characterization of Fibre Reinforced Silica Aerogel Blankets for Thermal Protection

    Directory of Open Access Journals (Sweden)

    S. Chakraborty

    2016-01-01

    Full Text Available Using tetraethoxysilane (TEOS as the source of silica, fibre reinforced silica aerogels were synthesized via fast ambient pressure drying using methanol (MeOH, trimethylchlorosilane (TMCS, ammonium fluoride (NH4F, and hexane. The molar ratio of TEOS/MeOH/(COOH2/NH4F was kept constant at 1 : 38 : 3.73 × 10−5 : 0.023 and the gel was allowed to form inside the highly porous meta-aramid fibrous batting. The wet gel surface was chemically modified (silylation process using various concentrations of TMCS in hexane in the range of 1 to 20% by volume. The fibre reinforced silica aerogel blanket was obtained subsequently through atmospheric pressure drying. The aerogel blanket samples were characterized by density, thermal conductivity, hydrophobicity (contact angle, and Scanning Electron Microscopy. The radiant heat resistance of the aerogel blankets was examined and compared with nonaerogel blankets. It has been observed that, compared to the ordinary nonaerogel blankets, the aerogel blankets showed a 58% increase in the estimated burn injury time and thus ensure a much better protection from heat and fire hazards. The effect of varying the concentration of TMCS on the estimated protection time has been examined. The improved thermal stability and the superior thermal insulation of the flexible aerogel blankets lead to applications being used for occupations that involve exposure to hazards of thermal radiation.

  2. Thermal Protection with 5% Dextrose Solution Blanket During Radiofrequency Ablation

    International Nuclear Information System (INIS)

    Chen, Enn Alexandria; Neeman, Ziv; Lee, Fred T.; Kam, Anthony; Wood, Brad

    2006-01-01

    A serious complication for any thermal radiofrequency ablation is thermal injury to adjacent structures, particularly the bowel, which can result in additional major surgery or death. Several methods using air, gas, fluid, or thermometry to protect adjacent structures from thermal injury have been reported. In the cases presented in this report, 5% dextrose water (D5W) was instilled to prevent injury to the bowel and diaphragm during radiofrequency ablation. Creating an Insulating envelope or moving organs with D5W might reduce risk for complications such as bowel perforation

  3. Protection against cold in prehospital care-thermal insulation properties of blankets and rescue bags in different wind conditions.

    Science.gov (United States)

    Henriksson, Otto; Lundgren, J Peter; Kuklane, Kalev; Holmér, Ingvar; Bjornstig, Ulf

    2009-01-01

    In a cold, wet, or windy environment, cold exposure can be considerable for an injured or ill person. The subsequent autonomous stress response initially will increase circulatory and respiratory demands, and as body core temperature declines, the patient's condition might deteriorate. Therefore, the application of adequate insulation to reduce cold exposure and prevent body core cooling is an important part of prehospital primary care, but recommendations for what should be used in the field mostly depend on tradition and experience, not on scientific evidence. The objective of this study was to evaluate the thermal insulation properties in different wind conditions of 12 different blankets and rescue bags commonly used by prehospital rescue and ambulance services. The thermal manikin and the selected insulation ensembles were setup inside a climatic chamber in accordance to the modified European Standard for assessing requirements of sleeping bags. Fans were adjusted to provide low (value, Itr (m2 C/Wclo; where C = degrees Celcius, and W = watts), was calculated from ambient air temperature (C), manikin surface temperature (C), and heat flux (W/m2). In the low wind condition, thermal insulation of the evaluated ensembles correlated to thickness of the ensembles, ranging from 2.0 to 6.0 clo (1 clo = 0.155 m2 C/W), except for the reflective metallic foil blankets that had higher values than expected. In moderate and high wind conditions, thermal insulation was best preserved for ensembles that were windproof and resistant to the compressive effect of the wind, with insulation reductions down to about 60-80% of the original insulation capacity, whereas wind permeable and/or lighter materials were reduced down to about 30-50% of original insulation capacity. The evaluated insulation ensembles might all be used for prehospital protection against cold, either as single blankets or in multiple layer combinations, depending on ambient temperatures. However, with extended

  4. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1978-01-01

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

  5. Epoxy blanket protects milled part during explosive forming

    Science.gov (United States)

    1966-01-01

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

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

    Science.gov (United States)

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

    1990-01-01

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

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

    Science.gov (United States)

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

    1993-01-01

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

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

    Science.gov (United States)

    Pippin, Gary; Stuckey, Wayne; Hemminger, Carol

    1993-01-01

    Silverized Teflon (Ag/FEP) is a widely used passive thermal control material for space applications. The material has a very low alpha/e ratio (less than 0.1) for low operating temperatures and is fabricated with various FEP thicknesses (as the Teflon thickness increases, the emittance increases). It is low outgassing and, because of its flexibility, can be applied around complex, curved shapes. Ag/FEP has achieved multiyear lifetimes under a variety of exposure conditions. This has been demonstrated by the Long Duration Exposure Facility (LDEF), Solar Max, Spacecraft Charging at High Altitudes (SCATHA), and other flight experiments. Ag/FEP material has been held in place on spacecraft by a variety of methods: mechanical clamping, direct adhesive bonding of tapes and sheets, and by Velcro(TM) tape adhesively bonded to back surfaces. On LDEF, for example, 5-mil blankets held by Velcro(TM) and clamping were used for thermal control over 3- by 4-ft areas on each of 17 trays. Adhesively bonded 2- and 5-mil sheets were used on other LDEF experiments, both for thermal control and as tape to hold other thermal control blankets in place. Performance data over extended time periods are available from a number of flights. The observed effects on optical properties, mechanical properties, and surface chemistry will be summarized in this paper. This leads to a discussion of performance life estimates and other design lessons for Ag/FEP thermal control material.

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

    International Nuclear Information System (INIS)

    Daly, E.; Pletzer, R.

    1992-04-01

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

  10. Vibration damage testing of thermal barrier fibrous blanket insulation

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.

    1984-01-01

    GA Technologies is engaged in a long-term, multiphase program to determine the vibration characteristics of thermal barrier components leading to qualification of assemblies for High Temperature Gas-Cooled Reactor (HTGR) service. The phase of primary emphasis described herein is the third in a series of acoustic tests and uses as background the more elemental tests preceding it. Two sizes of thermal barrier coverplates with one fibrous blanket insulation type were tested in an acoustic environment at sound pressure levels up to 160 dB. Three tests were conducted using sinusoidal and random noise for up to 200 h duration at room temperature. Frequent inspections were made to determine the progression of degradation using definition of stages from a prior test program. Initially the insulation surface adjacent to the metallic seal sheets (noise side) assumed a chafed or polished appearance. This was followed by flattening of the as-received pillowed surface. This stage was followed by a depression being formed in the vicinity of the free edge of the coverplate. Next, loose powder from within the blanket and from fiber erosion accumulated in the depression. Prior experience showed that the next stage of deterioration exhibited a consolidation of the powder to form a local crust. In this test series, this last stage generally failed to materialize. Instead, surface holes generated by solid ceramic particulates (commonly referred to as 'shot') constituted the stage following powdering. With the exception of some manufacturing-induced anomalies, the final stage, namely, gross fiber breakup, did not occur. It is this last stage that must be prevented for the thermal barrier to maintain its integrity. (orig./GL)

  11. Thermal mechanical analysis of a solid breeding blanket

    International Nuclear Information System (INIS)

    Aquaro, Donato

    2003-01-01

    This paper deals with a theoretical model of thermal mechanical behaviour of pebble beds, used as neutron multiplier or tritium breeder in the breeding blanket of a fusion nuclear reactor. The model tries to sum up the advantages of the two approaches ('discrete' method and macroscopic method), presently used for analysing the pebble bed behaviour, without their intrinsic disadvantages. The developed method has the capability to describe the microscopic behaviour of the single sphere (as the discrete approach does), and the capability to model complex structures under variable loads, typical of the macroscopic approach, without doing the unrealistic assumption of continuum homogeneous and isotropic material. The model describes the thermal mechanical behaviour of a single sphere compressed in elastic plastic conditions. The obtained relations have been extrapolated to regular lattices of spheres and subsequently to pebble beds (characterised by a macroscopic parameter called 'packing factor') of simple geometric shapes using statistical considerations. The results of the model have been assessed by comparison with results obtained by means of numerical simulations and experimental tests. The ongoing activity is the implementation in a FEM code of a new finite element, which represents one or several regular lattices of spheres, the non linear stiffness of which is obtained from the mono dimensional compression model of one sphere. The results of the numerical simulation permits to construct and display the strain and stress distribution of the single spheres by means of an implemented graphical interface

  12. Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie

    1998-01-01

    Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  14. Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

    1980-09-01

    The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design

  15. Thermal stresses and cyclic creep-fatigue in fusion reactor blanket

    International Nuclear Information System (INIS)

    Liu, K.C.

    1977-01-01

    Thermal stresses in the first walls of fusion reactor blankets were studied in detail. ORNL multibucket modules are emphasized. Practicality of using the bucket module rather than other blanket designs is examined. The analysis shows that applying intelligent engineering judgment in design can reduce the thermal stresses significantly. Arrangement of coolant flow and distribution of temperature are reviewed. Creep-fatigue property requirements for a first wall are discussed on the basis of existing design rules and criteria. Some major questions are pointed out and experiments needed to resolve basic uncertainties relative to key design decisions are discussed

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

    International Nuclear Information System (INIS)

    Aquaro, D.; Morellini, D.; Cerullo, N.

    2006-01-01

    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 (Li 4 SiO 4 and Li 2 TiO 3 ) with different enrichment in 6 Li 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 o C for the steel, 950 o C for the breeder and 650 o C 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 Li 4 SiO 4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give

  17. Thermal and mechanical design of WITAMIR-I blanket

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.N.

    1980-10-01

    The design philosophy of WITAMIR-I, a Wisconsin Tandem Mirror Reactor design study, uses the experience obtained from our previous tokamak studies and combines it with the unique features of the tandem mirror to obtain an attractive design of a TM power reactor. It is aimed at maximizing the strengths of the tandem mirror while mitigating its weaknesses. The end product should be a safe, reliable, maintainable and a relatively economic power reactor. The general description of the reactor, the plasma calculations, the magnet design, the neutronic calculations and the maintenance considerations are presented elsewhere. This paper presents the blanket design of this reactor study

  18. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    International Nuclear Information System (INIS)

    Li, Jia; Jiang, Kecheng; Zhang, Xiaokang; Nie, Xingchen; Zhu, Qinjun; Liu, Songlin

    2016-01-01

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  19. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jia, E-mail: lijia@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China); Nie, Xingchen [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Zhu, Qinjun; Liu, Songlin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2016-12-15

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  20. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1978-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  1. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1977-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  2. Evaluations of Silica Aerogel-Based Flexible Blanket as Passive Thermal Control Element for Spacecraft Applications

    Science.gov (United States)

    Hasan, Mohammed Adnan; Rashmi, S.; Esther, A. Carmel Mary; Bhavanisankar, Prudhivi Yashwantkumar; Sherikar, Baburao N.; Sridhara, N.; Dey, Arjun

    2018-03-01

    The feasibility of utilizing commercially available silica aerogel-based flexible composite blankets as passive thermal control element in applications such as extraterrestrial environments is investigated. Differential scanning calorimetry showed that aerogel blanket was thermally stable over - 150 to 126 °C. The outgassing behavior, e.g., total mass loss, collected volatile condensable materials, water vapor regained and recovered mass loss, was within acceptable range recommended for the space applications. ASTM tension and tear tests confirmed the material's mechanical integrity. The thermo-optical properties remained nearly unaltered in simulated space environmental tests such as relative humidity, thermal cycling and thermo-vacuum tests and confirmed the space worthiness of the aerogel. Aluminized Kapton stitched or anchored to the blanket could be used to control the optical transparency of the aerogel. These outcomes highlight the potential of commercial aerogel composite blankets as passive thermal control element in spacecraft. Structural and chemical characterization of the material was also done using scanning electron microscopy, Fourier transform infrared spectroscopy and x-ray photoelectron spectroscopy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-15

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

  4. Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

    International Nuclear Information System (INIS)

    Grotz, S.; Ghoniem, N.M.

    1986-02-01

    The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)

  5. Thermal comfort and safety of cotton blankets warmed at 130°F and 200°F.

    Science.gov (United States)

    Kelly, Patricia A; Cooper, Susan K; Krogh, Mary L; Morse, Elizabeth C; Crandall, Craig G; Winslow, Elizabeth H; Balluck, Julie P

    2013-12-01

    In 2009, the ECRI Institute recommended warming cotton blankets in cabinets set at 130°F or less. However, there is limited research to support the use of this cabinet temperature. To measure skin temperatures and thermal comfort in healthy volunteers before and after application of blankets warmed in cabinets set at 130 and 200°F, respectively, and to determine the time-dependent cooling of cotton blankets after removal from warming cabinets set at the two temperatures. Prospective, comparative, descriptive. Participants (n = 20) received one or two blankets warmed in 130 or 200°F cabinets. First, skin temperatures were measured, and thermal comfort reports were obtained at fixed timed intervals. Second, blanket temperatures (n = 10) were measured at fixed intervals after removal from the cabinets. No skin temperatures approached levels reported in the literature that cause epidermal damage. Thermal comfort reports supported using blankets from the 200°F cabinet, and blankets lost heat quickly over time. We recommend warming cotton blankets in cabinets set at 200°F or less to improve thermal comfort without compromising patient safety. Copyright © 2013 American Society of PeriAnesthesia Nurses. Published by Elsevier Inc. All rights reserved.

  6. First wall thermal hydraulic models for fusion blankets

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1980-01-01

    Subject to normal and off-normal reactor conditions, thermal hydraulic models of first walls, e.g., a thermal mass barrier, a tubular shield, and a radiating liner are reviewed. Under normal operation the plasma behaves as expected in a predicted way for transient and steady-state conditions. The most severe thermal loading on the first wall occurs when the plasma becomes unstable and dumps its energy on the wall in a very short period of time (milliseconds). Depending on the plasma dump time and area over which the energy is deposited may result in melting of the first wall surface, and if the temperature is high enough, vaporization

  7. Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

  8. Effective thermal conductivity of glass-fiber board and blanket standard reference materials

    International Nuclear Information System (INIS)

    Smith, D.R.; Hust, J.G.

    1983-01-01

    This chapter reports on measurements of effective thermal conductivity performed on a series of specimens of glass-fiber board and glass-fiber blanket. Explains that measurements of thermal conductivity were conducted as a function of temperature from 85 to 360 K, of temperature difference with T=10 to 100 K, of bulk density from 11 to 148 kg/m 3 and for nitrogen, argon, and helium inter-fiber fill gases at pressures from atmospheric to high vacuum. Analyzes and compares results with values from the published literature and National Bureau of Standards (NBS) certification data for similar material. Gives polynomial expressions for the functional relation between conductivity, temperature, and density for board and for blanket

  9. Ablative thermal protection systems

    International Nuclear Information System (INIS)

    Vaniman, J.; Fisher, R.; Wojciechowski, C.; Dean, W.

    1983-01-01

    The procedures used to establish the TPS (thermal protection system) design of the SRB (solid rocket booster) element of the Space Shuttle vehicle are discussed. A final evaluation of the adequacy of this design will be made from data obtained from the first five Shuttle flights. Temperature sensors installed at selected locations on the SRB structure covered by the TPS give information as a function of time throughout the flight. Anomalies are to be investigated and computer design thermal models adjusted if required. In addition, the actual TPS ablator material loss is to be measured after each flight and compared with analytically determined losses. The analytical methods of predicting ablator performance are surveyed. 5 references

  10. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    International Nuclear Information System (INIS)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available

  11. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  12. Comprehensive structural analysis of the HCPB demo blanket under thermal, mechanical, electromagnetic and radiation induced loads

    International Nuclear Information System (INIS)

    Boccaccini, L.V.; Norajitra, P.; Ruatto, P.; Scaffidi-Argentina, F.

    1998-01-01

    For the helium-cooled pebble bed (HCPB) blanket, which is one of the two reference concepts studied within the European Demo Development Program, a comprehensive finite element (FEM) structural analysis has been performed. The analysis refers to the steady-state operating conditions of an outboard blanket segment. On the basis of a three-dimensional model of radial-toroidal sections of the segment box, thermal stresses caused by the temperature gradients in the blanket structure have been calculated. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions as well as an accidental over-pressurization of the blanket box have been accounted for. The stresses caused by a central plasma major disruption from an initial current of 20 MA to zero in 20 ms have been also taken into account. Radiation-induced dimensional changes of breeder and multiplier material caused by both helium production and neutron damage, have also been evaluated and discussed. All the above loads have been combined as input for a FEM stress analysis and the resulting stress distribution has been evaluated according to the American Society of Mechanical Engineers (ASME) norms. (orig.)

  13. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

  14. Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yanagi, Yoshihiko; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto; Kuroda, Toshimasa; Kosaku, Yasuo; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2001-11-01

    Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li{sub 2}O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo-critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m{sup 2}, {sup 6}Li enrichment of 30% and coolant temperature at inlet of breeder region of 380degC. In the case of the higher coolant temperature 430degC of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%. (author)

  15. Ceramic fiber blanket wrap for fire protection of cable trays and conduits

    International Nuclear Information System (INIS)

    Chaille, C.E.; Reiman, R.J.

    1980-01-01

    In some areas of nuclear power plants, cables of redundant electrical systems, which are necessary for the safe shutdown of the reactor, are in close proximity. If a fire should occur in one of these areas, both electrical systems could be destroyed before the fire is extinguished and control of the reactor may be lost. A ceramic fiber blanket was evaluated as a fire protective wrap around cable trays and conduits. 2 refs

  16. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem mirror fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  17. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem-mirror fusion reactor

    International Nuclear Information System (INIS)

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

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  18. Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng; Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Lin, Shuang [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Huang, Kai [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2016-11-15

    Highlights: • The 3D thermal hydraulic analysis on the whole module of WCCB is performed by CFD method. • Temperature field and mass flow distribution have been obtained. • The design of WCCB is reasonable from the perspective of thermal-hydraulics. • The scheme for further optimization has been proposed. - Abstract: The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). The thermal-hydraulic analysis is essential because the blanket should remove the high heat flux from the plasma and the volumetric heat generated by neutrons. In this paper, the detailed three dimensional (3D) thermal hydraulic analysis on the whole module of WCCB blanket has been performed by Computational Fluid Dynamics (CFD) method, which is capable of solving conjugate heat transfer between solid structure and fluid. The main results, including temperature field, distribution of mass flow rate and coolant pressure drop, have been calculated simultaneously. These provides beneficial guidance data for the further structural optimization and for the design arrangement of primary and secondary circuit. Under the total heat source of 1.23 MW, the coolant mass flow rate of 5.457 kg/s is required to make coolant water corresponding to the Pressurized Water Reactor (PWR) condition (15.5 MPa, 285 °C–325 °C), generating the total coolant pressure drop (△P) of 0.467 MPa. The results show that the present structural design can make all the materials effectively cooled to the allowable temperature range, except for a few small modifications on the both sides of FW. The main components, including the first wall (FW), cooling plates (CPs), side wall (SWs)&stiffening plates (SPs) and the manifold(1–4), dominate 4.7%/41.7%/13%/40.6% of the total pressure drop, respectively. Additionally, the mass flow rate of each channel has been obtained, showing the peak relative deviation of 3.4% and 2% from the average for the paratactic

  19. Thermal hydraulic and power cycle analysis of liquid lithium blanket designs

    International Nuclear Information System (INIS)

    Misra, B.; Stevens, H.C.; Maroni, V.A.

    1977-01-01

    Thermal hydraulic and power cycle analyses were performed for the first-wall and blanket systems of tokamak-type fusion reactors under a typical set of design and operating conditions. The analytical results for lithium-cooled blanket cells show that with stainless steel as construction material and with no divertor present, the maximum allowable neutron wall loading is approximately 2 MW/m 2 and is limited by thermal stress criteria. With vanadium alloy as construction material and no divertor present, the maximum allowable neutron wall loading is approximately 8 MW/m 2 and is limited by an interplay of constraints imposed on the maximum allowable structural temperature and the minimum allowable coolant inlet temperature. With a divertor these wall loadings can be increased by from 40 to 90 percent. The cost of the vanadium system is found to be competitive with the stainless steel system because of the higher allowable structural temperatures and concomitant higher thermal efficiencies afforded by the vanadium alloys

  20. Thermal-hydraulic and neutronic considerations for designing a lithium-cooled tokamak blanket

    International Nuclear Information System (INIS)

    Chao, J.; Mikic, B.; Todreas, N.

    1978-12-01

    A methodology for the design of lithium cooled blankets is developed. The thermal-hydraulics, neutronics and interactions between them are extensively investigated. In thermal hydraulics, two models illustrate the methodology used to obtain the acceptable ranges for a set of design parameters. The methodology can be used to identify the limiting constraints for a particular design. A complete neutronic scheme is set up for the calculations of the volumetric heating rate as a function of the distance from the first wall, the breeding ratio as a function of the amount of structural material in the blanket, and the radiation damage in terms of atom displacements and gas production rate. Different values of the volume percent of Type-316 stainless steel are assigned in four breeding zones to represent a nonuniformly distributed structural material which satisfies various thermal-hydraulic requirements. The role that the radiation damage plays in the overall design methodology is described. The product of the first wall lifetime and neutron loading is limited by the radiation damage which degrades the mechanical properties of the material

  1. Thermal-hydraulic investigations on the CEA-ENEA DEMO relevant helium cooled poloidal blanket

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Polazzi, G.; Vallette, F.; Proust, E.; Eid, M.

    1994-01-01

    The CEA-ENEA design of an Helium Cooled Solid Breeder Blanket (HCSBB) for the DEMO reactor, with a breeder in tube (BIT) poloidal arrangement, is based on the use of lithium ceramic pellets, the ENEA γ-LiAlO 2 or the CEA Li 2 ZrO 3 . Due to the geometry of the DEMO reactor plasma chamber, these breeder bundles are adapted to the Vacuum Vessel with a strong poloidal curvature. This curvature influences the thermal-hydraulic behaviour of the coolant flowing inside the bundle. The paper presents the CEA-ENEA first results of the experimental and theoretical programme, aiming at optimizing the breeder module thermal hydraulic design. (author) 6 refs.; 7 figs.; 1 tab

  2. Neutronic and thermal estimation of blanket in-pile mockup with Li2TiO3 pebbles

    International Nuclear Information System (INIS)

    Nagao, Y.; Nakamichi, M.; Tsuchiya, K.; Kawamura, H.

    2001-01-01

    To evaluate exactly temperature distribution in large volume of tritium breeding materials during the blanket in-pile tests with the JMTR, neutronic and thermal calculations were conducted by using Monte Carlo code 'MCNP' with nuclear cross section library of 'FSXLIBJ3R2' and the transient and steady-state distribution code 'TRUMP'. From the results of preliminary estimation of temperature distribution in the blanket in-pile mockup, the calculated values were 24-28% higher than the measured values. One of the reasons is due to overestimation of calculated thermal neutron flux

  3. 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)

  4. Thermal-hydraulic analysis of water cooled breeding blanket of K-DEMO using MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-Hun; Park, Il Woong; Kim, Geon-Woo; Park, Goon-Cherl [Seoul National University, Seoul (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Highlights: • The thermal design of breeding blanket for the K-DEMO is evaluated using MARS-KS. • To confirm the prediction capability of MARS, the results were compared with the CFD. • The results of MARS-KS calculation and CFD prediction are in good agreement. • A transient simulation was carried out so as to show the applicability of MARS-KS. • A methodology to simulate the entire blanket system is proposed. - Abstract: The thermal design of a breeding blanket for the Korean Fusion DEMOnstration reactor (K-DEMO) is evaluated using the Multidimensional Analysis of Reactor Safety (MARS-KS) code in this study. The MARS-KS code has advantages in simulating transient two-phase flow over computational fluid dynamics (CFD) codes. In order to confirm the prediction capability of the code for the present blanket system, the calculation results were compared with the CFD prediction. The results of MARS-KS calculation and CFD prediction are in good agreement. Afterwards, a transient simulation for a conceptual problem was carried out so as to show the applicability of MARS-KS for a transient or accident condition. Finally, a methodology to simulate the multiple blanket modules is proposed.

  5. Microbiological sampling of spacecraft cabling, antennas, solar panels and thermal blankets

    Science.gov (United States)

    Koukol, R. C.

    1973-01-01

    Sampling procedures and techniques described resulted from various flight project microbiological monitoring programs of unmanned planetary spacecraft. Concurrent with development of these procedures, compatibility evaluations were effected with the cognizant spacecraft subsystem engineers to assure that degradation factors would not be induced during the monitoring program. Of significance were those areas of the spacecraft configuration for which special handling precautions and/or nonstandard sample gathering techniques were evolved. These spacecraft component areas were: cabling, high gain antenna, solar panels, and thermal blankets. The compilation of these techniques provides a historical reference for both the qualification and quantification of sampling parameters as applied to the Mariner Spacecraft of the late 1960's and early 1970's.

  6. High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Ioki, K.; Tivey, R.; Krassovski, D.; Kubik, D.

    1998-01-01

    Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest heat load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal-hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application of results from on-going high heat flux R and D and a brief description of future R and D effort to address remaining issues are also included. (orig.)

  7. Thermal Management and Thermal Protection Systems

    Science.gov (United States)

    Hasnain, Aqib

    2016-01-01

    's rays directly impinging on the system. Heating rate of the lamps were calculated by knowing fraction of emitted energy in a wavelength interval and the filament temperature. This version of the model can be used to predict performance of the system under vacuum with extreme cold or hot conditions. Initial testing of the PTMS showed promise, and the thermal math model predicts even better performance in thermal vacuum testing. ii) Thermal Protection Systems (TPS) are required for vehicles which enter earth's atmosphere to protect from aerodynamic heating caused by the friction between the vehicle and atmospheric gases. Orion's heat shield design has two aspects which needed to be analyzed thermally: i) a small excess of adhesive used to bond the outer AVCOAT layer to the inner composite structure tends to seep from under the AVCOAT and form a small bead in between two bricks of AVCOAT, ii) a silicone rubber with different thermophysical properties than AVCOAT fills the gap between two bricks of AVCOAT. I created a thermal model using TD to determine temperature differences that are caused by these two features. To prevent false results, all TD models must be verified against something known. In this case, the TD model was correlated to CHAR, an ablation modelling software used to analyze TPS. Analyzing a node far from the concerning features, we saw that the TD model data match CHAR data, verifying the TD model. Next, the temperature of the silicone rubber as well as the bead of adhesive were analyzed to determine if they exceeded allowable temperatures. It was determined that these two features do not have a significant effect on the max temperature of the heat shield. This model can be modified to check temperatures at various locations of the heat shield where the composite thickness varies.

  8. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-01

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  9. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-15

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  10. 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.

  11. Multiple Module Simulation of Water Cooled Breeding Blankets in K-DEMO Using Thermal-Hydraulic Analysis Code MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.

  12. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  13. Improved thermal/MHD design of self-cooled blankets for high-power-density fusion reactors

    International Nuclear Information System (INIS)

    Sedehi, S.; Lund, K.O.

    1986-01-01

    In this work, an improved self-cooled blanket design is conceived that seeks to minimize the induced current and pressure loss, while maintaining effective cooling and power output. Standard solutions for fully developed MHD flows in rectangular ducts are utilized to describe the magnetic pressure drop in rectangular ducts in terms of the duct aspects ratio. A newly available analytical result for developing and fully developed temperatures is utilized in determining the maximum wall temperature and outlet temperature. Based on results from rectangular ducts, improved annular-type duct designs are proposed and evaluated. The results from the rectangular duct analysis indicate reduced pressure drop and increased thermal performance for large aspect ratio (ratio of duct width in the toroidal B-field direction to width normal to B-field). An infinite aspect ratio occurs for the annular duct design and it is shown that this configuration has superior characteristics as a self-cooled blanket design concept

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  15. 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)

  16. Mirror reactor blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  17. Thermal convection in a toroidal duct of a liquid metal blanket. Part II. Effect of axial mean flow

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xuan, E-mail: xuanz@umich.edu; Zikanov, Oleg

    2017-03-15

    Highlights: • 2D convection flow develops with internal heating and strong axial magnetic field. • The flow is strongly modified by the buoyancy force associated with growing T{sub m}. • Thermal convection is suppressed at high Gr. • High temperature difference between top and bottom walls is expected at high Gr. - Abstract: The work continues the exploration of the effect of thermal convection on flows in toroidal ducts of a liquid metal blanket. This time we consider the effect of the mean flow along the duct and of the associated heat transfer diverting the heat deposited by captured neutrons. Numerical simulations are conducted for a model system with two-dimensional (streamwise-uniform) fully developed flow, purely toroidal magnetic field, and perfectly electrically and thermally insulating walls. Realistically high Grashof (up to 10{sup 11}) and Reynolds (up to 10{sup 6}) numbers are used. It is found that the flow develops thermal convection in the transverse plane at moderate Grashof numbers. At large Grashof numbers, the flow is dominated by the top-bottom asymmetry of the streamwise velocity and stable stratification of temperature, which are caused by the buoyancy force due to the mean temperature growing along the duct. This leads to suppression of thermal convection, weak mixing, and substantial gradients of wall temperature. Further analysis based on more realistic models is suggested.

  18. Transmutation and activation of stainless steel 316 SS in a thermal fusion reactor blanket

    International Nuclear Information System (INIS)

    Gruber, J.; Schneider, J.

    1977-10-01

    Using the program MATEXP (matrix exponential method) the influence of neutron flux is calculated for stainless steel 3s16 SS which is used as a structural material in a fusion reactor blanket (CTRD-I). The transmutations, activations and γ-dose rates are determined for an operation time of 20 years. Investigating the decay behaviour after operation time, we found that the long term activity and dose rate was mainly influenced by five nuclides: Fe55, Ni63, Ni59, Co60 and Nb94. (orig.) [de

  19. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  20. Proposal for the award of a blanket order contract for the supply of microprocessor-based protection and control devices for the CERN HV distribution network

    CERN Document Server

    2004-01-01

    This document concerns the award of a blanket contract for the supply of microprocessor-based protection and control devices for the CERN HV distribution network. The Finance Committee is invited to agree to the negotiation of a blanket order contract with SCHNEIDER ELECTRIC (PT), the lowest technically acceptable bidder after realignment, for the supply of microprocessor-based protection and control devices for the CERN HV distribution network for a total amount of 1 900 000 euros (2 924 128 Swiss francs), subject to revision for inflation after 1 January 2007. The rate of exchange used is that stipulated in the tender

  1. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    International Nuclear Information System (INIS)

    HEARD, F.J.

    1999-01-01

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels

  2. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  3. Limitations on blanket performance

    International Nuclear Information System (INIS)

    Malang, S.

    1999-01-01

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

  4. System code assessment with thermal-hydraulic experiment to develop helium cooled breeding blanket for nuclear fusion reactor

    International Nuclear Information System (INIS)

    Yum, S. B.; Park, I. W.; Park, G. C.; Lee, D. W.

    2012-01-01

    By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) Test Blanket Module (TBM) for testing in the International Thermonuclear Experimental Reactor (ITER). A performance analysis for the thermal-hydraulics and a safety analysis for an accident caused by a loss of coolant for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs Multicomponent Mixture Analysis), which was developed by the Gas Cooled Reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM First Wall (FW) mock-up made from the same material as tho KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 11, 19, and 29 bar, and under various ranges of flow rate from 0.63 to 2.44kg/min with a constant wall temperature condition. In the present study, a thermal-hydraulic test was performed with the newly constructed helium supplying system, In which the design pressure and temperature were 9 MPa and 500 .deg. C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 8 MPa pressure, 0.2 kg/s flow rate, which are almost same conditions of the KO TBM FW. One-side of the mock-up was heated with a constant heat flux of 0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea Heat Load Test Facility-2). The wall temperatures were measured using installed thermocouples, and they show a strong parity with the code results simulated under the same test conditions

  5. Large thermal protection system panel

    Science.gov (United States)

    Weinberg, David J. (Inventor); Myers, Franklin K. (Inventor); Tran, Tu T. (Inventor)

    2003-01-01

    A protective panel for a reusable launch vehicle provides enhanced moisture protection, simplified maintenance, and increased temperature resistance. The protective panel includes an outer ceramic matrix composite (CMC) panel, and an insulative bag assembly coupled to the outer CMC panel for isolating the launch vehicle from elevated temperatures and moisture. A standoff attachment system attaches the outer CMC panel and the bag assembly to the primary structure of the launch vehicle. The insulative bag assembly includes a foil bag having a first opening shrink fitted to the outer CMC panel such that the first opening and the outer CMC panel form a water tight seal at temperatures below a desired temperature threshold. Fibrous insulation is contained within the foil bag for protecting the launch vehicle from elevated temperatures. The insulative bag assembly further includes a back panel coupled to a second opening of the foil bag such that the fibrous insulation is encapsulated by the back panel, the foil bag, and the outer CMC panel. The use of a CMC material for the outer panel in conjunction with the insulative bag assembly eliminates the need for waterproofing processes, and ultimately allows for more efficient reentry profiles.

  6. ITER blanket designs

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  7. Thermal convection in a toroidal duct of a liquid metal blanket. Part I. Effect of poloidal magnetic field

    International Nuclear Information System (INIS)

    Zhang, Xuan; Zikanov, Oleg

    2017-01-01

    Highlights: • 2D convection flow develops with internal heating and strong axial magnetic field. • Poloidal magnetic field suppresses turbulence at high Hartmann number. • Flow structure is dominated by large-scale counter-rotation vortices. • Effective heat transfer is maintained by surviving convection structures. - Abstract: We explore the effect of poloidal magnetic field on the thermal convection flow in a toroidal duct of a generic liquid metal blanket. Non-uniform strong heating (the Grashof number up to 10 11 ) arising from the interaction of high-speed neutrons with the liquid breeder, and strong magnetic field (the Hartmann number up to 10 4 ) corresponding to the realistic reactor conditions are considered. The study continues our earlier work , where the problem was solved for a purely toroidal magnetic field and the convection was found to result in two-dimensional turbulence and strong mixing within the duct. Here, we find that the poloidal component of the magnetic field suppresses turbulence, reduces the flow's kinetic energy and high-amplitude temperature fluctuations, and, at high values of Hartmann number, leads to a steady-state flow. At the same time, the intense mixing by the surviving convection structures remains able to maintain effective heat transfer between the liquid metal and the walls.

  8. Development of Processing Techniques for Advanced Thermal Protection Materials

    Science.gov (United States)

    Selvaduray, Guna; Cox, Michael; Srinivasan, Vijayakumar

    1997-01-01

    Thermal Protection Materials Branch (TPMB) has been involved in various research programs to improve the properties and structural integrity of the existing aerospace high temperature materials. Specimens from various research programs were brought into the analytical laboratory for the purpose of obtaining and refining the material characterization. The analytical laboratory in TPMB has many different instruments which were utilized to determine the physical and chemical characteristics of materials. Some of the instruments that were utilized by the SJSU students are: Scanning Electron Microscopy (SEM), Energy Dispersive X-ray analysis (EDX), X-ray Diffraction Spectrometer (XRD), Fourier Transform-Infrared Spectroscopy (FTIR), Ultra Violet Spectroscopy/Visible Spectroscopy (UV/VIS), Particle Size Analyzer (PSA), and Inductively Coupled Plasma Atomic Emission Spectrometer (ICP-AES). The above mentioned analytical instruments were utilized in the material characterization process of the specimens from research programs such as: aerogel ceramics (I) and (II), X-33 Blankets, ARC-Jet specimens, QUICFIX specimens and gas permeability of lightweight ceramic ablators. In addition to analytical instruments in the analytical laboratory at TPMB, there are several on-going experiments. One particular experiment allows the measurement of permeability of ceramic ablators. From these measurements, physical characteristics of the ceramic ablators can be derived.

  9. Sprayable Phase Change Coating Thermal Protection Material

    Science.gov (United States)

    Richardson, Rod W.; Hayes, Paul W.; Kaul, Raj

    2005-01-01

    NASA has expressed a need for reusable, environmentally friendly, phase change coating that is capable of withstanding the heat loads that have historically required an ablative thermal insulation. The Space Shuttle Program currently relies on ablative materials for thermal protection. The problem with an ablative insulation is that, by design, the material ablates away, in fulfilling its function of cooling the underlying substrate, thus preventing the insulation from being reused from flight to flight. The present generation of environmentally friendly, sprayable, ablative thermal insulation (MCC-l); currently use on the Space Shuttle SRBs, is very close to being a reusable insulation system. In actual flight conditions, as confirmed by the post-flight inspections of the SRBs, very little of the material ablates. Multi-flight thermal insulation use has not been qualified for the Space Shuttle. The gap that would have to be overcome in order to implement a reusable Phase Change Coating (PCC) is not unmanageable. PCC could be applied robotically with a spray process utilizing phase change material as filler to yield material of even higher strength and reliability as compared to MCC-1. The PCC filled coatings have also demonstrated potential as cryogenic thermal coatings. In experimental thermal tests, a thin application of PCC has provided the same thermal protection as a much thicker and heavier application of a traditional ablative thermal insulation. In addition, tests have shown that the structural integrity of the coating has been maintained and phase change performance after several aero-thermal cycles was not affected. Experimental tests have also shown that, unlike traditional ablative thermal insulations, PCC would not require an environmental seal coat, which has historically been required to prevent moisture absorption by the thermal insulation, prevent environmental degradation, and to improve the optical and aerodynamic properties. In order to reduce

  10. Thermal power stations and environmental protection

    International Nuclear Information System (INIS)

    Gerking, E.

    1975-01-01

    In this book, the advantages of an optimum cooling concept for waters are compared with the disadvantages of an uncontrolled thermal pollution of waters by waste waters from thermal power plants. The book focuses on the problem of the cost of measures for environmental protection which has not yet received a detailed and complete treatment. The author suggests that perfectionist solutions and superfluos measures be abandoned in favour of a far-reaching, efficient environmental protection concept with a low expenditure of fuel and capital. A detailed treatment is given to false conclusions in the present estimations of the effects of thermal pollution of the waters and to the advantages of freshwater cooling and cooling in general. Also discussed are immission problems and attempts at their solution. (ORU/AK) [de

  11. Environmental protection in thermal power plants

    International Nuclear Information System (INIS)

    1987-01-01

    This workbook is a compilation of the most important facts and data that are relevant today for environmental protection in thermal power plants. Unlike the other issues the text is not in the form of a random collection of data but in the form of a complete presentation. Possible elaboration projects for pupils can be easily derived from the individual sections. These deal with: the discussion about environmental protection; forest decline; sources of emission; nuisances in the Federal Republic of Germany; environmental protection in fossil-fuel power plants - clean air - cooling water utilization and water protection - noise; environmental protection in nuclear power plants - radioactive material produced in nuclear reactors and the retention of such materials - radioactive waste materials - monitoring of radioactive emissions; accessory materials and hints. (orig./HSCH) [de

  12. Thermally induced outdiffusion studies of deuterium in ceramic breeder blanket materials after irradiation

    Energy Technology Data Exchange (ETDEWEB)

    González, Maria, E-mail: maria.gonzalez@ciemat.es [LNF-CIEMAT, Materials for Fusion Group, Madrid (Spain); Carella, Elisabetta; Moroño, Alejandro [LNF-CIEMAT, Materials for Fusion Group, Madrid (Spain); Kolb, Matthias H.H.; Knitter, Regina [Karlsruhe Institute of Technology, Institute for Applied Materials (IAM-WPT), Karlsruhe (Germany)

    2015-10-15

    Highlights: • Surface defects in Lithium-based ceramics are acting as trapping centres for deuterium. • Ionizing radiation affects the deuterium sorption and desorption processes. • By extension, the release of the tritium produced in a fusion breeder will be effective. - Abstract: Based on a KIT–CIEMAT collaboration on the radiation damage effects of light ions sorption/desorption in ceramic breeder materials, candidate materials for the ITER EU TBM were tested for their outgassing behavior as a function of temperature and radiation. Lithium orthosilicate based pebbles with different metatitanate contents and pellets of the individual oxide components were exposed to a deuterium atmosphere at room temperature. Then the thermally induced release of deuterium gas was registered up to 800 °C. This as-received behavior was studied in comparison with that after exposing the deuterium-treated samples to 4 MGy total dose of gamma radiation. The thermal desorption spectra reveal differences in deuterium sorption/desorption behavior depending on the composition and the induced ionizing damage. In these breeder candidates, strong desorption rate at approx. 300 °C takes place, which slightly increases with increasing amount of the titanate second phase. For all studied materials, ionizing radiation induces electronic changes disabling a number of trapping centers for D{sub 2} adsorption.

  13. The thermal response of the first wall of a fusion reactor blanket to plasma disruptions

    International Nuclear Information System (INIS)

    Klippel, H.Th.

    1983-09-01

    Major plasma disruptions in Tokamak power reactors are potentially dangerous because high thermal overloading of the first wall may occur, resulting in melting and evaporation. The present uncertainties of the disruption characteristics, in particular the space and time dependence of the energy deposition, lead to a wide variation in the prospective surface energy loads. The thermal response of a first wall of aluminium, stainless steel and of graphite subjected to disruption energy loads up to 1000 J cm -2 has been analysed including the effects of melting and surface evaporation, vapour recondensation, vapour shielding, and the moving of the surface boundary caused by the evaporation. A special calculation model has been developed for this purpose. The main results are the following: by values of local transient energy depositions over 1500 J cm -2 bare stainless steel walls are damaged severely. Further calculations are needed to estimate the endurance limit of several candidate first wall materials. Applications of coatings on surfaces need special attention. For the reference INTOR disruption (approx. 100 J cm -2 ) evaporation is not significant. The effect of vapour shielding on evaporation has been found to be significant. The effect on melting is less pronounced. In a complete analysis the stability and dynamic behaviour of the melted layer under electromagnetic forces should be included. Also a reliable set of plasma disruption characteristics should be gathered

  14. Proposal for the award of a blanket purchase contract for the design, supply, installation and maintenance of automatic fire-detection, fire-protection and voice-alarm systems for the Super Proton Synchrotron

    CERN Document Server

    2017-01-01

    Proposal for the award of a blanket purchase contract for the design, supply, installation and maintenance of automatic fire-detection, fire-protection and voice-alarm systems for the Super Proton Synchrotron

  15. INTOR first wall/blanket/shield activity

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  16. Magnetoconvection in HCLL blankets

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Buehler, L.

    2014-01-01

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

  17. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

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

  18. Advanced materials for thermal protection system

    Science.gov (United States)

    Heng, Sangvavann; Sherman, Andrew J.

    1996-03-01

    Reticulated open-cell ceramic foams (both vitreous carbon and silicon carbide) and ceramic composites (SiC-based, both monolithic and fiber-reinforced) were evaluated as candidate materials for use in a heat shield sandwich panel design as an advanced thermal protection system (TPS) for unmanned single-use hypersonic reentry vehicles. These materials were fabricated by chemical vapor deposition/infiltration (CVD/CVI) and evaluated extensively for their mechanical, thermal, and erosion/ablation performance. In the TPS, the ceramic foams were used as a structural core providing thermal insulation and mechanical load distribution, while the ceramic composites were used as facesheets providing resistance to aerodynamic, shear, and erosive forces. Tensile, compressive, and shear strength, elastic and shear modulus, fracture toughness, Poisson's ratio, and thermal conductivity were measured for the ceramic foams, while arcjet testing was conducted on the ceramic composites at heat flux levels up to 5.90 MW/m2 (520 Btu/ft2ṡsec). Two prototype test articles were fabricated and subjected to arcjet testing at heat flux levels of 1.70-3.40 MW/m2 (150-300 Btu/ft2ṡsec) under simulated reentry trajectories.

  19. Outer skin protection of columbium Thermal Protection System (TPS) panels

    Science.gov (United States)

    Culp, J. D.

    1973-01-01

    A coated columbium alloy material system 0.04 centimeter thick was developed which provides for increased reliability to the load bearing character of the system in the event of physical damage to and loss of the exterior protective coating. The increased reliability to the load bearing columbium alloy (FS-85) was achieved by interposing an oxidation resistant columbium alloy (B-1) between the FS-85 alloy and a fused slurry silicide coating. The B-1 alloy was applied as a cladding to the FS-85 and the composite was fused slurry silicide coated. Results of material evaluation testing included cyclic oxidation testing of specimens with intentional coating defects, tensile testing of several material combinations exposed to reentry profile conditions, and emittance testing after cycling of up to 100 simulated reentries. The clad material, which was shown to provide greater reliability than unclad materials, holds significant promise for use in the thermal protection system of hypersonic reentry vehicles.

  20. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  1. Thermal Protection Systems: Past, Present and Future

    Science.gov (United States)

    Johnson, Sylvia M.

    2015-01-01

    Thermal protection materials and systems (TPS) have been critical to fulfilling humankinds desire to explore space. Composite and ceramic materials have enabled the early missions to orbit, the moon, the space station, Mars with robots, and sample return. Crewed missions to Mars are being considered, and this places even more demands on TPS materials. This talk will give some history on the materials used for earth and planetary entry and the demands placed upon such materials. TPS needs for future missions, especially to Mars, will be identified and potential solutions discussed.

  2. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-04-01

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

  3. 3D Multifunctional Ablative Thermal Protection System

    Science.gov (United States)

    Feldman, Jay; Venkatapathy, Ethiraj; Wilkinson, Curt; Mercer, Ken

    2015-01-01

    NASA is developing the Orion spacecraft to carry astronauts farther into the solar system than ever before, with human exploration of Mars as its ultimate goal. One of the technologies required to enable this advanced, Apollo-shaped capsule is a 3-dimensional quartz fiber composite for the vehicle's compression pad. During its mission, the compression pad serves first as a structural component and later as an ablative heat shield, partially consumed on Earth re-entry. This presentation will summarize the development of a new 3D quartz cyanate ester composite material, 3-Dimensional Multifunctional Ablative Thermal Protection System (3D-MAT), designed to meet the mission requirements for the Orion compression pad. Manufacturing development, aerothermal (arc-jet) testing, structural performance, and the overall status of material development for the 2018 EM-1 flight test will be discussed.

  4. Thermal Protection Test Bed Pathfinder Development Project

    Science.gov (United States)

    Snapp, Cooper

    2015-01-01

    In order to increase thermal protection capabilities for future reentry vehicles, a method to obtain relevant test data is required. Although arc jet testing can be used to obtain some data on materials, the best method to obtain these data is to actually expose them to an atmospheric reentry. The overprediction of the Orion EFT-1 flight data is an example of how the ground test to flight traceability is not fully understood. The RED-Data small reentry capsule developed by Terminal Velocity Aerospace is critical to understanding this traceability. In order to begin to utilize this technology, ES3 needs to be ready to build and integrate heat shields onto the RED-Data vehicle. Using a heritage Shuttle tile material for the heat shield will both allow valuable insight into the environment that the RED-Data vehicle can provide and give ES3 the knowledge and capability to build and integrate future heat shields for this vehicle.

  5. Basic Theoretical Principles Pertaining to Thermal Protection of Oil Transformer

    Directory of Open Access Journals (Sweden)

    O. G. Shirokov

    2008-01-01

    Full Text Available The paper contains formulation of basic theoretical principles pertaining to thermal protection of an oil transformer in accordance with classical theory of relay protection and theory of diagnostics with the purpose of unification of terminological and analytical information which is presently available in respect of this problem. Classification of abnormal thermal modes of an oil transformer and also algorithms and methods for operation of diagnostic thermal protection of a transformer have been proposed.

  6. The second advanced lead lithium blanket concept using ODS steel as structural material and SiCf/SiC flow channel inserts as electrical and thermal insulators (Task PPA 2.5). Final report

    International Nuclear Information System (INIS)

    Norajitra, P.; Buehler, L.; Fischer, U.

    1999-12-01

    Preparatory work on the advanced dual coolant (A-DCL) blanket concept using SiC f /SiC flow channel inserts as electrical and thermal insulators has been carried out at the Forschungszentrum Karlsruhe in co-operation with CEA as a conceptual design proposal to the EU fusion power plant study planned to be launched in 2000 within the framework of the EU fusion programme with the main objective of specifying the characteristics of an attractive and viable commercial D-T fusion power plant. The basic principles and design characteristics of this A-DCL blanket concept are presented and its potential with regard to performance (neutron wall load, lifetime, availability) is discussed in this report. The results of this study show that the A-DCL blanket concept has a high potential for further development due to its high thermal efficiency and its simple concept solution. (orig.) [de

  7. The usage of phase change materials in fire fighter protective clothing: its effect on thermal protection

    Science.gov (United States)

    Zhao, Mengmeng

    2017-12-01

    The thermal protective performance of the fire fighter protective clothing is of vital importance for fire fighters. In the study fabrics treated by phase change materials (PCMs) were applied in the multi-layered fabrics of the fire fighter protective clothing ensemble. The PCM fabrics were placed at the different layers of the clothing and their thermal protective performance were measured by a TPP tester. Results show that with the application of the PCM fabrics the thermal protection of the multi-layered fabrics was greatly increased. The time to reach a second degree burn was largely reduced. The location of the PCM fabrics at the different layers did not affect much on the thermal protective performance. The higher amount of the PCM adds on, the higher thermal protection was brought. The fabrics with PCMs of a higher melting temperature could contribute to higher thermal protection.

  8. The preliminary thermal-hydraulic design of one superheated steam water cooled blanket concept based on RELAP5 and MELCOR codes - 15147

    International Nuclear Information System (INIS)

    Guo, Y.; Wang, G.; Cheng, Y.; Peng, C.

    2015-01-01

    Water Cooled Blanket (WCB) is very important in the concept design and energy transfer in future fusion power plant. One concept design of WCB is under computational testing. RELAP5 and MELCOR codes, which are mature and often used in nuclear engineering, are selected as simulation tools. The complex inner flow channels and heat sources are simplified according to its thermal-hydraulic characteristics. Then the nodal models for RELAP5 and MELCOR are built for approximating the concept design. The superheated steam scheme is analyzed by two codes separately under different power levels. After some adjustments of the inlet flow resistance coefficients of some flow channels, the reasonable stable conditions can be obtained. The stable fluid and wall temperature distributions and pressure drops are studied. The results of two codes are compared and some advices are given. (authors)

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

    International Nuclear Information System (INIS)

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

    1989-09-01

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

  10. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  11. Displacements of Metallic Thermal Protection System Panels During Reentry

    Science.gov (United States)

    Daryabeigi, Kamran; Blosser, Max L.; Wurster, Kathryn E.

    2006-01-01

    Bowing of metallic thermal protection systems for reentry of a previously proposed single-stage-to-orbit reusable launch vehicle was studied. The outer layer of current metallic thermal protection system concepts typically consists of a honeycomb panel made of a high temperature nickel alloy. During portions of reentry when the thermal protection system is exposed to rapidly varying heating rates, a significant temperature gradient develops across the honeycomb panel thickness, resulting in bowing of the honeycomb panel. The deformations of the honeycomb panel increase the roughness of the outer mold line of the vehicle, which could possibly result in premature boundary layer transition, resulting in significantly higher downstream heating rates. The aerothermal loads and parameters for three locations on the centerline of the windward side of this vehicle were calculated using an engineering code. The transient temperature distributions through a metallic thermal protection system were obtained using 1-D finite volume thermal analysis, and the resulting displacements of the thermal protection system were calculated. The maximum deflection of the thermal protection system throughout the reentry trajectory was 6.4 mm. The maximum ratio of deflection to boundary layer thickness was 0.032. Based on previously developed distributed roughness correlations, it was concluded that these defections will not result in tripping the hypersonic boundary layer.

  12. Kinetic Integrated Thermal Protection System (KnITPS)

    Data.gov (United States)

    National Aeronautics and Space Administration — Use the flexibility and shape formation possibilities inherent in knitting to form thermal protection systems that can be custom fitted to a heat shield carrier...

  13. NDE for Ablative Thermal Protection Systems, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This program addresses the need for non-destructive evaluation (NDE) methods for quality assessment and defect evaluation of thermal protection systems (TPS),...

  14. NDE for Ablative Thermal Protection Systems, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — This program addresses the need for non-destructive evaluation (NDE) methods for quality assessment and defect evaluation of thermal protection systems (TPS). Novel...

  15. Thermal Protection System Materials (TPSM): 3D MAT

    Data.gov (United States)

    National Aeronautics and Space Administration — The 3D MAT Project seeks to design and develop a game changing Woven Thermal Protection System (TPS) technology tailored to meet the needs of the Orion Multi-Purpose...

  16. Theoretical prediction of thermal conductivity for thermal protection systems

    International Nuclear Information System (INIS)

    Gori, F.; Corasaniti, S.; Worek, W.M.; Minkowycz, W.J.

    2012-01-01

    The present work is aimed to evaluate the effective thermal conductivity of an ablative composite material in the state of virgin material and in three paths of degradation. The composite material is undergoing ablation with formation of void pores or char and void pores. The one dimensional effective thermal conductivity is evaluated theoretically by the solution of heat conduction under two assumptions, i.e. parallel isotherms and parallel heat fluxes. The paper presents the theoretical model applied to an elementary cubic cell of the composite material which is made of two crossed fibres and a matrix. A numerical simulation is carried out to compare the numerical results with the theoretical ones for different values of the filler volume fraction. - Highlights: ► Theoretical models of the thermal conductivity of an ablative composite. ► Composite material is made of two crossed fibres and a matrix. ► Three mechanisms of degradation are investigated. ► One dimensional thermal conductivity is evaluated by the heat conduction equation. ► Numerical simulations to be compared with the theoretical models.

  17. Aero-thermal optimization of in-flight electro-thermal ice protection systems in transient de-icing mode

    International Nuclear Information System (INIS)

    Pourbagian, Mahdi; Habashi, Wagdi G.

    2015-01-01

    Highlights: • We introduce an efficient methodology for the optimization of a de-icing system. • We can replace the expensive CHT simulation by ROM without loosing much accuracy. • We propose different criteria affecting the energy usage and aerodynamic performance. • These criteria can significantly improve the performance of the de-icing system. - Abstract: Even if electro-thermal ice protection systems (IPS) consume less energy when operating in de-icing mode than in anti-icing mode, they still need to be optimized for energy usage. The optimization, however, should also take into account the effect of the de-icing system on the aerodynamic performance. The present work offers an optimization framework in which both thermal and aerodynamic viewpoints are taken into account in formulating various objective and constraint functions by considering the energy consumption, the thickness, the volume, the shape and the location of the accreted ice on the surface as the key parameters affecting the energy usage and the aerodynamic performance. The design variables include the power density and the activation time of the electric heating blankets. A derivative-free technique, called the mesh adaptive direct search (MADS) method, is used to carry out the optimization process, which would normally need a large number of unsteady conjugate heat transfer (CHT) calculations for the IPS simulation. To avoid such prohibitive computations, reduced-order modeling (ROM) is used to construct simplified low-dimensional CHT models. The approach is illustrated through several test cases, in which different combinations of objective and constraint functions, design variables and cycling sequence patterns are examined. In these test cases, the energy consumption is significantly reduced compared to the experiments by improving the spatial and temporal distribution of the thermal energy usage. The results show the benefits of the approach in bringing energy, safety and

  18. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    Proust, E.; Giancarli, L.

    1992-01-01

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

  19. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  1. Thermonuclear blankets

    International Nuclear Information System (INIS)

    Kubota, Tadashi.

    1986-01-01

    Purpose: To increase the effective thermal conductivity between lithium ceramic spheres and a metal container. Constitution: The surface of lithium ceramic spheres is coated with a nickel metal, which is further coated with a thin copper layer. Then, copper spheres with a diameter smaller than that of the lithium ceramic spheres are packed in admixture together with the lithium ceramic spheres in an appropriate volume ratio. Since, copper as a relatively soft metal is coated on the surface of the lithium ceramic spheres and the copper spheres are charged to the gaps between each of the lithium ceramic spheres, the area of contact between the lithium ceramic spheres to each other and that between the lithium ceramic spheres and the metal container are easily increased to improve the effective thermal conductivity, by which the heat removing performance of the plant can be improved. (Yoshino, Y.)

  2. Study of the multiplication and kinetic effects of salt mixtures and salt blanket micromodels on thermal neutron spectra of heavy water MAKET facility

    International Nuclear Information System (INIS)

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

    2009-10-01

    The main goal of the Project is to study and evaluate nuclear characteristics of materials and isotopes involved in processes of irradiated nuclear fuel transmutation. This principal task is subdivided into 9 subtasks subject to the neutron or proton source used, the type of the nuclear process under study, isotope collection, characteristics of which are to be investigated, etc. In the presented extract of the Project Activity report the measurements there were used the MAKET zero-power heavy-water reactor in the measurements there was employed a large set of minor actinide samples highly enriched with the main isotope. The samples were obtained with mass-separator SM-2 (VNIIEF). At the heavy-water reactor MAKET (ITEP) there were measured multiplying and kinetic characteristics of salt mixtures basing on the spectra of fast and thermal neutrons. The salt mixtures of zirconium and sodium fluorides were available in salt blanket models (SBM) of cylindrical shape. There were measured the neutron spectra formed by this micro-model as well as the effective fission cross-sections of neptunium, plutonium, americium and curium isotopes caused by SBM neutrons. The neutron spectra in the measurement positions were determined from activation reaction rates. (author)

  3. Materials for breeding blankets

    International Nuclear Information System (INIS)

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

    1995-09-01

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

  4. Materials for breeding blankets

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. A Novel Adjustable Concept for Permeable Gas/Vapor Protective Clothing: Balancing Protection and Thermal Strain.

    Science.gov (United States)

    Bogerd, Cornelis Peter; Langenberg, Johannes Pieter; DenHartog, Emiel A

    2018-02-13

    Armed forces typically have personal protective clothing (PPC) in place to offer protection against chemical, biological, radiological and nuclear (CBRN) agents. The regular soldier is equipped with permeable CBRN-PPC. However, depending on the operational task, these PPCs pose too much thermal strain to the wearer, which results in a higher risk of uncompensable heat stress. This study investigates the possibilities of adjustable CBRN-PPC, consisting of different layers that can be worn separately or in combination with each other. This novel concept aims to achieve optimization between protection and thermal strain during operations. Two CBRN-PPC (protective) layers were obtained from two separate manufacturers: (i) a next-to-skin (NTS) and (ii) a low-burden battle dress uniform (protective BDU). In addition to these layers, a standard (non-CBRN protective) BDU (sBDU) was also made available. The effect of combining clothing layers on the levels of protection were investigated with a Man-In-Simulant Test. Finally, a mechanistic numerical model was employed to give insight into the thermal burden of the evaluated CBRN-PPC concepts. Combining layers results in substantially higher protection that is more than the sum of the individual layers. Reducing the airflow on the protective layer closest to the skin seems to play an important role in this, since combining the NTS with the sBDU also resulted in substantially higher protection. As expected, the thermal strain posed by the different clothing layer combinations decreases as the level of protection decreases. This study has shown that the concept of adjustable protection and thermal strain through multiple layers of CBRN-PPC works. Adjustable CBRN-PPC allows for optimization of the CBRN-PPC in relation to the threat level, thermal environment, and tasks at hand in an operational setting. © The Author(s) 2017. Published by Oxford University Press on behalf of the British Occupational Hygiene Society.

  6. Study of skin model and geometry effects on thermal performance of thermal protective fabrics

    Science.gov (United States)

    Zhu, Fanglong; Ma, Suqin; Zhang, Weiyuan

    2008-05-01

    Thermal protective clothing has steadily improved over the years as new materials and improved designs have reached the market. A significant method that has brought these improvements to the fire service is the NFPA 1971 standard on structural fire fighters’ protective clothing. However, this testing often neglects the effects of cylindrical geometry on heat transmission in flame resistant fabrics. This paper deals with methods to develop cylindrical geometry testing apparatus incorporating novel skin bioheat transfer model to test flame resistant fabrics used in firefighting. Results show that fabrics which shrink during the test can have reduced thermal protective performance compared with the qualities measured with a planar geometry tester. Results of temperature differences between skin simulant sensors of planar and cylindrical tester are also compared. This test method provides a new technique to accurately and precisely characterize the thermal performance of thermal protective fabrics.

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

    International Nuclear Information System (INIS)

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

    1979-09-01

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

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

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

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

  9. Active Wireless Temperature Sensors for Aerospace Thermal Protection Systems

    Science.gov (United States)

    Milos, Frank S.; Karunaratne, K.; Arnold, Jim (Technical Monitor)

    2002-01-01

    Health diagnostics is an area where major improvements have been identified for potential implementation into the design of new reusable launch vehicles in order to reduce life-cycle costs, to increase safety margins, and to improve mission reliability. NASA Ames is leading the effort to advance inspection and health management technologies for thermal protection systems. This paper summarizes a joint project between NASA Ames and Korteks to develop active wireless sensors that can be embedded in the thermal protection system to monitor sub-surface temperature histories. These devices are thermocouples integrated with radio-frequency identification circuitry to enable acquisition and non-contact communication of temperature data through aerospace thermal protection materials. Two generations of prototype sensors are discussed. The advanced prototype collects data from three type-k thermocouples attached to a 2.54-cm square integrated circuit.

  10. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  11. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  12. Alternative High Performance Polymers for Ablative Thermal Protection Systems

    Science.gov (United States)

    Boghozian, Tane; Stackpoole, Mairead; Gonzales, Greg

    2015-01-01

    Ablative thermal protection systems are commonly used as protection from the intense heat during re-entry of a space vehicle and have been used successfully on many missions including Stardust and Mars Science Laboratory both of which used PICA - a phenolic based ablator. Historically, phenolic resin has served as the ablative polymer for many TPS systems. However, it has limitations in both processing and properties such as char yield, glass transition temperature and char stability. Therefore alternative high performance polymers are being considered including cyanate ester resin, polyimide, and polybenzoxazine. Thermal and mechanical properties of these resin systems were characterized and compared with phenolic resin.

  13. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  14. Damage Detection/Locating System Providing Thermal Protection

    Science.gov (United States)

    Woodard, Stanley E. (Inventor); Jones, Thomas W. (Inventor); Taylor, Bryant D. (Inventor); Qamar, A. Shams (Inventor)

    2010-01-01

    A damage locating system also provides thermal protection. An array of sensors substantially tiles an area of interest. Each sensor is a reflective-surface conductor having operatively coupled inductance and capacitance. A magnetic field response recorder is provided to interrogate each sensor before and after a damage condition. Changes in response are indicative of damage and a corresponding location thereof.

  15. Thermal limits for passive safety of fusion reactors

    International Nuclear Information System (INIS)

    Kazimi, M.S.; Massidda, J.E.; Oshima, M.

    1989-01-01

    The thermal response of the first wall and blanket due to power/cooling mismatch in the absence of operation action is examined. The analyses of coolant and power transients are carried out on six reference blanket designs representing a broad range of fusion first wall and blanket technology. It is concluded that the requirement of plant protection will impose sufficiently stringent peak neutron wall loading limits to avoid a serious threat to the public. It is found that for the D-T design,s the operating wall loading may have to be limited to 3 - 8 MW/m/sup 2/ for passive plant protection, depending on the plant design

  16. Fusion fuel blanket technology

    International Nuclear Information System (INIS)

    Hastings, I.J.; Gierszewski, P.

    1987-05-01

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

  17. Blanket testing in NET

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  18. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

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

  20. Impact Testing of Orbiter Thermal Protection System Materials

    Science.gov (United States)

    Kerr, Justin

    2006-01-01

    This viewgraph presentation reviews the impact testing of the materials used in designing the shuttle orbiter thermal protection system (TPS). Pursuant to the Columbia Accident Investigation Board recommendations a testing program of the TPS system was instituted. This involved using various types of impactors in different sizes shot from various sizes and strengths guns to impact the TPS tiles and the Leading Edge Structural Subsystem (LESS). The observed damage is shown, and the resultant lessons learned are reviewed.

  1. Heat Transfer Analysis of Thermal Protection Structures for Hypersonic Vehicles

    Science.gov (United States)

    Zhou, Chen; Wang, Zhijin; Hou, Tianjiao

    2017-11-01

    This research aims to develop an analytical approach to study the heat transfer problem of thermal protection systems (TPS) for hypersonic vehicles. Laplace transform and integral method are used to describe the temperature distribution through the TPS subject to aerodynamic heating during flight. Time-dependent incident heat flux is also taken into account. Two different cases with heat flux and radiation boundary conditions are studied and discussed. The results are compared with those obtained by finite element analyses and show a good agreement. Although temperature profiles of such problems can be readily accessed via numerical simulations, analytical solutions give a greater insight into the physical essence of the heat transfer problem. Furthermore, with the analytical approach, rapid thermal analyses and even thermal optimization can be achieved during the preliminary TPS design.

  2. Thermal assault and polyurethane foam-evaluating protective mechanisms

    International Nuclear Information System (INIS)

    Williamson, C.L.; Iams, Z.L.

    2004-01-01

    Rigid polyurethane foam utilizes a variety of mechanisms to mitigate the thermal assault of a ''regulatory burn''. Polymer specific heat and foam k-factor are of limited usefulness in predicting payload protection. Properly formulated rigid polyurethane foam provides additional safeguards by employing ablative mechanisms which are effective even when the foam has been crushed or fractured as a result of trauma. The dissociative transitions from polymer to gas and char, and the gas transport of heat from inside the package out into the environment are also thermal mitigators. Additionally, the in-situ production of an intumescent, insulative, carbonaceous char, confers thermal protection even when a package's outer steel skin has been breached. In this test program, 19 liter, ''Five gallon'' steel pails are exposed on one end to the flame of an ''Oil Burner'' as described in the US Federal Aviation Administration (FAA) ''Aircraft Materials Fire Test Handbook''. When burning 2 diesel at a nominal rate of 8.39 kg (18.5 pounds)/hr, the burner generates a high emissivity flame that impinges on the pail face with the thermal intensity of a full scale pool-fire environment. Results of these tests, TGA and MDSC analysis on the subject foams are reported, and their relevance to full size packages and pool fires are discussed

  3. Rigid Polyurethane Foam Thermal Insulation Protected with Mineral Intumescent Mat

    Directory of Open Access Journals (Sweden)

    Kirpluks Mikelis

    2014-12-01

    Full Text Available One of the biggest disadvantages of rigid polyurethane (PU foams is its low thermal resistance, high flammability and high smoke production. Greatest advantage of this thermal insulation material is its low thermal conductivity (λ, which at 18-28 mW/(m•K is superior to other materials. To lower the flammability of PU foams, different flame retardants (FR are used. Usually, industrially viable are halogenated liquid FRs but recent trends in EU regulations show that they are not desirable any more. Main concern is toxicity of smoke and health hazard form volatiles in PU foam materials. Development of intumescent passive fire protection for foam materials would answer problems with flammability without using halogenated FRs. It is possible to add expandable graphite (EG into PU foam structure but this increases the thermal conductivity greatly. Thus, the main advantage of PU foam is lost. To decrease the flammability of PU foams, three different contents 3%; 9% and 15% of EG were added to PU foam formulation. Sample with 15% of EG increased λ of PU foam from 24.0 to 30.0 mW/(m•K. This paper describes the study where PU foam developed from renewable resources is protected with thermally expandable intumescent mat from Technical Fibre Products Ltd. (TFP as an alternative to EG added into PU material. TFP produces range of mineral fibre mats with EG that produce passive fire barrier. Two type mats were used to develop sandwich-type PU foams. Also, synergy effect of non-halogenated FR, dimethyl propyl phosphate and EG was studied. Flammability of developed materials was assessed using Cone Calorimeter equipment. Density, thermal conductivity, compression strength and modulus of elasticity were tested for developed PU foams. PU foam morphology was assessed from scanning electron microscopy images.

  4. Tritium breeding blanket

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  5. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

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

  6. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

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

  7. 46 CFR 199.214 - Immersion suits and thermal protective aids.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Immersion suits and thermal protective aids. 199.214... Passenger Vessels § 199.214 Immersion suits and thermal protective aids. (a) Each passenger vessel must... an immersion suit. (c) The immersion suits and thermal protective aids required under paragraphs (a...

  8. Thermal protection system gap analysis using a loosely coupled fluid-structural thermal numerical method

    Science.gov (United States)

    Huang, Jie; Li, Piao; Yao, Weixing

    2018-05-01

    A loosely coupled fluid-structural thermal numerical method is introduced for the thermal protection system (TPS) gap thermal control analysis in this paper. The aerodynamic heating and structural thermal are analyzed by computational fluid dynamics (CFD) and numerical heat transfer (NHT) methods respectively. An interpolation algorithm based on the control surface is adopted for the data exchanges on the coupled surface. In order to verify the analysis precision of the loosely coupled method, a circular tube example was analyzed, and the wall temperature agrees well with the test result. TPS gap thermal control performance was studied by the loosely coupled method successfully. The gap heat flux is mainly distributed in the small region at the top of the gap which is the high temperature region. Besides, TPS gap temperature and the power of the active cooling system (CCS) calculated by the traditional uncoupled method are higher than that calculated by the coupled method obviously. The reason is that the uncoupled method doesn't consider the coupled effect between the aerodynamic heating and structural thermal, however the coupled method considers it, so TPS gap thermal control performance can be analyzed more accurately by the coupled method.

  9. Thermal-Acoustic Fatigue of a Multilayer Thermal Protection System in Combined Extreme Environments

    Directory of Open Access Journals (Sweden)

    Liu Liu

    2014-06-01

    Full Text Available In order to ensure integrity of thermal protection system (TPS structure for hypersonic vehicles exposed to severe operating environments, a study is undertaken to investigate the response and thermal-acoustic fatigue damage of a representative multilayer TPS structure under combined thermal and acoustic loads. An unsteady-state flight of a hypersonic vehicle is composed of a series of steady-state snapshots, and for each snapshot an acoustic load is imposed to a static steady-state TPS structure. A multistep thermal-acoustic fatigue damage intensity analysis procedure is given and consists of a heat transfer analysis, a nonlinear thermoelastic analysis, and a random response analysis under a combined loading environment and the fatigue damage intensity has been evaluated with two fatigue analysis techniques. The effects of thermally induced deterministic stress and nondeterministic dynamic stress due to the acoustic loading have been considered in the damage intensity estimation with a maximum stress fatigue model. The results show that the given thermal-acoustic fatigue intensity estimation procedure is a viable approach for life prediction of TPS structures under a typical mission cycle with combined loadings characterized by largely different time-scales. A discussion of the effects of the thermal load, the acoustic load, and fatigue analysis methodology on the fatigue damage intensity has been provided.

  10. Transpiration cooling assisted ablative thermal protection of aerospace substructures

    International Nuclear Information System (INIS)

    Khan, M.B.; Iqbal, N.; Haider, Z.

    2009-01-01

    Ablatives are heat-shielding materials used to protect aerospace substructures. These materials are sacrificial in nature and provide protection primarily through the large endothermic transformation during exposure to hyper thermal environment such as encountered in re-entry modules. The performance of certain ablatives was reported in terms of their TGA/DTA in Advanced Materials-97 (pp 57-65). The focus of this earlier research resided in the consolidation of interface between the refractory inclusion and the host polymeric matrix to improve thermal resistance. In the present work we explore the scope of transpiration cooling in ablative performance through flash evaporation of liquid incorporated in the host EPDM (Ethylene Propylene Diene Monomer) matrix. The compression-molded specimens were exposed separately to plasma flame (15000 C) and oxyacetylene torch (3000 C) and the back face transient temperature is recorded in situ employing a thermocouple/data logger system. Both head on impingement (HOI) and parallel flow (PF) through a central cavity in the ablator were used. It is observed that transpiration cooling is effective and yields (a) rapid thermal equilibrium in the specimen, (b) lower back face temperature and (c) lower ablation rate, compared to conventional ablatives. SEM/EDS analysis is presented to amplify the point. (author)

  11. Fusion blanket high-temperature heat transfer

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1983-01-01

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

  12. Thermal Protective Coating for High Temperature Polymer Composites

    Science.gov (United States)

    Barron, Andrew R.

    1999-01-01

    The central theme of this research is the application of carboxylate-alumoxane nanoparticles as precursors to thermally protective coatings for high temperature polymer composites. In addition, we will investigate the application of carboxylate-alumoxane nanoparticle as a component to polymer composites. The objective of this research was the high temperature protection of polymer composites via novel chemistry. The significance of this research is the development of a low cost and highly flexible synthetic methodology, with a compatible processing technique, for the fabrication of high temperature polymer composites. We proposed to accomplish this broad goal through the use of a class of ceramic precursor material, alumoxanes. Alumoxanes are nano-particles with a boehmite-like structure and an organic periphery. The technical goals of this program are to prepare and evaluate water soluble carboxylate-alumoxane for the preparation of ceramic coatings on polymer substrates. Our proposed approach is attractive since proof of concept has been demonstrated under the NRA 96-LeRC-1 Technology for Advanced High Temperature Gas Turbine Engines, HITEMP Program. For example, carbon and Kevlar(tm) fibers and matting have been successfully coated with ceramic thermally protective layers.

  13. Ballistic Performance of Porous-Ceramic, Thermal-Protection-Systems

    Science.gov (United States)

    Christiansen, E. L.; Davis, B. A.; Miller, J. E.; Bohl, W. E.; Foreman, C. D.

    2009-01-01

    Porous-ceramic, thermal protection systems are used heavily in current reentry vehicles like the Space Shuttle and are currently being proposed for the next generation of manned spacecraft, Orion. These materials insulate the structural components of a spacecraft against the intense thermal environments of atmospheric reentry. Furthermore, these materials are also highly exposed to space environmental hazards like meteoroid and orbital debris impacts. This paper discusses recent impact testing up to 9 km/s, and the findings of the influence of material equation-of-state on the simulation of the impact event to characterize the ballistic performance of these materials. These results will be compared with heritage models1 for these materials developed from testing at lower velocities. Assessments of predicted spacecraft risk based upon these tests and simulations will also be discussed.

  14. Preliminary design of the thermal protection system for solar probe

    Science.gov (United States)

    Dirling, R. B., Jr.; Loomis, W. C.; Heightland, C. N.

    1982-01-01

    A preliminary design of the thermal protection system for the NASA Solar Probe spacecraft is presented. As presently conceived, the spacecraft will be launched by the Space Shuttle on a Jovian swing-by trajectory and at perihelion approach to three solar radii of the surface of the Earth's sun. The system design satisfies maximum envelope, structural integrity, equipotential, and mass loss/contamination requirements by employing lightweight carbon-carbon emissive shields. The primary shield is a thin shell, 15.5-deg half-angle cone which absorbs direct solar flux at up to 10-deg off-nadir spacecraft pointing angles. Secondary shields of sandwich construction and low thickness-direction thermal conductivity are used to reduce the primary shield infrared radiation to the spacecraft payload.

  15. An assessment of the base blanket for ITER

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  16. Thermal stress analysis of space shuttle orbiter wing skin panel and thermal protection system

    Science.gov (United States)

    Ko, William L.; Jenkins, Jerald M.

    1987-01-01

    Preflight thermal stress analysis of the space shuttle orbiter wing skin panel and the thermal protection system (TPS) was performed. The heated skin panel analyzed was rectangular in shape and contained a small square cool region at its center. The wing skin immediately outside the cool region was found to be close to the state of elastic instability in the chordwise direction based on the conservative temperature distribution. The wing skin was found to be quite stable in the spanwise direction. The potential wing skin thermal instability was not severe enough to tear apart the strain isolation pad (SIP) layer. Also, the preflight thermal stress analysis was performed on the TPS tile under the most severe temperature gradient during the simulated reentry heating. The tensile thermal stress induced in the TPS tile was found to be much lower than the tensile strength of the TPS material. The thermal bending of the TPS tile was not severe enough to cause tearing of the SIP layer.

  17. Ballistic Performance of Porous-Ceramic, Thermal Protection Systems

    Science.gov (United States)

    Miller, J. E.; Bohl, W. E.; Christiansen, Eric C.; Davis, B. A.; Foreman, C. D.

    2011-01-01

    Porous-ceramic, thermal protection systems are used heavily in current reentry vehicles like the Orbiter, and they are currently being proposed for the next generation of US manned spacecraft, Orion. These systems insulate reentry critical components of a spacecraft against the intense thermal environments of atmospheric reentry. Additionally, these materials are highly exposed to space environment hazards like solid particle impacts. This paper discusses impact studies up to 10 km/s on 8 lb/cu ft alumina-fiber-enhanced-thermal-barrier (AETB8) tiles coated with a toughened-unipiece-fibrous-insulation/ reaction-cured-glass layer (TUFI/RCG). A semi-empirical, first principals impact model that describes projectile dispersion is described that provides excellent agreement with observations over a broad range of impact velocities, obliquities and projectile materials. Model extensions to look at the implications of greater than 10 GPa equation of state is also discussed. Predicted penetration probabilities for a vehicle visiting the International Space Station is 60% lower for orbital debris and 95% lower for meteoroids with this model compared to an energy scaled approach.

  18. Self-cooled liquid-metal blanket concept

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

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

  20. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

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

    1997-05-01

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

  1. Novel blanket design for ICTR's

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  2. Integrated Thermal Protection Systems and Heat Resistant Structures

    Science.gov (United States)

    Pichon, Thierry; Lacoste, Marc; Glass, David E.

    2006-01-01

    In the early stages of NASA's Exploration Initiative, Snecma Propulsion Solide was funded under the Exploration Systems Research & Technology program to develop integrated thermal protection systems and heat resistant structures for reentry vehicles. Due to changes within NASA's Exploration Initiative, this task was cancelled early. This presentation provides an overview of the work that was accomplished prior to cancellation. The Snecma team chose an Apollo-type capsule as the reference vehicle for the work. They began with the design of a ceramic aft heatshield (CAS) utilizing C/SiC panels as the capsule heatshield, a C/SiC deployable decelerator and several ablators. They additionally developed a health monitoring system, high temperature structures testing, and the insulation characterization. Though the task was pre-maturely cancelled, a significant quantity of work was accomplished.

  3. Terahertz Computed Tomography of NASA Thermal Protection System Materials

    Science.gov (United States)

    Roth, D. J.; Reyes-Rodriguez, S.; Zimdars, D. A.; Rauser, R. W.; Ussery, W. W.

    2011-01-01

    A terahertz axial computed tomography system has been developed that uses time domain measurements in order to form cross-sectional image slices and three-dimensional volume renderings of terahertz-transparent materials. The system can inspect samples as large as 0.0283 cubic meters (1 cubic foot) with no safety concerns as for x-ray computed tomography. In this study, the system is evaluated for its ability to detect and characterize flat bottom holes, drilled holes, and embedded voids in foam materials utilized as thermal protection on the external fuel tanks for the Space Shuttle. X-ray micro-computed tomography was also performed on the samples to compare against the terahertz computed tomography results and better define embedded voids. Limits of detectability based on depth and size for the samples used in this study are loosely defined. Image sharpness and morphology characterization ability for terahertz computed tomography are qualitatively described.

  4. Flexible Thermal Protection System Development for Hypersonic Inflatable Aerodynamic Decelerators

    Science.gov (United States)

    DelCorso, Joseph A.; Bruce, Walter E., III; Hughes, Stephen J.; Dec, John A.; Rezin, Marc D.; Meador, Mary Ann B.; Guo, Haiquan; Fletcher, Douglas G.; Calomino, Anthony M.; Cheatwood, McNeil

    2012-01-01

    The Hypersonic Inflatable Aerodynamic Decelerators (HIAD) project has invested in development of multiple thermal protection system (TPS) candidates to be used in inflatable, high downmass, technology flight projects. Flexible TPS is one element of the HIAD project which is tasked with the research and development of the technology ranging from direct ground tests, modelling and simulation, characterization of TPS systems, manufacturing and handling, and standards and policy definition. The intent of flexible TPS is to enable large deployable aeroshell technologies, which increase the drag performance while significantly reducing the ballistic coefficient of high-mass entry vehicles. A HIAD requires a flexible TPS capable of surviving aerothermal loads, and durable enough to survive the rigors of construction, handling, high density packing, long duration exposure to extrinsic, in-situ environments, and deployment. This paper provides a comprehensive overview of key work being performed within the Flexible TPS element of the HIAD project. Included in this paper is an overview of, and results from, each Flexible TPS research and development activity, which includes ground testing, physics-based thermal modelling, age testing, margins policy, catalysis, materials characterization, and recent developments with new TPS materials.

  5. Enabling Technology for Thermal Protection on HIAD and Other Hypersonic Missions, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — S. D. Miller and Associates proposes to investigate a new class of thermal insulations that will enable thermal protection systems (TPS) on ceramic matrix composite...

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

    Science.gov (United States)

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

    2017-12-01

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

  7. Study of organic ablative thermal-protection coating for solid rocket motor

    Science.gov (United States)

    Hua, Zenggong

    1992-06-01

    A study is conducted to find a new interior thermal-protection material that possesses good thermal-protection performance and simple manufacturing possibilities. Quartz powder and Cr2O3 are investigated using epoxy resin as a binder and Al2O3 as the burning inhibitor. Results indicate that the developed thermal-protection coating is suitable as ablative insulation material for solid rocket motors.

  8. Flexible Transpiration Cooled Thermal Protection System for Inflatable Atmospheric Capture and Entry Systems, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Andrews Space, Inc. proposes an innovative transpiration cooled aerobrake TPS design that is thermally protective, structurally flexible, and lightweight. This...

  9. Flexible Transpiration Cooled Thermal Protection System for Inflatable Atmospheric Capture and Entry Systems, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Andrews Space, Inc. proposes an innovative transpiration cooled aerobrake TPS design that is thermally protective, structurally flexible, and lightweight. This...

  10. Advanced Thermal Protection Systems (ATPS), Aerospace Grade Carbon Bonded Carbon Fiber Material, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Carbon bonded carbon fiber (CBCF) insulating material is the basis for several highly successful NASA developed thermal protection systems (TPS). Included among...

  11. On the Thermal Protection Systems of Landers for Venus Exploration

    Science.gov (United States)

    Ekonomov, A. P.; Ksanfomality, L. V.

    2018-01-01

    The landers of the Soviet Venera series—from Venera-9 to Venera-14—designed at the Lavochkin Association are a man-made monument to spectacular achievements of Soviet space research. For more than 40 years, they have remained the uneclipsed Soviet results in space studies of the Solar System. Within the last almost half a century, the experiments carried out by the Venera-9 to Venera-14 probes for studying the surface of the planet have not been repeated by any space agency in the world, mainly due to quite substantial technical problems. Since that time, no Russian missions with landers have been sent to Venus either. On Venus, there is an anoxic carbon dioxide atmosphere, where the pressure is 9.2 MPa and the temperature is 735 K near the surface. A long-lived lander should experience these conditions for an appreciable length of time. What technical solutions could provide a longer operation time for a new probe investigating the surface of Venus, if its thermal scheme is constructed similar to that of the Venera series? Onboard new landers, there should be a sealed module, where the physical conditions required for operating scientific instruments are maintained for a long period. At the same time, new high-temperature electronic equipment that remains functional under the above-mentioned conditions have appeared. In this paper, we consider and discuss different variants of the system for a long-lived sealed lander, in particular, the absorption of the penetrating heat due to water evaporation and the thermal protection construction for the instruments with intermediate characteristics.

  12. Development of a test device to characterize thermal protective performance of fabrics against hot steam and thermal radiation

    International Nuclear Information System (INIS)

    Su, Yun; Li, Jun

    2016-01-01

    Steam burns severely threaten the life of firefighters in the course of their fire-ground activities. The aim of this paper was to characterize thermal protective performance of flame-retardant fabrics exposed to hot steam and low-level thermal radiation. An improved testing apparatus based on ASTM F2731-11 was developed in order to simulate the routine fire-ground conditions by controlling steam pressure, flow rate and temperature of steam box. The thermal protective performance of single-layer and multi-layer fabric system with/without an air gap was studied based on the calibrated tester. It was indicated that the new testing apparatus effectively evaluated thermal properties of fabric in hot steam and thermal radiation. Hot steam significantly exacerbated the skin burn injuries while the condensed water on the skin’s surface contributed to cool down the skin tissues during the cooling. Also, the absorbed thermal energy during the exposure and the cooling was mainly determined by the fabric’s configuration, the air gap size, the exposure time and the existence of hot steam. The research provides a effective method to characterize the thermal protection of fabric in complex conditions, which will help in optimization of thermal protection performance of clothing and reduction of steam burn. (paper)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  14. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  15. Mechanical Testing of Carbon Based Woven Thermal Protection Materials

    Science.gov (United States)

    Pham, John; Agrawal, Parul; Arnold, James O.; Peterson, Keith; Venkatapathy, Ethiraj

    2013-01-01

    Three Dimensional Woven thermal protection system (TPS) materials are one of the enabling technologies for mechanically deployable hypersonic decelerator systems. These materials have been shown capable of serving a dual purpose as TPS and as structural load bearing members during entry and descent operations. In order to ensure successful structural performance, it is important to characterize the mechanical properties of these materials prior to and post exposure to entry-like heating conditions. This research focuses on the changes in load bearing capacity of woven TPS materials after being subjected to arcjet simulations of entry heating. Preliminary testing of arcjet tested materials [1] has shown a mechanical degradation. However, their residual strength is significantly more than the requirements for a mission to Venus [2]. A systematic investigation at the macro and microstructural scales is reported here to explore the potential causes of this degradation. The effects of heating on the sizing (an epoxy resin coating used to reduce friction and wear during fiber handling) are discussed as one of the possible causes for the decrease in mechanical properties. This investigation also provides valuable guidelines for margin policies for future mechanically deployable entry systems.

  16. Preparation and thermal performance of paraffin/Nano-SiO2 nanocomposite for passive thermal protection of electronic devices

    International Nuclear Information System (INIS)

    Wang, Yaqin; Gao, Xuenong; Chen, Peng; Huang, Zhaowen; Xu, Tao; Fang, Yutang; Zhang, Zhengguo

    2016-01-01

    Highlights: • Three types of paraffin/nano-SiO 2 nanocomposites were prepared and characterized. • Thermo-physical properties of these composites were determined and compared. • One composite with lower thermal conductivity showed better thermal insulation properties. • This composite was identified as thermal insulation material for electronic components. - Abstract: In this paper, three grades of nano silicon dioxide (nano-SiO 2 ), NS1, NS2 and NS3, were mixed into paraffin to prepare nanocomposites as novel insulation materials for electronic passive thermal protection applications. The optimal mass percentages of paraffin for the three composites, NS1P, NS2P and NS3P, were determined to be 75%, 70% and 65%, respectively. Investigations by means of scanning electron micrographs (SEM), differential scanning calorimeter (DSC), thermogravimetric analysis (TG), hot disk analyzer and thermal protection performance tests were devoted to the morphology, thermal properties and thermal protection performance analysis of composites. Experimental results showed that paraffin uniformly distributed into the pores and on the surface of nano-SiO 2 . Melting points of composites declined and experimental latent heat became lower than the calculated values with the decrease of nano-SiO 2 pore size. The NS1P composite had larger thermal storage capacity, better reliability and stability compared with NS2P and NS3P. In addition, compared with 90% wt.% paraffin/EG composite, the incorporation of NS1 (25 wt.%) into paraffin caused not only 63.2% reduction in thermal conductivity, but also 21.8% increase in thermal protection time affected by the ambient temperature. Thus those good properties confirmed that NS1P (75 wt.%) composite was a viable candidate for protecting electronic devices under high temperature environment.

  17. Investigation for thermal stability of U3Si2 and protection methods

    International Nuclear Information System (INIS)

    Zhang Huiying; Sun Jichang; Sun Rongxian

    1994-08-01

    The thermal stability of U 3 Si 2 in Ar, N 2 and air, and the interaction between U 3 Si 2 and Al, Zr have been investigated by thermal analysis method. According to the results of thermal analysis, protection measures for various procedures have been improved. From the practice, it shows that the protection measures can ensure the safety of production and raise the product quality as well as reduce the cost effectively

  18. Blanket options for high-efficiency fusion power

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  19. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  20. Fusion blanket for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  1. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-12-01

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

  4. 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)

    Neves, S.F.; Couto, S.; Campos, J.B.L.M.; Mayor, T.S.

    2015-01-01

    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

  5. Application of waterproof breathable fabric in thermal protective clothing exposed to hot water and steam

    Science.gov (United States)

    Su, Y.; Li, R.; Song, G.; Li, J.

    2017-10-01

    A hot water and steam tester was used to examine thermal protective performance of waterproof and breathable fabric against hot water and steam hazards. Time to cause skin burn and thermal energy absorbed by skin during exposure and cooling phases was employed to characterize the effect of configuration, placing order and properties of waterproof and breathable fabric on the thermal protective performance. The difference of thermal protective performance due to hot water and steam hazards was discussed. The result showed that the configuration of waterproof and breathable fabric presented a significant effect on the thermal protective performance of single- and double-layer fabric system, while the difference between different configurations in steam hazard was greater than that in hot water hazard. The waterproof and breathable fabric as outer layer provided better protection than that as inner layer. Increasing thickness and moisture regain improved the thermal protective performance of fabric system. Additionally, the thermal energy absorbed by skin during the cooling phase was affected by configuration, thickness and moisture regain of fabric. The findings will provide technical data to improve performance of thermal protective clothing in hot water and steam hazards.

  6. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

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

  7. Limiter and first wall of the fusion reactor blanket

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  9. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Yican

    2000-01-01

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

  11. Calculations of tritium breeding ratio and inventory distributions of FEB blanket

    International Nuclear Information System (INIS)

    Deng Baiquan

    2001-01-01

    Based on the design features of FEB reactor blanket, the tritium breeding ratio and tritium concentrations in liquid lithium of each breeding zone have been calculated after 10 days full power operation for outboard blanket and one day operation for inboard blanket. The comparisons with the results calculated by Monte-Carlo code MORSE-CGT are made. Meanwhile the inventory in beryllium multiplier after one-year full power operation has also been estimated. An important conclusion has been drew the thermal hydraulic design should be careful to guarantee the blanket temperature should not rise as high as 680 degree C

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

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

    1984-01-01

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

  13. Nuclear Analysis of an ITER Blanket Module

    Science.gov (United States)

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

    2013-08-01

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

  14. NanoSatellite Thermal Overload Protection System (nSTOPS)

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to develop and demonstrate a laboratory version of a means to electrically dissipate excess thermal energy from 3-cube (and larger) nanosatellites:...

  15. An overview of recent projects to study thermal protection in life rafts, lifeboats and immersion suits

    Energy Technology Data Exchange (ETDEWEB)

    Mak, L.; DuCharme, M. B.; Farnworth, B.; Wissler, E. H.; Brown, R.; Kuczora, A. [Maritime and Arctic Survival Scientific and Engineering Ressearch Team (Canada)

    2011-07-01

    Survival during a marine evacuation in cold regions is very challenging. However international regulations do not require specific thermal protection or ventilation performance criteria for lifeboats. In the same way, the testing methods for approval testing of immersion suits are not standardised. This paper investigated recent projects completed or on-going to study thermal protection in life rafts, lifeboats and immersion suits. An overview of several projects from the Maritime and Arctic Survival Scientific and Engineering Research Team (MASSERT) was conducted. This review provided the necessary knowledge to advance international standards and develop the thermal protection requirements for survival in the Arctic. The results showed the MASSERT correlated thermal insulation values between human subjects and thermal manikins in life rafts and in immersion suits. It was found that the manikins are a valuable evaluation tool, as well as the computerised models used as prediction tools.

  16. Liquid lithium blanket processing studies

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  17. On-Orbit Health Monitoring and Repair Assessment of Thermal Protection Systems, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR project delivers On-orbit health MoNItoring and repair assessment of THERMal protection systems (OMNI_THERM). OMNI_THERM features impedance-based...

  18. Monitoring of Thermal Protection Systems using Robust Self-Organizing Optical Fiber Sensing Networks

    Data.gov (United States)

    National Aeronautics and Space Administration — Effective thermal protection systems are crucial for spacecraft or future hypersonic transports re-entering the atmosphere. Micro-meteoroids and orbital debris...

  19. In-Situ Real-Time Temperature Monitoring of Thermal Protection Systems, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This program addresses the need for interfacial and in-depth temperature monitoring of thermal protection systems (TPS). Novel, linear drive, eddy current methods...

  20. Development of High-Fidelity Material Response Modeling for Resin-Infused Woven Thermal Protection Systems

    Data.gov (United States)

    National Aeronautics and Space Administration — For future space exploration missions, it is essential for the thermal protection system (TPS) found on hypersonic vehicles or atmospheric entry probes to be...

  1. Enabling Technology for Thermal Protection on HIAD and Other Hypersonic Missions, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Gas conduction and radiation are the two important heat transfer mechanisms in highly porous reusable thermal protection systems used for planetary entry of space...

  2. Utilization of Self-Healing Materials in Thermal Protection System Applications

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed project is the Utilization of Self-Healing Materials for Thermal Protection System (TPS) Applications. Currently, the technology for repairing TPS from...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  4. Thermal assessment of sunlight impinging on OSIRIS-REx OCAMS PolyCam, OTES, and IMU-sunshade MLI blankets in flight

    Science.gov (United States)

    Choi, Michael K.

    2017-09-01

    The NASA Origins, Spectral Interpretation, Resource Identification, Security, Regolith Explorer (OSIRIS-REx) spacecraft was successfully launched into orbit on September 8, 2016. It is traveling to a near-Earth asteroid (101955) Bennu, study it in detail, and bring back a pristine sample to Earth for scientific analyses. At the Outbound Cruise nominal spacecraft attitude, with Sun on +X, sunlight impinges on the OSIRIS-REx camera suite (OCAMS) PolyCam sunshade multilayer insulation (MLI) with microporous black polytetrafluoroethylene (PTFE), a portion of the PolyCam optics support tube (MLI with germanium black Kapton (GBK)), a portion of the OSIRIS-REx Thermal Emission Spectrometer (OTES) sunshade (MLI with GBK), the Inertia Measurement Unit (IMU) sunshade (MLI with GBK), and the OSIRIS-REx Laser Altimeter (OLA) sunshade (MLI with GBK). Sunlight is reflected or scattered by the above MLIs to the other components on the forward (+Z) deck. It illuminates the forward deck. A detailed thermal assessment on the solar impingement has been performed for the Proximity Ops at the asteroid, Touch-and-Go sample acquisition, and Return Cruise mission phases.

  5. Heat Exchange in “Human body - Thermal protection - Environment” System

    Science.gov (United States)

    Khromova, I. V.

    2017-11-01

    This article is devoted to the issues of simulation and calculation of thermal processes in the system called “Human body - Thermal protection - Environment” under low temperature conditions. It considers internal heat sources and convective heat transfer between calculated elements. Overall this is important for the Heat Transfer Theory. The article introduces complex heat transfer calculation method and local thermophysical parameters calculation method in the system called «Human body - Thermal protection - Environment», considering passive and active thermal protections, thermophysical and geometric properties of calculated elements in a wide range of environmental parameters (water, air). It also includes research on the influence that thermal resistance of modern materials, used in special protective clothes development, has on heat transfer in the system “Human body - Thermal protection - Environment”. Analysis of the obtained results allows adding of the computer research data to experiments and optimizing of individual life-support system elements, which are intended to protect human body from exposure to external factors.

  6. About possible technologies of creation nanostructures blankets

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  7. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  8. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    1983-10-01

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

  9. Thermal protection from a finite period of heat exposure – Heat survival of flight data recorders

    International Nuclear Information System (INIS)

    Rana, Ruhul Amin; Li, Ri

    2015-01-01

    This work relates to developing thermal protection for a finite period of exposure to a high temperature environment. This type of transient heat transfer problem starts with a heating period, which is then followed by a cooling period once the high temperature environment disappears. The study is particularly relevant to the thermal protection of flight data recorders from high temperature flame. In this work, transient heat conduction through a three-concentric-layer configuration is numerically studied, which includes a metal housing, a thermal insulation, and a phase change material. The thermal performance is evaluated using the center temperature changing with time. It is found that the center temperature reaches a peak during cooling period rather than heating period. Time taken to reach the peak and the peak value depend on the sizes and properties of the layers. The properties include latent heat of fusion, melting temperature, heat capacities, and thermal conductivities. Parametric study is conducted to analyze and distinguish the influence of these parameters. The study provides general guidance for determining sizes and selecting materials for the thermal design of flight data recorders. Additionally, the study is also useful for other similar applications, for which thermal management and protection over a period of time is needed. In this paper, analysis starts with a baseline configuration composed of specific materials and sizes. Finite changes are applied to sizes, properties of the materials, and the results are compared to understand the roles of the varied parameters in affecting the thermal protection performance. - Highlights: • We study the thermal design of flight data recorders for heat survival. • Consecutive heating and cooling of 3-layer configuration is investigated. • Influences of sizes and material properties on thermal protection are explored

  10. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

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

  11. Multiscale Modeling of Carbon/Phenolic Composite Thermal Protection Materials: Atomistic to Effective Properties

    Science.gov (United States)

    Arnold, Steven M.; Murthy, Pappu L.; Bednarcyk, Brett A.; Lawson, John W.; Monk, Joshua D.; Bauschlicher, Charles W., Jr.

    2016-01-01

    Next generation ablative thermal protection systems are expected to consist of 3D woven composite architectures. It is well known that composites can be tailored to achieve desired mechanical and thermal properties in various directions and thus can be made fit-for-purpose if the proper combination of constituent materials and microstructures can be realized. In the present work, the first, multiscale, atomistically-informed, computational analysis of mechanical and thermal properties of a present day - Carbon/Phenolic composite Thermal Protection System (TPS) material is conducted. Model results are compared to measured in-plane and out-of-plane mechanical and thermal properties to validate the computational approach. Results indicate that given sufficient microstructural fidelity, along with lowerscale, constituent properties derived from molecular dynamics simulations, accurate composite level (effective) thermo-elastic properties can be obtained. This suggests that next generation TPS properties can be accurately estimated via atomistically informed multiscale analysis.

  12. ITER blanket module shield block design and analysis

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  13. Ceramic sphere-pac breeder design for fusion blankets

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1990-12-01

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

  17. High temperature blankets for the production of synthetic fuels

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. Transient thermal protection of film covering circular aperture by sublimation and weak decomposition

    Energy Technology Data Exchange (ETDEWEB)

    Havstad, Mark A.; Miles, Robin R.; Hsieh, Henry, E-mail: hsieh6@llnl.gov

    2015-03-15

    Highlights: • Precise sublimating layers can provide protection in transient thermal environments. • Sensitivity analysis shows that the uncertainty in properties has modest influence. • It is likely that methane layers are a good choice for IFE targets. - Abstract: Unwanted heating of sensitive surfaces in harsh thermal environments can be prevented by precise application of sacrificial materials such as sublimation layers and pyrolyzing films. The use of sublimation for the protection of circular polyimide membranes subjected to brief (∼100 ms) heating by infrared radiation and hot (6000 K) inert gas convection is analyzed. Selection of sublimation material and sublimation layer and membrane thickness is considered with emphasis on providing sufficient thermal protection yet negligible unwanted material remaining at the end of a specified heating period. Though the analysis here is general, the motivation is protection of the polyimide films covering the laser entrance holes on IFE (inertial fusion energy) hohlraums being injected into the hot gas (xenon) protecting IFE reactor chambers. Both one and two dimensional thermal models are used to develop a robust thermal concept. Sensitivity analyses (SA) methods are exercised to show where the design may be vulnerable and which input parameters have the greatest effect on performance and likelihood of success. For the design and conditions considered, methane sublimating layers are probably preferred over xenon or pentane.

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

    Science.gov (United States)

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

    2017-04-01

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

  20. The transpiration cooled first wall and blanket concept

    International Nuclear Information System (INIS)

    Barleon, Leopold; Wong, Clement

    2002-01-01

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

  1. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

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

  2. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

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

    1986-02-01

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

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

    Science.gov (United States)

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

    2014-11-01

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

  4. A coupled systems code-CFD MHD solver for fusion blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Wolfendale, Michael J., E-mail: m.wolfendale11@imperial.ac.uk; Bluck, Michael J.

    2015-10-15

    Highlights: • A coupled systems code-CFD MHD solver for fusion blanket applications is proposed. • Development of a thermal hydraulic systems code with MHD capabilities is detailed. • A code coupling methodology based on the use of TCP socket communications is detailed. • Validation cases are briefly discussed for the systems code and coupled solver. - Abstract: The network of flow channels in a fusion blanket can be modelled using a 1D thermal hydraulic systems code. For more complex components such as junctions and manifolds, the simplifications employed in such codes can become invalid, requiring more detailed analyses. For magnetic confinement reactor blanket designs using a conducting fluid as coolant/breeder, the difficulties in flow modelling are particularly severe due to MHD effects. Blanket analysis is an ideal candidate for the application of a code coupling methodology, with a thermal hydraulic systems code modelling portions of the blanket amenable to 1D analysis, and CFD providing detail where necessary. A systems code, MHD-SYS, has been developed and validated against existing analyses. The code shows good agreement in the prediction of MHD pressure loss and the temperature profile in the fluid and wall regions of the blanket breeding zone. MHD-SYS has been coupled to an MHD solver developed in OpenFOAM and the coupled solver validated for test geometries in preparation for modelling blanket systems.

  5. The blanket interface to TSTA

    International Nuclear Information System (INIS)

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

    1988-01-01

    The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

  6. PLZT thermal/flash protective goggles: device concepts and constraints

    International Nuclear Information System (INIS)

    Cutchen, J.T.

    1979-01-01

    In 1975 Sandia Laboratories began the design and development of PLZT Goggles for the US Air Force to provide protection from temporary flashblindness and permanent retinal burns caused by the brilliant flash of nuclear explosions. The user requirements, system and physical constraints, and use/storage environments were all considered in arriving at the final design goals. When the program began, there was no industrial capability to manufacture large-aperture PLZT materials or bonded lens assemblies. The technology has been established from a laboratory baseline in a brief period, and operational testing and evaluation by the Air Force has been completed. The goggles, identified as the EEU-2/P,, are now in production

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

    International Nuclear Information System (INIS)

    Werner, R.W.

    1978-01-01

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

  8. Design of thermal protection system for 8 foot HTST combustor

    Science.gov (United States)

    Moskowitz, S.

    1973-01-01

    The combustor in the 8-foot high temperature structures tunnel at the NASA-Langley Research Center has encountered cracking over a period of 50-250 tunnel tests within a limited range of the required operating envelope. A program was conducted which analyzed the failed combustor liner hardware and determined that the mechanism of failure was vibratory fatigue. A vibration damper system using wave springs located axially between the liner T-bar and the liner support was designed as an intermediate solution to extend the life of the current two-pass regenerative air-cooled liner. The effects of liner wall thickness, cooling air passage height, stiffener ring geometry, reflective coatings, and liner material selection were investigated for these designs. Preliminary layout design arrangements including the external water-cooling system requirements, weight estimates, installation requirements and preliminary estimates of manufacturing costs were prepared for the most promissing configurations. A state-of-the-art review of thermal barrier coatings and an evaluation of reflective coatings for the gasside surface of air-cooled liners are included.

  9. Effect of Air Gap Entrapped in Firefighter Protective Clothing on Thermal Resistance and Evaporative Resistance

    Directory of Open Access Journals (Sweden)

    He Hualing

    2018-03-01

    Full Text Available Heat and water vapor transfer behavior of thermal protective clothing is greatly influenced by the air gap entrapped in multilayer fabric system. In this study, a sweating hot plate method was used to investigate the effect of air gap position and size on thermal resistance and evaporative resistance of firefighter clothing under a range of ambient temperature and humidity. Results indicated that the presence of air gap in multilayer fabric system decreased heat and water vapor transfer abilities under normal wear. Moreover, the air gap position slightly influenced the thermal and evaporative performances of the firefighter clothing. In this study, the multilayer fabric system obtained the highest thermal resistance, when the air space was located at position B. Furthermore, the effect of ambient temperature on heat and water vapor transfer properties of the multilayer fabric system was also investigated in the presence of a specific air gap. It was indicated that ambient temperature did not influence the evaporative resistance of thermal protective clothing. A thermographic image was used to test the surface temperature of multilayer fabric system when an air gap was incorporated. These results suggested that a certain air gap entrapped in thermal protective clothing system could affect wear comfort.

  10. Evaluation of potential blanket concepts for a Demonstration Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Chapin, D.L.; Chi, J.W.H.; Kelly, J.L.

    1978-01-01

    An evaluation has been made of several different blanket concepts for use in a near-term Demonstration Tokamak Hybrid Reactor (DTHR), whose main objective would be to produce a significant amount of fissile fuel while demonstrating the feasibility of the tokamak hybrid reactor concept. The desirability of a simple design using proven technology plus a proliferation resistant fuel cycle led to the selection of a low temperature and pressure water-cooled, zircaloy clad ThO 2 blanket concept to breed 233 U. The nuclear performance and thermal-hydraulics characteristics of the blanket were evaluated to arrive at a consistent design. The blanket was found to be feasible for producing a significant amount of fissile fuel even with the limited operating conditions and blanket coverage in the DTHR

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

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    Lee, Jeong-Hun; Kim, Geon-Woo; Park, Goon-Cherl; Cho, Hyoung-Kyu; Im, Kihak

    2014-01-01

    In the present study, thermal-hydraulic analyses for the blanket concept are being conducted using the Multidimensional Analysis of Reactor Safety (MARSKS) code, which has been used for the safety analysis of a pressurized water reactor. The purposes of the analyses are to verify the applicability of the code for the proposed blanket system, to investigate the departure of nucleate boiling (DNB) occurrence during the normal and transient conditions, and to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. In this paper, the thermal analysis results of the proposed blanket design using the MARS-KS code are presented for the normal operation and an accident condition of a reduced coolant flow rate. Afterwards, the plan for the whole blanket system analysis using MARSKS is introduced and the result of the first trial for the multiple blanket module analysis is summarized. In the present study, thermal-hydraulic analyses for the blanket concept were conducted using the MARS-KS code for a single blanket module. By comparing the MARS calculation results with the CFD analysis results, it was found that MARS-KS can be applied for the blanket thermal analysis with less number of computational meshes. Moreover, due to its capability on the two-phase flow analysis, it can be used for the transient or accident simulation where a phase change may be resulted in. In the future, the MARS-KS code will be applied for the anticipated transient and design based accident analyses. The investigation of the DNB occurrence during the normal and transient conditions will be of special interest of the analysis using it. After that, a methodology to simulate the entire blanket system was proposed by using the DLL version of MARS-KS. A supervisor program, which controls the multiple DLL files, was developed for the common header modelling. The program explicitly determines the flow rates of each module which can equalize

  13. THERMAL PROTECTION AND THERMAL STABILIZATION OF FIBER-OPTICAL GYROSCOPE INCLUDED IN STRAPDOWN INERTIAL NAVIGATION SYSTEM

    Directory of Open Access Journals (Sweden)

    D. S. Gromov

    2014-03-01

    Full Text Available It is known, that temperature perturbations and thermal modes have significant influence on the accuracy of a fiber-optical gyroscope. Nowadays, thermal perturbations are among the main problems in the field of navigation accuracy. Review of existing methods for decrease of temperature influences on the accuracy of a strapdown inertial navigation system with fiberoptical gyros showed, that the usage of constructive and compensation methods only is insufficient and, therefore, thermostabilization is required. Reversible thermostabilization system is offered, its main executive elements are thermoelectric modules (Peltier’s modules, heat transfer from which is provided by heatsinks at work surfaces of modules. This variant of thermostabilization maintenance is considered; Peltier’s modules and temperature sensors for the system are chosen. Parameters of heatsinks for heat transfer intensification are calculated. Fans for necessary air circulation in the device are chosen and thickness of thermal isolation is calculated. Calculations of thermal modes of navigation system with thermostabilization are made in modern software Autodesk Simulation CFD. Comparison of results for present and previous researches and calculations shows essential decrease in gradients of temperature on gyro surfaces and better uniformity of temperature field in the whole device. Conclusions about efficiency of the given method usage in view of accuracy improvement of navigation system are made. Thermostabilization provision of a strapdown inertial navigation system with fiberoptical gyros is proved. Thermostabilization application in combination with compensational methods can reach a necessary accuracy of navigation system.

  14. Nonablative lightweight thermal protection system for Mars Aeroflyby Sample collection mission

    Science.gov (United States)

    Suzuki, Toshiyuki; Aoki, Takuya; Ogasawara, Toshio; Fujita, Kazuhisa

    2017-07-01

    In this study, the concept of a nonablative lightweight thermal protection system (NALT) were proposed for a Mars exploration mission currently under investigation in Japan. The NALT consists of a carbon/carbon (C/C) composite skin, insulator tiles, and a honeycomb sandwich panel. Basic thermal characteristics of the NALT were obtained by conducting heating tests in high-enthalpy facilities. Thermal conductivity values of the insulator tiles as well as the emissivity values of the C/C skin were measured to develop a numerical analysis code for predicting NALT's thermal performance in flight environments. Finally, a breadboard model of a 600-mm diameter NALT aeroshell was developed and qualified through vibration and thermal vacuum tests.

  15. Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

    International Nuclear Information System (INIS)

    Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

    1985-01-01

    Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

  16. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

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

  17. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  18. Development of blanket remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  19. The TFTR lithium blanket module program

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  20. Ablation, Thermal Response, and Chemistry Program for Analysis of Thermal Protection Systems

    Science.gov (United States)

    Milos, Frank S.; Chen, Yih-Kanq

    2010-01-01

    In previous work, the authors documented the Multicomponent Ablation Thermochemistry (MAT) and Fully Implicit Ablation and Thermal response (FIAT) programs. In this work, key features from MAT and FIAT were combined to create the new Fully Implicit Ablation, Thermal response, and Chemistry (FIATC) program. FIATC is fully compatible with FIAT (version 2.5) but has expanded capabilities to compute the multispecies surface chemistry and ablation rate as part of the surface energy balance. This new methodology eliminates B' tables, provides blown species fractions as a function of time, and enables calculations that would otherwise be impractical (e.g. 4+ dimensional tables) such as pyrolysis and ablation with kinetic rates or unequal diffusion coefficients. Equations and solution procedures are presented, then representative calculations of equilibrium and finite-rate ablation in flight and ground-test environments are discussed.

  1. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-06-01

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

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

    International Nuclear Information System (INIS)

    Norajitra, P.

    1995-06-01

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

  3. Thermal comfort sustained by cold protective clothing in Arctic open-pit mining-a thermal manikin and questionnaire study.

    Science.gov (United States)

    Jussila, Kirsi; Rissanen, Sirkka; Aminoff, Anna; Wahlström, Jens; Vaktskjold, Arild; Talykova, Ljudmila; Remes, Jouko; Mänttäri, Satu; Rintamäki, Hannu

    2017-12-07

    Workers in the Arctic open-pit mines are exposed to harsh weather conditions. Employers are required to provide protective clothing for workers. This can be the outer layer, but sometimes also inner or middle layers are provided. This study aimed to determine how Arctic open-pit miners protect themselves against cold and the sufficiency, and the selection criteria of the garments. Workers' cold experiences and the clothing in four Arctic open-pit mines in Finland, Sweden, Norway and Russia were evaluated by a questionnaire (n=1,323). Basic thermal insulation (I cl ) of the reported clothing was estimated (ISO 9920). The I cl of clothing from the mines were also measured by thermal manikin (standing/walking) in 0.3 and 4.0 m/s wind. The questionnaire showed that the I cl of the selected clothing was on average 1.2 and 1.5 clo in mild (-5 to +5°C) and dry cold (-20 to -10°C) conditions, respectively. The I cl of the clothing measured by thermal manikin was 1.9-2.3 clo. The results show that the Arctic open-pit miners' selected their clothing based on occupational (time outdoors), environmental (temperature, wind, moisture) and individual factors (cold sensitivity, general health). However, the selected clothing was not sufficient to prevent cooling completely at ambient temperatures below -10°C.

  4. THE INFLUENCE OF THERMAL EVOLUTION IN THE MAGNETIC PROTECTION OF TERRESTRIAL PLANETS

    Energy Technology Data Exchange (ETDEWEB)

    Zuluaga, Jorge I.; Bustamante, Sebastian; Cuartas, Pablo A. [Instituto de Fisica-FCEN, Universidad de Antioquia, Calle 67 No. 53-108, Medellin (Colombia); Hoyos, Jaime H., E-mail: jzuluaga@fisica.udea.edu.co, E-mail: sbustama@pegasus.udea.edu.co, E-mail: p.cuartas@fisica.udea.edu.co, E-mail: jhhoyos@udem.edu.co [Departamento de Ciencias Basicas, Universidad de Medellin, Carrera 87 No. 30-65, Medellin (Colombia)

    2013-06-10

    Magnetic protection of potentially habitable planets plays a central role in determining their actual habitability and/or the chances of detecting atmospheric biosignatures. Here we develop a thermal evolution model of potentially habitable Earth-like planets and super-Earths (SEs). Using up-to-date dynamo-scaling laws, we predict the properties of core dynamo magnetic fields and study the influence of thermal evolution on their properties. The level of magnetic protection of tidally locked and unlocked planets is estimated by combining simplified models of the planetary magnetosphere and a phenomenological description of the stellar wind. Thermal evolution introduces a strong dependence of magnetic protection on planetary mass and rotation rate. Tidally locked terrestrial planets with an Earth-like composition would have early dayside magnetopause distances between 1.5 and 4.0 R{sub p} , larger than previously estimated. Unlocked planets with periods of rotation {approx}1 day are protected by magnetospheres extending between 3 and 8 R{sub p} . Our results are robust in comparison with variations in planetary bulk composition and uncertainties in other critical model parameters. For illustration purposes, the thermal evolution and magnetic protection of the potentially habitable SEs GL 581d, GJ 667Cc, and HD 40307g were also studied. Assuming an Earth-like composition, we found that the dynamos of these planets are already extinct or close to being shut down. While GL 581d is the best protected, the protection of HD 40307g cannot be reliably estimated. GJ 667Cc, even under optimistic conditions, seems to be severely exposed to the stellar wind, and, under the conditions of our model, has probably suffered massive atmospheric losses.

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  6. Comparative Analysis of the Thermal Insulation of Traditional and Newly Designed Protective Clothing for Foundry Workers

    Directory of Open Access Journals (Sweden)

    Iwona Frydrych

    2016-09-01

    Full Text Available An objective of the undertaken research was checking the applicability of aluminized basalt fabrics for the production of clothing for foundry workers. The results of flammability, the resistance to contact, convective and radiation heat, as well as the resistance to big molten metal splashes confirmed the thesis of applicability of the packages with the use of aluminized basalt fabric content for the assumed purpose; therefore, such protective clothing was produced. Thermal comfort of foundry workers is very important and related to many factors, i.e., the structure of the protective clothing package, the number of layers, their thickness, the distance between the body and appropriate underwear. In the paper, a comparison of the results of thermal insulation measurement of two kinds of protective clothing is presented: the traditional one made of aluminized glass fabrics and the new one made of aluminized basalt fabrics. Measurements of clothing thermal insulation were conducted using a thermal manikin dressed in the protective clothing and three kinds of underwear products covering the upper and lower part of the manikin.

  7. Robotic system for the servicing of the orbiter thermal protection system

    Science.gov (United States)

    Graham, Todd; Bennett, Richard; Dowling, Kevin; Manouchehri, Davoud; Cooper, Eric; Cowan, Cregg

    1994-01-01

    This paper describes the design and development of a mobile robotic system to process orbiter thermal protection system (TPS) tiles. This work was justified by a TPS automation study which identified tile rewaterproofing and visual inspection as excellent applications for robotic automation.

  8. 77 FR 11598 - Thermal Overload Protection for Electric Motors on Motor-Operated Valves

    Science.gov (United States)

    2012-02-27

    ... application of thermal overload protection devices that are integral with the motor starter for electric... Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 301-251-7455; email... Development Branch, Division of Engineering, Office of Nuclear Regulatory Research. [FR Doc. 2012-4552 Filed 2...

  9. Surface Catalytic Efficiency of Advanced Carbon Carbon Candidate Thermal Protection Materials for SSTO Vehicles

    Science.gov (United States)

    Stewart, David A.

    1996-01-01

    The catalytic efficiency (atom recombination coefficients) for advanced ceramic thermal protection systems was calculated using arc-jet data. Coefficients for both oxygen and nitrogen atom recombination on the surfaces of these systems were obtained to temperatures of 1650 K. Optical and chemical stability of the candidate systems to the high energy hypersonic flow was also demonstrated during these tests.

  10. The Relationship between Physical Activity and Thermal Protective Clothing on Functional Balance in Firefighters

    Science.gov (United States)

    Kong, Pui W.; Suyama, Joe; Cham, Rakie; Hostler, David

    2012-01-01

    We investigated the relationship between baseline physical training and the use of firefighting thermal protective clothing (TPC) with breathing apparatus on functional balance. Twenty-three male firefighters performed a functional balance test under four gear/clothing conditions. Participants were divided into groups by physical training status,…

  11. GCD TechPort Data Sheets Thermal Protection System Materials (TPSM) Project

    Science.gov (United States)

    Chinnapongse, Ronald L.

    2014-01-01

    The Thermal Protection System Materials (TPSM) Project consists of three distinct project elements: the 3-Dimensional Multifunctional Ablative Thermal Protection System (3D MAT) project element; the Conformal Ablative Thermal Protection System (CA-TPS) project element; and the Heatshield for Extreme Entry Environment Technology (HEEET) project element. 3D MAT seeks to design, develop and deliver a game changing material solution based on 3-dimensional weaving and resin infusion approach for manufacturing a material that can function as a robust structure as well as a thermal protection system. CA-TPS seeks to develop and deliver a conformal ablative material designed to be efficient and capable of withstanding peak heat flux up to 500 W/ sq cm, peak pressure up to 0.4 atm, and shear up to 500 Pa. HEEET is developing a new ablative TPS that takes advantage of state-of-the-art 3D weaving technologies and traditional manufacturing processes to infuse woven preforms with a resin, machine them to shape, and assemble them as a tiled solution on the entry vehicle substructure or heatshield.

  12. Protection Heater Design Validation for the LARP Magnets Using Thermal Imaging

    CERN Document Server

    Marchevsky, M; Cheng, D W; Felice, H; Sabbi, G; Salmi, T; Stenvall, A; Chlachidze, G; Ambrosio, G; Ferracin, P; Izquierdo Bermudez, S; Perez, J C; Todesco, E

    2016-01-01

    Protection heaters are essential elements of a quench protection scheme for high-field accelerator magnets. Various heater designs fabricated by LARP and CERN have been already tested in the LARP high-field quadrupole HQ and presently being built into the coils of the high-field quadrupole MQXF. In order to compare the heat flow characteristics and thermal diffusion timescales of different heater designs, we powered heaters of two different geometries in ambient conditions and imaged the resulting thermal distributions using a high-sensitivity thermal video camera. We observed a peculiar spatial periodicity in the temperature distribution maps potentially linked to the structure of the underlying cable. Two-dimensional numerical simulation of heat diffusion and spatial heat distribution have been conducted, and the results of simulation and experiment have been compared. Imaging revealed hot spots due to a current concentration around high curvature points of heater strip of varying cross sections and visuali...

  13. A new approach to characterize the effect of fabric deformation on thermal protective performance

    International Nuclear Information System (INIS)

    Li, Jun; Li, Xiaohui; Lu, Yehu; Wang, Yunyi

    2012-01-01

    It is very important to evaluate thermal protective performance (TPP) in laboratory-simulated fire scenes as accurately as possible. For this paper, to thoroughly understand the effect of fabric deformation on basic physical properties and TPP of flame-retardant fabrics exposed to flash fire, a new modified TPP testing apparatus was developed. Different extensions were employed to simulate the various extensions displayed during different body motions. The tests were also carried out with different air gaps. The results showed a significant decrease in air permeability after deformation. However, the change of thickness was slight. The fabric deformation had a complicated effect on thermal protection with different air gaps. The change of TPP depended on the balance between the surface contact area and the thermal insulation. The newly developed testing apparatus could be well employed to evaluate the effect of deformation on TPP of flame-resistant fabrics. (paper)

  14. A new approach to characterize the effect of fabric deformation on thermal protective performance

    Science.gov (United States)

    Li, Jun; Li, Xiaohui; Lu, Yehu; Wang, Yunyi

    2012-04-01

    It is very important to evaluate thermal protective performance (TPP) in laboratory-simulated fire scenes as accurately as possible. For this paper, to thoroughly understand the effect of fabric deformation on basic physical properties and TPP of flame-retardant fabrics exposed to flash fire, a new modified TPP testing apparatus was developed. Different extensions were employed to simulate the various extensions displayed during different body motions. The tests were also carried out with different air gaps. The results showed a significant decrease in air permeability after deformation. However, the change of thickness was slight. The fabric deformation had a complicated effect on thermal protection with different air gaps. The change of TPP depended on the balance between the surface contact area and the thermal insulation. The newly developed testing apparatus could be well employed to evaluate the effect of deformation on TPP of flame-resistant fabrics.

  15. Lightweight Ablative and Ceramic Thermal Protection System Materials for NASA Exploration Systems Vehicles

    Science.gov (United States)

    Valentine, Peter G.; Lawrence, Timothy W.; Gubert, Michael K.; Milos, Frank S.; Kiser, James D.; Ohlhorst, Craig W.; Koenig, John R.

    2006-01-01

    As a collaborative effort among NASA Centers, the "Lightweight Nonmetallic Thermal Protection Materials Technology" Project was set up to assist mission/vehicle design trade studies, to support risk reduction in thermal protection system (TPS) material selections, to facilitate vehicle mass optimization, and to aid development of human-rated TPS qualification and certification plans. Missions performing aerocapture, aerobraking, or direct aeroentry rely on advanced heatshields that allow reductions in spacecraft mass by minimizing propellant requirements. Information will be presented on candidate materials for such reentry approaches and on screening tests conducted (material property and space environmental effects tests) to evaluate viable candidates. Seventeen materials, in three classes (ablatives, tiles, and ceramic matrix composites), were studied. In additional to physical, mechanical, and thermal property tests, high heat flux laser tests and simulated-reentry oxidation tests were performed. Space environmental effects testing, which included exposures to electrons, atomic oxygen, and hypervelocity impacts, was also conducted.

  16. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

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

  17. A new method of thermal protection by opposing jet for a hypersonic aeroheating strut

    Science.gov (United States)

    Qin, Jiang; Ning, Dongpo; Feng, Yu; Zhang, Junlong; Feng, Shuo; Bao, Wen

    2017-06-01

    This paper presents the numerical investigation of thermal protection of scramjet strut by opposing jet in supersonic stream of Mach number 6 with a hydrogen fueled scramjet strut model using CFD software. Simulation results indicate that when a small amount of fuel is injected from the nose of the strut, the bow shock is pushed away from the strut, and the heat flux is reduced in the strut, especially at the leading edge. Opposing jet forms a recirculation region near the nozzle so that the strut is covered with low temperature fuel and separated from free stream. An appropriate total pressure ratio can be used to reduce not only aerodynamic heating but also the drag of strut. It is therefore concluded that thermal protection of scramjet strut by opposing jet is one of the promising ways to protect scramjet strut in high enthalpy stream.

  18. Feasibility study of incore fission chamber application for neutron flux measurements on the NET blanket

    International Nuclear Information System (INIS)

    Bertalot, L.

    1987-01-01

    A feasibility study has been carried out on the use of in-core fission chambers as neutron diagnostic tools to perform neutron flux measurements on the blanket component of NET. The high neutron and gamma fluxes and the severe thermal-mechanical and magnetic conditions of the blanket structure have been taken into account in this analysis. Preliminary design criteria and specifications of an in-core detector are presented for NET application. A research and development programme is outlined which aims to obtain more information on the tecnological constraints arising from the severe conditions of the NET blanket

  19. Beryllium data base for in-pile mockup test on blanket of fusion reactor, (1)

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hiroshi; Ishitsuka, Etsuo (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Sakamoto, Naoki; Kato, Masakazu; Takatsu, Hideyuki.

    1992-11-01

    Beryllium has been used in the fusion blanket designs with ceramic breeder as a neutron multiplier to increase the net tritium breeding ratio (TBR). The properties of beryllium, that is physical properties, chemical properties, thermal properties, mechanical properties, nuclear properties, radiation effects, etc. are necessary for the fusion blanket design. However, the properties of beryllium have not been arranged for the fusion blanket design. Therefore, it is indispensable to check and examine the material data of beryllium reported previously. This paper is the first one of the series of papers on beryllium data base, which summarizes the reported material data of beryllium. (author).

  20. Evaluation of antioxidants stability by thermal analysis and its protective effect in heated edible vegetable oil

    Directory of Open Access Journals (Sweden)

    Seme Youssef Reda

    2011-06-01

    Full Text Available In this work, through the use of thermal analysis techniques, the thermal stabilities of some antioxidants were investigated, in order to evaluate their resistance to thermal oxidation in oils, by heating canola vegetable oil, and to suggest that antioxidants would be more appropriate to increase the resistance of vegetable oils in the thermal degradation process in frying. The techniques used were: Thermal Gravimetric (TG and Differential Scanning Calorimetry (DSC analyses, as well as an allusion to a possible protective action of the vegetable oils, based on the thermal oxidation of canola vegetable oil in the laboratory under constant heating at 180 ºC/8 hours for 10 days. The studied antioxidants were: ascorbic acid, sorbic acid, citric acid, sodium erythorbate, BHT (3,5-di-tert-butyl-4-hydroxytoluene, BHA (2, 3-tert-butyl-4-methoxyphenol, TBHQ (tertiary butyl hydroquinone, PG (propyl gallate - described as antioxidants by ANVISA and the FDA; and also the phytic acid antioxidant and the SAIB (sucrose acetate isobutyrate additive, which is used in the food industry, in order to test its behavior as an antioxidant in vegetable oil. The following antioxidants: citric acid, sodium erythorbate, BHA, BHT, TBHQ and sorbic acid decompose at temperatures below 180 ºC, and therefore, have little protective action in vegetable oils undergoing frying processes. The antioxidants below: phytic acid, ascorbic acid and PG, are the most resistant and begin their decomposition processes at temperatures between 180 and 200 ºC. The thermal analytical techniques have also shown that the SAIB antioxidant is the most resistant to oxidative action, and it can be a useful choice in the thermal decomposition prevention of edible oils, improving stability regarding oxidative processes.

  1. Effective evaluation of privacy protection techniques in visible and thermal imagery

    Science.gov (United States)

    Nawaz, Tahir; Berg, Amanda; Ferryman, James; Ahlberg, Jörgen; Felsberg, Michael

    2017-09-01

    Privacy protection may be defined as replacing the original content in an image region with a (less intrusive) content having modified target appearance information to make it less recognizable by applying a privacy protection technique. Indeed, the development of privacy protection techniques also needs to be complemented with an established objective evaluation method to facilitate their assessment and comparison. Generally, existing evaluation methods rely on the use of subjective judgments or assume a specific target type in image data and use target detection and recognition accuracies to assess privacy protection. An annotation-free evaluation method that is neither subjective nor assumes a specific target type is proposed. It assesses two key aspects of privacy protection: "protection" and "utility." Protection is quantified as an appearance similarity, and utility is measured as a structural similarity between original and privacy-protected image regions. We performed an extensive experimentation using six challenging datasets (having 12 video sequences), including a new dataset (having six sequences) that contains visible and thermal imagery. The new dataset is made available online for the community. We demonstrate effectiveness of the proposed method by evaluating six image-based privacy protection techniques and also show comparisons of the proposed method over existing methods.

  2. Structural analysis under the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Majumdar, S.

    1985-01-01

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

  3. Disruption problematics in segmented blanket concepts

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  4. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  5. An Approximate Ablative Thermal Protection System Sizing Tool for Entry System Design

    Science.gov (United States)

    Dec, John A.; Braun, Robert D.

    2006-01-01

    A computer tool to perform entry vehicle ablative thermal protection systems sizing has been developed. Two options for calculating the thermal response are incorporated into the tool. One, an industry-standard, high-fidelity ablation and thermal response program was integrated into the tool, making use of simulated trajectory data to calculate its boundary conditions at the ablating surface. Second, an approximate method that uses heat of ablation data to estimate heat shield recession during entry has been coupled to a one-dimensional finite-difference calculation that calculates the in-depth thermal response. The in-depth solution accounts for material decomposition, but does not account for pyrolysis gas energy absorption through the material. Engineering correlations are used to estimate stagnation point convective and radiative heating as a function of time. The sizing tool calculates recovery enthalpy, wall enthalpy, surface pressure, and heat transfer coefficient. Verification of this tool is performed by comparison to past thermal protection system sizings for the Mars Pathfinder and Stardust entry systems and calculations are performed for an Apollo capsule entering the atmosphere at lunar and Mars return speeds.

  6. Effect of blanket assembly shuffling on LMR neutronic performance

    International Nuclear Information System (INIS)

    Khalil, H.; Fujita, E.K.

    1987-01-01

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

  7. Laying Stress on Energy-Saving and Environmental Protection of Thermal Generation

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    @@ The most attraetive spot of the 11th Five-Year Plan is to change China's present mode of cconomic growth and take a road of circulative cconomy based on effective utilization of resources and environmental protection. Electric power as a basic industry,energy conservation and environmental protection will become one of its working cmphases in a period of time to come. In this connection, the journalist (Zhao Ran) from China Electric Power has exclusively interviewed Tang Yunlin, the former president of the China Electric Power Planning and Engineering Institute. He thought that the most important thing for power industry to save energy and protect environment is to bring about the energy conservation and environmental protection in thermal power plants rather than first devclop hydropower, nuclear power and renewable energy. His viewpoints and suggestions have been recognized by many insiders.

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

    International Nuclear Information System (INIS)

    1979-11-01

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

  9. Thermal buffering performance of composite phase change materials applied in low-temperature protective garments

    Science.gov (United States)

    Yang, Kai; Jiao, Mingli; Yu, Yuanyuan; Zhu, Xueying; Liu, Rangtong; Cao, Jian

    2017-07-01

    Phase change material (PCM) is increasingly being applied in the manufacturing of functional thermo-regulated textiles and garments. This paper investigated the thermal buffering performance of different composite PCMs which are suitable for the application in functional low-temperature protective garments. First, according to the criteria selecting PCM for functional textiles/garments, three kinds of pure PCM were selected as samples, which were n-hexadecane, n-octadecane and n-eicosane. To get the adjustable phase change temperature range and higher phase change enthalpy, three kinds of composite PCM were prepared using the above pure PCM. To evaluate the thermal buffering performance of different composite PCM samples, the simulated low-temperature experiments were performed in the climate chamber, and the skin temperature variation curves in three different low temperature conditions were obtained. Finally composite PCM samples’ thermal buffering time, thermal buffering capacity and thermal buffering efficiency were calculated. Results show that the comprehensive thermal buffering performance of n-octadecane and n-eicosane composite PCM is the best.

  10. Ballistic Performance of Porous Ceramic Thermal Protection Systems at 9 km/s

    Science.gov (United States)

    Miller, Joshua E.; Bohl, W. E.; Foreman, C. D.; Christiansen, Eric L.; Davis, B. A.

    2009-01-01

    Porous-ceramic, thermal-protection-systems are used heavily in current reentry vehicles like the Orbiter, and they are currently being proposed for the next generation of manned spacecraft, Orion. These materials insulate the structural components and sensitive electronic components of a spacecraft against the intense thermal environments of atmospheric reentry. Furthermore, these materials are also highly exposed to space environmental hazards like meteoroid and orbital debris impacts. This paper discusses recent impact testing up to 9 km/s on ceramic tiles similar to those used on the Orbiter. These tiles have a porous-batting of nominally 8 lb/cubic ft alumina-fiber-enhanced-thermal-barrier (AETB8) insulating material coated with a damage-resistant, toughened-unipiece-fibrous-insulation (TUFI) layer.

  11. Ballistic Performance of Porous-Ceramic, Thermal Protection Systems to 9 km/s

    Science.gov (United States)

    Miller, Joshua E.; Bohl, William E.; Foreman, Cory D.; Christiansen, Eric C.; Davis, Bruce A.

    2010-01-01

    Porous-ceramic, thermal protection systems are used heavily in current reentry vehicles like the Orbiter, and they are currently being proposed for the next generation of US manned spacecraft, Orion. These materials insulate the structural components and sensitive components of a spacecraft against the intense thermal environments of atmospheric reentry. These materials are also highly exposed to solid particle space environment hazards. This paper discusses recent impact testing up to 9.65 km/s on ceramic tiles similar to those used on the Orbiter. These tiles are a porous-ceramic insulator of nominally 8 lb/ft(exp 3) alumina-fiber-enhanced-thermal-barrier (AETB8) coated with a damage-resistant, toughened-unipiece-fibrous-insulation/reaction-cured-glass layer (TUFI/RCG).

  12. The monolithic carbon aerogels and aerogel composites for electronics and thermal protection applications

    Science.gov (United States)

    Lu, Sheng; Guo, Hui; Zhou, Yugui; Liu, Yuanyuan; Jin, Zhaoguo; Liu, Bin; Zhao, Yingmin

    2017-09-01

    Monolithic carbon aerogels have been prepared by condensation polymerization and high temperature pyrolysis. The morphology of carbon aerogels are characterized by SEM. The pore structure is characterized by N2 adsorption-desorption technique. Monolithic carbon aerogels are mesoporous nanomaterials. Carbon fiber reinforced carbon aerogel composites are prepared by in-situ sol-gel process. Fiber reinforced carbon aerogel composites are of high mechanical strength. The thermal response of the fiber reinforced aerogel composite samples are tested in an arc plasma wind tunnel. Carbon aerogel composites show good thermal insulation capability and high temperature resistance in inert atmosphere even at ultrahigh temperature up to 1800 °C. The results show that they are suitable for applications in electrodes for supercapacitors/ Lithium-ion batteries and aerospace thermal protection area.

  13. The Development of HfO2-Rare Earth Based Oxide Materials and Barrier Coatings for Thermal Protection Systems

    Science.gov (United States)

    Zhu, Dongming; Harder, Bryan James

    2014-01-01

    Advanced hafnia-rare earth oxides, rare earth aluminates and silicates have been developed for thermal environmental barrier systems for aerospace propulsion engine and thermal protection applications. The high temperature stability, low thermal conductivity, excellent oxidation resistance and mechanical properties of these oxide material systems make them attractive and potentially viable for thermal protection systems. This paper will focus on the development of the high performance and high temperature capable ZrO2HfO2-rare earth based alloy and compound oxide materials, processed as protective coating systems using state-or-the-art processing techniques. The emphasis has been in particular placed on assessing their temperature capability, stability and suitability for advanced space vehicle entry thermal protection systems. Fundamental thermophysical and thermomechanical properties of the material systems have been investigated at high temperatures. Laser high-heat-flux testing has also been developed to validate the material systems, and demonstrating durability under space entry high heat flux conditions.

  14. Power cables thermal protection by interval simulation of imprecise dynamical systems

    Energy Technology Data Exchange (ETDEWEB)

    Bontempi, G. [Universite Libre de Brussels (Belgium). Dept. d' Informatique; Vaccaro, A.; Villacci, D. [Universita del Sannio Benevento (Italy). Dept. of Engineering

    2004-11-01

    The embedding of advanced simulation techniques in power cables enables improved thermal protection because of higher accuracy, adaptiveness and. flexibility. In particular, they make possible (i) the accurate solution of differential equations describing the cables thermal dynamics and (ii) the adoption of the resulting solution in the accomplishment of dedicated protective functions. However, the use of model-based protective systems is exposed to the uncertainty affecting some model components (e.g. weather along the line route, thermophysical properties of the soil, cable parameters). When uncertainty can be described in terms of probability distribution, well-known techniques, such as Monte Carlo, are used to simulate the system behaviour. On the other hand, when the description of uncertainty in probabilistic terms is unfeasible or problematic, nonprobabilistic alternatives should be taken into consideration. This paper will discuss and compare three interval-based techniques as alternatives to probabilistic methods in the simulation of power cable dynamics. The experimental session will assess the interval-based approaches by simulating the thermal behaviour of medium voltage power cables.(author)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  16. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

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

  17. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

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

    2000-07-01

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

  18. A modernized high-pressure heater protection system for nuclear and thermal power stations

    Science.gov (United States)

    Svyatkin, F. A.; Trifonov, N. N.; Ukhanova, M. G.; Tren'kin, V. B.; Koltunov, V. A.; Borovkov, A. I.; Klyavin, O. I.

    2013-09-01

    Experience gained from operation of high-pressure heaters and their protection systems serving to exclude ingress of water into the turbine is analyzed. A formula for determining the time for which the high-pressure heater shell steam space is filled when a rupture of tubes in it occurs is analyzed, and conclusions regarding the high-pressure heater design most advisable from this point of view are drawn. A typical structure of protection from increase of water level in the shell of high-pressure heaters used in domestically produced turbines for thermal and nuclear power stations is described, and examples illustrating this structure are given. Shortcomings of components used in the existing protection systems that may lead to an accident at the power station are considered. A modernized protection system intended to exclude the above-mentioned shortcomings was developed at the NPO Central Boiler-Turbine Institute and ZioMAR Engineering Company, and the design solutions used in this system are described. A mathematical model of the protection system's main elements (the admission and check valves) has been developed with participation of specialists from the St. Petersburg Polytechnic University, and a numerical investigation of these elements is carried out. The design version of surge tanks developed by specialists of the Central Boiler-Turbine Institute for excluding false operation of the high-pressure heater protection system is proposed.

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

    International Nuclear Information System (INIS)

    Tanigawa, Hisashi; Enoeda, Mikio

    2010-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-15

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

  1. Self-cooled blanket concepts using Pb-17Li as liquid breeder and coolant

    International Nuclear Information System (INIS)

    Malang, S.; Deckers, H.; Fischer, U.; John, H.; Meyder, R.; Norajitra, P.; Reimann, J.; Reiser, H.; Rust, K.

    1991-01-01

    A blanket design concept using Pb-17Li eutectic alloy as both breeder material and coolant is described. Such a self-cooled blanket for the boundary conditions of a DEMO-reactor is under development at the Kernforschungszentrum Karlsruhe (KfK) in the frame of the European blanket development program. Results of investigations in the areas of design, neutronics, magneto-hydrodynamics, thermo-mechanics, ancillary loop systems, and safety are reported. Based on recent progress, it can be concluded that the boundary conditions of a DEMO-reactor can be met, tritium self-sufficiency can be obtained without using beryllium as an additional neutron multiplier, and tritium inventory and permeation are acceptably low. However, to complete judge the feasibility of the proposed concept, further studies are necessary to obtain a better understanding of the magneto-hydrodynamic phenomena and their effects on the thermal-hydraulic performance of a fusion reactor blanket. (orig.)

  2. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  3. NOEL: a no-leak fusion blanket concept

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  5. Methods of evaluating protective clothing relative to heat and cold stress: thermal manikin, biomedical modeling, and human testing.

    Science.gov (United States)

    O'Brien, Catherine; Blanchard, Laurie A; Cadarette, Bruce S; Endrusick, Thomas L; Xu, Xiaojiang; Berglund, Larry G; Sawka, Michael N; Hoyt, Reed W

    2011-10-01

    Personal protective equipment (PPE) refers to clothing and equipment designed to protect individuals from chemical, biological, radiological, nuclear, and explosive hazards. The materials used to provide this protection may exacerbate thermal strain by limiting heat and water vapor transfer. Any new PPE must therefore be evaluated to ensure that it poses no greater thermal strain than the current standard for the same level of hazard protection. This review describes how such evaluations are typically conducted. Comprehensive evaluation of PPE begins with a biophysical assessment of materials using a guarded hot plate to determine the thermal characteristics (thermal resistance and water vapor permeability). These characteristics are then evaluated on a thermal manikin wearing the PPE, since thermal properties may change once the materials have been constructed into a garment. These data may be used in biomedical models to predict thermal strain under a variety of environmental and work conditions. When the biophysical data indicate that the evaporative resistance (ratio of permeability to insulation) is significantly better than the current standard, the PPE is evaluated through human testing in controlled laboratory conditions appropriate for the conditions under which the PPE would be used if fielded. Data from each phase of PPE evaluation are used in predictive models to determine user guidelines, such as maximal work time, work/rest cycles, and fluid intake requirements. By considering thermal stress early in the development process, health hazards related to temperature extremes can be mitigated while maintaining or improving the effectiveness of the PPE for protection from external hazards.

  6. Replacement of Ablators with Phase-Change Material for Thermal Protection of STS Elements

    Science.gov (United States)

    Kaul, Raj K.; Stuckey, Irvin; Munafo, Paul M. (Technical Monitor)

    2002-01-01

    As part of the research and development program to develop new Thermal Protection System (TPS) materials for aerospace applications at NASA's Marshall Space Flight Center (MSFC), an experimental study was conducted on a new concept for a non-ablative TPS material. Potential loss of TPS material and ablation by-products from the External Tank (ET) or Solid Rocket Booster (SRB) during Shuttle flight with the related Orbiter tile damage necessitates development of a non-ablative thermal protection system. The new Thermal Management Coating (TMC) consists of phase-change material encapsulated in micro spheres and a two-part resin system to adhere the coating to the structure material. The TMC uses a phase-change material to dissipate the heat produced during supersonic flight rather than an ablative material. This new material absorbs energy as it goes through a phase change during the heating portion of the flight profile and then the energy is slowly released as the phase-change material cools and returns to its solid state inside the micro spheres. The coating was subjected to different test conditions simulating design flight environments at the NASA/MSFC Improved Hot Gas Facility (IHGF) to study its performance.

  7. Saturation of SERCA's lipid annulus may protect against its thermal inactivation

    International Nuclear Information System (INIS)

    Fajardo, Val Andrew; Trojanowski, Natalie; Castelli, Laura M.; Miotto, Paula M.; Amoye, Foyinsola; Ward, Wendy E.; Tupling, A. Russell; LeBlanc, Paul J.

    2017-01-01

    The sarco(endo)plasmic reticulum Ca 2+ -ATPase (SERCA) pumps are integral membrane proteins that catalyze the active transport of Ca 2+ into the sarcoplasmic reticulum, thereby eliciting muscle relaxation. SERCA pumps are highly susceptible to oxidative damage, and cytoprotection of SERCA dampens thermal inactivation and is a viable therapeutic strategy in combating diseases where SERCA activity is impaired, such as muscular dystrophy. Here, we sought to determine whether increasing the percent of saturated fatty acids (SFA) within SERCA's lipid annulus through diet could protect SERCA pumps from thermal inactivation. Female Wistar rats were fed either a semi-purified control diet (AIN93G, 7% soybean oil by weight) or a modified AIN93G diet containing high SFA (20% lard by weight) for 17 weeks. Soleus muscles were extracted and SERCA lipid annulus and activity under thermal stress were analyzed. Our results show that SERCA's lipid annulus is abundant with short-chain (12–14 carbon) fatty acids, which corresponds well with SERCA's predicted bilayer thickness of 21 Å. Under control-fed conditions, SERCA's lipid annulus was already highly saturated (79%), and high-fat feeding did not increase this any further. High-fat feeding did not mitigate the reductions in SERCA activity seen with thermal stress; however, correlational analyses revealed significant and strong associations between % SFA and thermal stability of SERCA activity with greater %SFA being associated with lower thermal inactivation and greater % polyunsaturation and unsaturation index being associated with increased thermal inactivation. Altogether, these findings show that SERCA's lipid annulus may influence its susceptibility to oxidative damage, which could have implications in muscular dystrophy and age-related muscle wasting. - Highlights: • SERCA's lipid annulus in rat soleus was measured after immunoconcentration. • Short fatty acid chains surround SERCA and

  8. Characterization of thermally sprayed coatings for high-temperature wear-protection applications

    International Nuclear Information System (INIS)

    Li, C.C.

    1980-03-01

    Under normal high-temperature gas-cooled reactor (HTGR) operating conditions, faying surfaces of metallic components under high contact pressure are prone to friction, wear, and self-welding damage. Component design calls for coatings for the protection of the mating surfaces. Anticipated operating temperatures up to 850 to 950 0 C (1562 to 1742 0 F) and a 40-y design life require coatings with excellent thermal stability and adequate wear and spallation resistance, and they must be compatible with the HTGR coolant helium environment. Plasma and detonation-gun (D-gun) deposited chromium carbide-base and stabilized zirconia coatings are under consideration for wear protection of reactor components such as the thermal barrier, heat exchangers, control rods, and turbomachinery. Programs are under way to address the structural integrity, helium compatibility, and tribological behavior of relevant sprayed coatings. In this paper, the need for protection of critical metallic components and the criteria for selection of coatings are discussed. The technical background to coating development and the experience with the steam cycle HTGR (HTGR-SC) are commented upon. Coating characterization techniques employed at General Atomic Company (GA) are presented, and the progress of the experimental programs is briefly reviewed. In characterizing the coatings for HTGR applications, it is concluded that a systems approach to establish correlation between coating process parameters and coating microstructural and tribological properties for design consideration is required

  9. Thermal protection system for the concrete core support floor at Fort St. Vrain

    International Nuclear Information System (INIS)

    Jones, H.; Hedgecock, P.D.

    1976-01-01

    A unique feature of the Fort St. Vrain HTGR is its steel jacketed concrete core support floor. The construction of this floor generally resembles that of the prestressed concrete reactor vessel, but its location immediately below the core hot gas outlets generates some particularly severe thermal protection requirements. A thermal barrier is used over the entire outer surface of the floor and in the twelve hot gas ducts which convey the primary coolant through the floor to the steam generators. A cooling water system of square tubes welded to the inside of the steel jacket is used to remove that heat which does pass through the thermal barrier and to maintain the concrete at acceptable temperatures. The design approach to the floor itself and to the thermal barriers and cooling system will be described, but the main emphasis of the paper will be on the total experience gained during construction and pre-operational testing. A particular problem experienced during construction was leakage from some cooling tubes, after their embedment in concrete. The solution to that problem was to develop a method for injecting catalyzed epoxy into the leaking tube. This method, which has general usefulness for in-service repairs, will be described. (author)

  10. An assessment of thermal spray coating technologies for high temperature corrosion protection

    International Nuclear Information System (INIS)

    Heath, G.R.; Heimgartner, P.; Gustafsson, S.; Irons, G.; Miller, R.

    1997-01-01

    The use of thermally sprayed coatings in combating high temperature corrosion continues to grow in the major industries of chemical, waste incineration, power generation and pulp and paper. This has been driven partially by the development of corrosion resistant alloys, improved knowledge and quality in the thermal spray industry and continued innovation in thermal spray equipment. There exists today an extensive range of thermal spray process options, often with the same alloy solution. In demanding corrosion applications it is not sufficient to just specify alloy and coating method. For the production of reliable coatings the whole coating production envelope needs to be considered, including alloy selection, spray parameters, surface preparation, base metal properties, heat input etc. Combustion, arc-wire, plasma, HVOF and spray+fuse techniques are reviewed and compared in terms of their strengths and limitations to provide cost-effective solutions for high temperature corrosion protection. Arc wire spraying, HP/HVOF and spray+fuse are emerging as the most promising techniques to optimise both coating properties and economic/practical aspects. (orig.)

  11. Monitoring of Thermal Protection Systems Using Robust Self-Organizing Optical Fiber Sensing Networks

    Science.gov (United States)

    Richards, Lance

    2013-01-01

    The general aim of this work is to develop and demonstrate a prototype structural health monitoring system for thermal protection systems that incorporates piezoelectric acoustic emission (AE) sensors to detect the occurrence and location of damaging impacts, and an optical fiber Bragg grating (FBG) sensor network to evaluate the effect of detected damage on the thermal conductivity of the TPS material. Following detection of an impact, the TPS would be exposed to a heat source, possibly the sun, and the temperature distribution on the inner surface in the vicinity of the impact measured by the FBG network. A similar procedure could also be carried out as a screening test immediately prior to re-entry. The implications of any detected anomalies in the measured temperature distribution will be evaluated for their significance in relation to the performance of the TPS during re-entry. Such a robust TPS health monitoring system would ensure overall crew safety throughout the mission, especially during reentry

  12. Moisture absorption characteristics of the Orbiter thermal protection system and methods used to prevent water ingestion

    Science.gov (United States)

    Schomburg, C.; Dotts, R. L.; Tillian, D. J.

    1983-01-01

    The Space Shuttle Orbiter's silica tile Thermal Protection System (TPS) is beset by the moisture absorption problems inherently associated with low density, highly porous insulation systems. Attention is presently given to the comparative success of methods for the minimization and/or prevention of water ingestion by the TPS tiles, covering the development of water-repellent agents and their tile application techniques, flight test program results, and materials improvements. The use of external films for rewaterproofing of the TPS tiles after each mission have demonstrated marginal to unacceptable performance. By contrast, a tile interior waterproofing agent has shown promise.

  13. About properties of ZrO2 thermal protective coatings obtained from spherical powder mixtures

    Science.gov (United States)

    Berdnik, O. B.; Tsareva, I. N.; Tarasenko, Yu P.

    2017-05-01

    It is developed the technology of high-energy plasma spraying of the zirconium dioxide (ZrO2) thermal protective coating on the basis of ZrO2 tetragonal and cubic phases with the spheroidal grain shape and the columnar substructure, with the total porosity P = 4 %, the hardness HV = 12 GPa, the roughness parameter R a ˜ 6 μm, the thickness 0.3-3 mm. As a sublayer it is used the heat-resistant coating of “Ni-Co-Cr-Al-Y” system with an intermetallic phase composition and the layered microstructure of the grains.

  14. Development of a helmet-mounted PLZT thermal/flash protection system

    International Nuclear Information System (INIS)

    Harris, J.O. Jr.; Cutchen, J.T.; Pfoff, B.J.

    1976-01-01

    Sandia Laboratories is developing PLZT thermal/flash protective devices (TFPD's) goggles to prevent exposure and resultant eye damage from nuclear weapon detonations. The primary emphasis of the present program is to transfer technology and establish production capability for helmet-mounted PLZT/TFPD goggles for USAF flight crews, with a non-helmet-mounted configuration to follow. The first production units are anticipated in the fall of 1977. The operating principles of the PLZT/TFPD goggle device are briefly outlined, and the device configuration and operational characteristics are described

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

    Science.gov (United States)

    Enoeda, M.; Kosaku, Y.; Hatano, T.; Kuroda, T.; Miki, N.; Honma, T.; Akiba, M.; Konishi, S.; Nakamura, H.; Kawamura, Y.; Sato, S.; Furuya, K.; Asaoka, Y.; Okano, K.

    2003-12-01

    This paper presents results of conceptual design activities and associated R&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&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

  16. (D,T) Driven thorium hybrid blankets

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-01

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

  18. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

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

    1982-06-01

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

  19. Conformal Ablative Thermal Protection System for Planetary and Human Exploration Missions:An Overview of the Technology Maturation Effort

    Science.gov (United States)

    Beck, Robin A S.; Arnold, James O.; Gasch, Matthew J.; Stackpoole, Margaret M.; Prabhu, Dinesh K.; Szalai, Christine E.; Wercinski, Paul F.; Venkatapathy, Ethiraj

    2013-01-01

    The Office of Chief Technologist, NASA identified the need for research and technology development in part from NASAs Strategic Goal 3.3 of the NASA Strategic Plan to develop and demonstrate the critical technologies that will make NASAs exploration, science, and discovery missions more affordable and more capable. Furthermore, the Game Changing Development Program is a primary avenue to achieve the Agencys 2011 strategic goal to Create the innovative new space technologies for our exploration, science, and economic future. The National Research Council (NRC) Space Technology Roadmaps and Priorities report highlights six challenges and they are: Mass to Surface, Surface Access, Precision Landing, Surface Hazard Detection and Avoidance, Safety and Mission Assurance, and Affordability. In order for NASA to meet these challenges, the report recommends immediate focus on Rigid and Flexible Thermal Protection Systems. Rigid TPS systems such as Avcoat or SLA are honeycomb based and PICA is in the form of tiles. The honeycomb systems are manufactured using techniques that require filling of each (38 cell) by hand, and in a limited amount of time all of the cells must be filled and the heatshield must be cured. The tile systems such as PICA pose a different challenge as the low strain-to-failure and manufacturing size limitations require large number of small tiles with gap-fillers between the tiles. Recent investments in flexible ablative systems have given rise to the potential for conformal ablative TPS. A conformal TPS over a rigid aeroshell has the potential to solve a number of challenges faced by traditional rigid TPS materials. The high strain-to-failure nature of the conformal ablative materials will allow integration of the TPS with the underlying aeroshell structure much easier and enable monolithic-like configuration and larger segments (or parts) to be used. By reducing the overall part count, the cost of installation (based on cost comparisons between blanket

  20. Measuring the spectral emissivity of thermal protection materials during atmospheric reentry simulation

    Science.gov (United States)

    Marble, Elizabeth

    1996-01-01

    Hypersonic spacecraft reentering the earth's atmosphere encounter extreme heat due to atmospheric friction. Thermal Protection System (TPS) materials shield the craft from this searing heat, which can reach temperatures of 2900 F. Various thermophysical and optical properties of TPS materials are tested at the Johnson Space Center Atmospheric Reentry Materials and Structures Evaluation Facility, which has the capability to simulate critical environmental conditions associated with entry into the earth's atmosphere. Emissivity is an optical property that determines how well a material will reradiate incident heat back into the atmosphere upon reentry, thus protecting the spacecraft from the intense frictional heat. This report describes a method of measuring TPS emissivities using the SR5000 Scanning Spectroradiometer, and includes system characteristics, sample data, and operational procedures developed for arc-jet applications.

  1. Thermally Sprayed Aluminum Coatings for the Protection of Subsea Risers and Pipelines Carrying Hot Fluids

    Directory of Open Access Journals (Sweden)

    Nataly Ce

    2016-11-01

    Full Text Available This paper reports the effect of boiling synthetic seawater on the performance of damaged Thermally Sprayed Aluminum (TSA on carbon steel. Small defects (4% of the sample’s geometric surface area were drilled, exposing the steel, and the performance of the coating was analyzed for corrosion potential for different exposure times (2 h, 335 h, and 5000 h. The samples were monitored using linear polarization resistance (LPR in order to obtain their corrosion rate. Scanning electron microscopy (SEM/energy dispersive X-ray spectroscopy (EDX and X-ray diffraction (XRD were used for post-test characterization. The results showed that a protective layer of Mg(OH2 formed in the damaged area, which protected the underlying steel. Additionally, no coating detachment from the steel near the defect region was observed. The corrosion rate was found to be 0.010–0.015 mm/year after 5000 h in boiling synthetic seawater.

  2. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

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

  3. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

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

  4. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

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

  5. Thermal Protection Performance of Carbon Aerogels Filled with Magnesium Chloride Hexahydrate as a Phase Change Material

    Directory of Open Access Journals (Sweden)

    Ali Kazemi

    2014-02-01

    Full Text Available Carbon aerogels are comprised of a class of low density open-cell foams with large void space, nanometer pore size and composed of sparsely semi-colloidal nanometer sized particles forming an open porous structure. Phase change materials are those with high heat of fusion that could absorb and release a large amount of energy at the time of phase transition. These materials are mostly used as thermal energy storage materials but in addition they could serve as an obstacle for passage of heat during phase changes and this has led to their use in thermal protection systems. In this study, the effect of magnesium chloride hexahydrate, as a phase change material (melting point 115°C, on thermal properties of carbon aerogels is investigated. Thermal performance tests are designed and used for comparing the temperature-time behavior of the samples. DSC is applied to obtain the latent heat of melting of the phase change materials and the SEM tests are used to analyze the microstructure and morphology of carbon aerogels. The results show that the low percentage of phase change materials in carbon aerogels does not have any significant positive effect on carbon aerogels thermal properties. However, these properties are improved by increasing the percentage of phase change materials. With high percentage of phase change materials, a sample surface at 300°C would display an opposite surface with a significant drop in temperature increases, while at 115-200°C, with carbon aerogels, having no phase change materials, there is a severe reduction in the rate of temperature increase of the sample.

  6. Thermal stability and electrochemical properties of PVP-protected Ru nanoparticles synthesized at room temperature

    Science.gov (United States)

    Kumar, Manish; Devi, Pooja; Shivling, V. D.

    2017-08-01

    Stable ruthenium nanoparticles (RuNPs) have been synthesized by the chemical reduction of ruthenium trichloride trihydrate (RuCl3 · 3H2O) using sodium borohydride (NaBH4) as a reductant and polyvinylpyrrolidone (PVP) as a protecting agent in the aqueous medium at room temperature. The nanoparticles thus prepared were characterized by their morphology and structural analysis from transmission electron microscopy (TEM), X-ray powder diffraction (XRD), UV-vis spectroscopy, Fourier transformation infrared and thermogravimetric analysis (TGA) techniques. The TEM image suggested a homogeneous distribution of PVP-protected RuNPs having a small average diameter of 2-4 nm with a chain-like network structure. The XRD pattern also confirmed that a crystallite size is around 2 nm of PVP-protected RuNPs having a single broad peak. The thermal stability studied using TGA, indicated good stability and the electrochemical properties of these nanoparticles revealed that saturation current increases for PVP-protected RuNPs/GC.

  7. Comparison of the leading candidate combinations of blanket materials, thermodynamic cycles, and tritium systems for full scale fusion power plants

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-01-01

    The many possible combinations of blanket materials, tritium generation and recovery systems, and power conversion systems were surveyed and a comprehensive set of designs were generated by using a common set of ground rules that include all of the boundary conditions that could be envisioned for a full-scale commercial fusion power plant. Particular attention was given to the effects of blanket temperature on power plant cycle efficiency and economics, the interdependence of the thermodynamic cycle and the tritium recovery system, and to thermal and pressure stresses in the blanket structure. The results indicate that, of the wide variety of systems that have been considered, the most promising employs lithium recirculated in a closed loop within a niobium blanket structure and cooled with boiling potassium or cesium. This approach gives the simplest and lowest cost tritium recovery system, the lowest pressure and thermal stresses, the simplest structure with the lowest probability of a leak, the greatest resistance to damage from a plasma energy dump, and the lowest rate of plasma contamination by either outgassing or sputtering. The only other blanket materials combination that appears fairly likely to give a satisfactory tritium generation and recovery system is a lithium-beryllium fluoride-Incoloy blanket, and even this system involves major uncertainties in the effectiveness, size, and cost of the tritium recovery system. Further, the Li 2 BeF 4 blanket system has the disadvantage that the world reserves of beryllium are too limited to support a full-blown fusion reactor economy, its poor thermal conductivity leads to cooling difficulties and a requirement for a complex structure with intricate cooling passages, and this inherently leads to an expansive blanket with a relatively high probability of leaks. The other blanket materials combinations yield even less attractive systems

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  9. Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

    Science.gov (United States)

    Deepak, SHARMA; Paritosh, CHAUDHURI

    2018-04-01

    The Indian test blanket module (TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices (ITER relevant and DEMO). The Indian Lead–Lithium Cooled Ceramic Breeder (LLCB) blanket concept is one of the Indian DEMO relevant TBM, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept that consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Specifically, attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions. These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  11. Moving ring field-reversed mirror blanket design considerations

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-09-01

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

  13. Ballistic Performance Model of Crater Formation in Monolithic, Porous Thermal Protection Systems

    Science.gov (United States)

    Miller, J. E.; Christiansen, E. L.; Deighton, K. D.

    2014-01-01

    Porous monolithic ablative systems insulate atmospheric reentry vehicles from reentry plasmas generated by atmospheric braking from orbital and exo-orbital velocities. Due to the necessity that these materials create a temperature gradient up to several thousand Kelvin over their thickness, it is important that these materials are near their pristine state prior to reentry. These materials may also be on exposed surfaces to space environment threats like orbital debris and meteoroids leaving a probability that these exposed surfaces will be below their prescribed values. Owing to the typical small size of impact craters in these materials, the local flow fields over these craters and the ablative process afford some margin in thermal protection designs for these locally reduced performance values. In this work, tests to develop ballistic performance models for thermal protection materials typical of those being used on Orion are discussed. A density profile as a function of depth of a typical monolithic ablator and substructure system is shown in Figure 1a.

  14. A Collaborative Analysis Tool for Thermal Protection Systems for Single Stage to Orbit Launch Vehicles

    Science.gov (United States)

    Alexander, Reginald; Stanley, Thomas Troy

    2001-01-01

    Presented is a design tool and process that connects several disciplines which are needed in the complex and integrated design of high performance reusable single stage to orbit (SSTO) vehicles. Every system is linked to all other systems, as is the case with SSTO vehicles with air breathing propulsion, which is currently being studied by the National Aeronautics and Space Administration (NASA). In particular, the thermal protection system (TPS) is linked directly to almost every major system. The propulsion system pushes the vehicle to velocities on the order of 15 times the speed of sound in the atmosphere before pulling up to go to orbit which results in high temperatures on the external surfaces of the vehicle. Thermal protection systems to maintain the structural integrity of the vehicle must be able to mitigate the heat transfer to the structure and be lightweight. Herein lies the interdependency, in that as the vehicle's speed increases, the TPS requirements are increased. And as TPS masses increase the effect on the propulsion system and all other systems is compounded. To adequately calculate the TPS mass of this type of vehicle several engineering disciplines and analytical tools must be used preferably in an environment that data is easily transferred and multiple iterations are easily facilitated.

  15. New method for evaluating the kinetic constant of thermal protection materials

    International Nuclear Information System (INIS)

    Bae, Ji Yeul; Yi, Jong Ju; Park, Sul Ki; Cho, Hyung Hee; Bae, Ju Chan; Ham, Hee Cheol

    2013-01-01

    Thermal protection material (TPM) is used to protect rocket structures from extreme conditions created by the hot exhaust of the rocket. Designing TPM is an important step in the rocket design process. Considering that an increase in the system weight decreases the overall performance of a rocket, the amount of TPM is carefully determined during the design process. Therefore, the precise properties of TPM guarantee an accurate thermal analysis and the successful design of the rocket. Among the many properties of TPM, the kinetic constant and activation energy, which govern the thermochemical reaction of the TPM, are the most important. Thus, an experiment to measure the kinetic constant and activation energy is conducted as part of this research. A theoretical approach to deduce the properties from measured data is discussed, and a method to apply the theory to experimental data, termed the R 2 method, is developed. Compared to a previous method which was difficult to apply, the R 2 method reduces unclear selections of the reaction time and does not require intervention by an interpreter. The properties deduced by the R 2 method show good agreement with the other method despite the limited number of experimental results.

  16. Heat transfer characteristics and limitations analysis of heat-pipe-cooled thermal protection structure

    International Nuclear Information System (INIS)

    Guangming, Xiao; Yanxia, Du; Yewei, Gui; Lei, Liu; Xiaofeng, Yang; Dong, Wei

    2014-01-01

    The theories of heat transfer, thermodynamics and fluid dynamics are employed to develop the coupled heat transfer analytical methods for the heat-pipe-cooled thermal protection structure (HPC TPS), and a three-dimensional numerical method considering the sonic limit of heat pipe is proposed. To verify the calculation correctness, computations are carried out for a typical heat pipe and the results agree well with experimental data. Then, the heat transfer characteristics and limitations of HPC TPS are mainly studied. The studies indicate that the use of heat pipe can reduce the temperature at high heat flux region of structure efficiently. However, there is a frozen startup period before the heat pipe reaching a steady operating state, and the sonic limit will be a restriction on the heat transfer capability. Thus, the effects of frozen startup must be considered for the design of HPC TPS. The simulation model and numerical method proposed in this paper can predict the heat transfer characteristics of HPC TPS quickly and exactly, and the results will provide important references for the design or performance evaluation of HPC TPS. - Highlights: • Numerical methods for the heat-pipe-cooled thermal protection structure are studied. • Three-dimensional simulation model considering sonic limit of heat pipe is proposed. • The frozen startup process of the embedded heat pipe can be predicted exactly. • Heat transfer characteristics of TPS and limitations of heat pipe are discussed

  17. New method for evaluating the kinetic constant of thermal protection materials

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ji Yeul; Yi, Jong Ju; Park, Sul Ki; Cho, Hyung Hee [Yonsei University, Seoul (Korea, Republic of); Bae, Ju Chan; Ham, Hee Cheol [Agency for Defense Development, Daegu (Korea, Republic of)

    2013-06-15

    Thermal protection material (TPM) is used to protect rocket structures from extreme conditions created by the hot exhaust of the rocket. Designing TPM is an important step in the rocket design process. Considering that an increase in the system weight decreases the overall performance of a rocket, the amount of TPM is carefully determined during the design process. Therefore, the precise properties of TPM guarantee an accurate thermal analysis and the successful design of the rocket. Among the many properties of TPM, the kinetic constant and activation energy, which govern the thermochemical reaction of the TPM, are the most important. Thus, an experiment to measure the kinetic constant and activation energy is conducted as part of this research. A theoretical approach to deduce the properties from measured data is discussed, and a method to apply the theory to experimental data, termed the R{sup 2} method, is developed. Compared to a previous method which was difficult to apply, the R{sup 2} method reduces unclear selections of the reaction time and does not require intervention by an interpreter. The properties deduced by the R{sup 2} method show good agreement with the other method despite the limited number of experimental results.

  18. Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

    2016-11-15

    Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

  19. Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

    2016-11-01

    Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

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

    Science.gov (United States)

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

    2017-12-01

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

  1. Engineering design of a direct-cycle steam-generating blanket for a long-pulse fusion reactor

    International Nuclear Information System (INIS)

    Cort, G.E.; Hagenson, R.L.; Teasdale, R.W.; Fox, W.E.; Soran, P.D.; Cullingford, H.S.; Bathke, C.G.; Krakowski, R.A.

    1979-01-01

    A comprehensive neutronics, thermohydraulic, and mechanical design of a tritium-breeding blanket for use by a conceptual long-pulse Reversed-Field Pinch Reactor (RFPR) is described. On the basis of constraints imposed by cost and the desire to use existing technology, a direct-cycle steam system and stainless-steel construction were used. For reasons of plasma stability, the RFPR blanket supports a 20-mm-thick copper first wall. Located behind the 1.5-m-radius first wall is a 0.50-m-thick stainless-steel blanket containing a granular bed of Li 2 O through which flows low-pressure helium (0.1 MPa) for tritium extraction. Water/steam tubes radially penetrate this packed bed. The large thermal capacity and low thermal diffusivity of the Li 2 O blanket are sufficient to maintain a nearly constant temperature during the approx. 25-s burn period

  2. Investigation of the thermal behavior of 2 1/2 ton cylinder protective overpack

    International Nuclear Information System (INIS)

    Park, S.H.

    1988-01-01

    UF 6 cylinders containing reactor grade enriched uranium are transported in protective overpacks. Recently, the design of the 2 1/2 ton UF 6 cylinder overpack was modified to insure the safety of the cylinder inside the overpack. Modifications include a continuous stainless steel liner from the outer surface to the inner surface of the overpack and step joints between the upper and lower halves of the overpack. The effects of a continuous stainless steel liner and moisture in the insulation layer of a UF 6 cylinder protective overpack were investigated with a numerical code. Results were compared with limited available field data. The purpose of comparing the numerical results with field data is to insure the validity of the numerical analysis and the physical properties used in the analysis. The study indicates that the continuous stainless steel liner did not influence the heat transfer rate much from the outer surface of the overpack to the 30B cylinder inside. The effect of step joints was not modeled due to the difficulty of quantifying the leakage rate through the gap. With a continuous stainless steel liner from the outside of the overpack to the inside, the overpack satisfies the thermal design criteria of protecting the cylinder inside for a minimum of 30 minutes when the overpack is exposed to a fire. The effect of moisture inside the insulation layer in the overpack is to reduce the energy to the cylinder with its high thermal capacity. The high pressure steam generated from the moisture will be relieved externally through the vent holes on the outer surface of the overpack. Although these holes are sealed after the overpack is dried, the plug sealing the holes will melt when the overpack is exposed to a fire

  3. Space Shuttle Thermal Protection System Repair Flight Experiment Induced Contamination Impacts

    Science.gov (United States)

    Smith, Kendall A.; Soares, Carlos E.; Mikatarian, Ron; Schmidl, Danny; Campbell, Colin; Koontz, Steven; Engle, Michael; McCroskey, Doug; Garrett, Jeff

    2006-01-01

    NASA s activities to prepare for Flight LF1 (STS-114) included development of a method to repair the Thermal Protection System (TPS) of the Orbiter s leading edge should it be damaged during ascent by impacts from foam, ice, etc . Reinforced Carbon-Carbon (RCC) is used for the leading edge TPS. The repair material that was developed is named Non- Oxide Adhesive eXperimental (NOAX). NOAX is an uncured adhesive material that acts as an ablative repair material. NOAX completes curing during the Orbiter s descent. The Thermal Protection System (TPS) Detailed Test Objective 848 (DTO 848) performed on Flight LF1 (STS-114) characterized the working life, porosity void size in a micro-gravity environment, and the on-orbit performance of the repairs to pre-damaged samples. DTO 848 is also scheduled for Flight ULF1.1 (STS-121) for further characterization of NOAX on-orbit performance. Due to the high material outgassing rates of the NOAX material and concerns with contamination impacts to optically sensitive surfaces, ASTM E 1559 outgassing tests were performed to determine NOAX condensable outgassing rates as a function of time and temperature. Sensitive surfaces of concern include the Extravehicular Mobility Unit (EMU) visor, cameras, and other sensors in proximity to the experiment during the initial time after application. This paper discusses NOAX outgassing characteristics, how the amount of deposition on optically sensitive surfaces while the NOAX is being manipulated on the pre-damaged RCC samples was determined by analysis, and how flight rules were developed to protect those optically sensitive surfaces from excessive contamination where necessary.

  4. Mathematical Foundation Based Inter-Connectivity modelling of Thermal Image processing technique for Fire Protection

    Directory of Open Access Journals (Sweden)

    Sayantan Nath

    2015-09-01

    Full Text Available In this paper, integration between multiple functions of image processing and its statistical parameters for intelligent alarming series based fire detection system is presented. The proper inter-connectivity mapping between processing elements of imagery based on classification factor for temperature monitoring and multilevel intelligent alarm sequence is introduced by abstractive canonical approach. The flow of image processing components between core implementation of intelligent alarming system with temperature wise area segmentation as well as boundary detection technique is not yet fully explored in the present era of thermal imaging. In the light of analytical perspective of convolutive functionalism in thermal imaging, the abstract algebra based inter-mapping model between event-calculus supported DAGSVM classification for step-by-step generation of alarm series with gradual monitoring technique and segmentation of regions with its affected boundaries in thermographic image of coal with respect to temperature distinctions is discussed. The connectedness of the multifunctional operations of image processing based compatible fire protection system with proper monitoring sequence is presently investigated here. The mathematical models representing the relation between the temperature affected areas and its boundary in the obtained thermal image defined in partial derivative fashion is the core contribution of this study. The thermal image of coal sample is obtained in real-life scenario by self-assembled thermographic camera in this study. The amalgamation between area segmentation, boundary detection and alarm series are described in abstract algebra. The principal objective of this paper is to understand the dependency pattern and the principles of working of image processing components and structure an inter-connected modelling technique also for those components with the help of mathematical foundation.

  5. Effect of thermal protectants on the stability of bovine milk immunoglobulin G

    Energy Technology Data Exchange (ETDEWEB)

    Chen, C. C. [National Taiwan University, Taipei, Taiwan (China); Chang, H. M.

    1998-09-15

    pH stability, thermal stability, and the effect of homogenization and ultrasonic treatment on the stability of bovine milk immunoglobulin G (IgG) in model systems was studied. Separated IgG (0.02 mg/mL) was found to be unstable and susceptible to denaturation when incubated at pH 4 or 10 or thermally treated at temperature 75 degrees C. IgG in the colostrum, on the other hand, was found to be much more stable than in whey or in PBS when thermally treated at temperatures in the range of 75-100 degrees C. The residual IgC content reduced more sharply with increasing heating times, and almost no IgG content was detected when IgG in PBS (0.15 M NaCl/0.01 M phosphate buffer, pH 7.0) was heated at 95 degrees C for 15 s, whereas the corresponding residual IgG contents in whey and colostrum were found to be 42 and 59%, respectively. For IgG in PBS heated at 95 degrees C for 15 s, addition of 5% fructose or maltose displayed most remarkable protection effects by raising the residual IgG content to 31%, followed by sucrose, lactose, glucose, and galactose. However, extravagant addition ( 30%) to IgG in PBS led to a decline in residual IgG content. Addition of 0.4% glutamic acid and 2% glycine to IgG in PBS heated at 95 degrees C for 15 s also remarkably improved the residual IgG content by 13.5 and 26.7%, respectively. Glycerol and sugar alcohol, such as sorbitol, stabilized IgG during the thermal treatment.

  6. Effect of thermal protectants on the stability of bovine milk immunoglobulin G

    International Nuclear Information System (INIS)

    Chen, C.C.; Chang, H.M.

    1998-01-01

    pH stability, thermal stability, and the effect of homogenization and ultrasonic treatment on the stability of bovine milk immunoglobulin G (IgG) in model systems was studied. Separated IgG (0.02 mg/mL) was found to be unstable and susceptible to denaturation when incubated at pH 4 or 10 or thermally treated at temperature 75 degrees C. IgG in the colostrum, on the other hand, was found to be much more stable than in whey or in PBS when thermally treated at temperatures in the range of 75-100 degrees C. The residual IgC content reduced more sharply with increasing heating times, and almost no IgG content was detected when IgG in PBS (0.15 M NaCl/0.01 M phosphate buffer, pH 7.0) was heated at 95 degrees C for 15 s, whereas the corresponding residual IgG contents in whey and colostrum were found to be 42 and 59%, respectively. For IgG in PBS heated at 95 degrees C for 15 s, addition of 5% fructose or maltose displayed most remarkable protection effects by raising the residual IgG content to 31%, followed by sucrose, lactose, glucose, and galactose. However, extravagant addition ( 30%) to IgG in PBS led to a decline in residual IgG content. Addition of 0.4% glutamic acid and 2% glycine to IgG in PBS heated at 95 degrees C for 15 s also remarkably improved the residual IgG content by 13.5 and 26.7%, respectively. Glycerol and sugar alcohol, such as sorbitol, stabilized IgG during the thermal treatment

  7. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

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

    1995-05-01

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

  8. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

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

    1999-07-01

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

  9. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

    Tokuhiro, A.

    1991-11-01

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

  10. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  11. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  12. Beryllium R&D for blanket application

    Science.gov (United States)

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

    1998-10-01

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

  13. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  14. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  15. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

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

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-22

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

  17. Investigation of heat treatment conditions of structural material for blanket fabrication process

    International Nuclear Information System (INIS)

    Hirose, Takanori; Suzuki, Satoshi; Akiba, Masato; Shiba, Kiyoyuki; Sawai, Tomotsugu; Jitsukawa, Shiro

    2004-01-01

    This paper presents recent results of thermal hysteresis effects on ceramic breeder blanket structural material. Reduced activation ferritic/martensitic (RAF) steel is the leading candidates for the first wall structural materials of breeding blankets. RAF steel demonstrates superior resistance to high dose neutron irradiation, because the steel has tempered martensite structure which contains the number of sink site for radiation defects. This microstructure obtained by two-step heat treatment, first is normalizing at temperature above 1200 K and the second is tempering at temperature below 1100 K. Recent study revealed the thermal hysteresis has significant impacts on the post-irradiation mechanical properties. The breeding blanket has complicated structure, which consists of tungsten armor and thin first wall with cooling pipe. The blanket fabrication requires some high temperature joining processes. Especially hot isostatic pressing (HIP) is examined as a near-net-shape fabrication process for this structure. The process consists of heating above 1300 K and isostatic pressing at the pressure above 150 MPa followed by tempering. Moreover ceramics pebbles are packed into blanket module and the module is to be seamed by welding followed by post weld heat treatment in the final assemble process. Therefore the final microstructural features of RAFs strongly depend on the blanket fabrication process. The objective of this work is to evaluate the effects of thermal hysteresis corresponding to blanket fabrication process on RAFs microstructure in order to establish appropriate blanket fabrication process. Japanese RAFs F82H (Fe-0.1C-8Cr-2W-0.2V-0.05Ta) was investigated by metallurgical method after isochronal heat treatment up to 1473 K simulating high temperature bonding process. Although F82H showed significant grain growth after conventional solid HIP conditions (1313 K x 2 hr.), this coarse grained microstructure was refined by the post HIP normalizing at

  18. Electrical connectors for blanket modules in ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  19. Preconceptual design of a packed fluidized bed blanket for a fission suppressed thorium-fueled CTHR

    International Nuclear Information System (INIS)

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

    1981-01-01

    This paper describes a thorium-fueled PFB blanket concept for a Commercial Tokamak Hybrid Reactor. A preliminary mechanical concept is presented and the results of neutronics, thermal-hydraulics and economics analyses are discussed. Futher work needed to design and advance the concept is recommended

  20. Thermal Protection Performance of Phase Changing Material Based on Polyethylene Glycol

    Directory of Open Access Journals (Sweden)

    Leila Sadat Ahmadi

    2012-12-01

    Full Text Available Phase change materials (PCM are substances with a high heat of fusion which, through melting and solidifying at certain temperatures, are capable to store or release a large amount of energy. This phenomenon can be utilized in designing heat protective materials as well as in thermal energy storage systems. One of the approaches to avoid materials leaching from a structure, where PCMs are incorporated, is to blend them with suitable polymers. To have a proper blend it is necessary to choose a compatible polymer with a PCM. It is important to assess the optimized concentration of PCM in polymer matrix and the phase structure and morphology of the blend, which causes the best heat protection. In this work, the influence of polyethylene glycol (PEG as PCMs in epoxy resin matrix on heat protection was investigated. A special performance test was designed to study timetemperature behavior of the prepared samples and DSC and SEM tests to observe the melting point, heat of fusion and morphology of the samples. The results indicated that increases in PCM content led to better heat protection and the best concentration for PEG was found to be 60% wt. Time-temperature curves show that increases of temperature for PCM samples is very slow compared with net epoxy sample. PCM samples curves show plateau in melting region. In this region, they show nearly 15°C temperature lower than a net epoxy sample. The plateau region makes a delay time in temperature increment, which is about 22 min for PEG samples compared with a net epoxy.

  1. Determination of Acreage Thermal Protection Foam Loss From Ice and Foam Impacts

    Science.gov (United States)

    Carney, Kelly S.; Lawrence, Charles

    2015-01-01

    A parametric study was conducted to establish Thermal Protection System (TPS) loss from foam and ice impact conditions similar to what might occur on the Space Launch System. This study was based upon the large amount of testing and analysis that was conducted with both ice and foam debris impacts on TPS acreage foam for the Space Shuttle Project External Tank. Test verified material models and modeling techniques that resulted from Space Shuttle related testing were utilized for this parametric study. Parameters varied include projectile mass, impact velocity and impact angle (5 degree and 10 degree impacts). The amount of TPS acreage foam loss as a result of the various impact conditions is presented.

  2. Fracture Toughness Evaluation of Space Shuttle External Tank Thermal Protection System Polyurethane Foam Insulation Materials

    Science.gov (United States)

    McGill, Preston; Wells, Doug; Morgan, Kristin

    2006-01-01

    Experimental evaluation of the basic fracture properties of Thermal Protection System (TPS) polyurethane foam insulation materials was conducted to validate the methodology used in estimating critical defect sizes in TPS applications on the Space Shuttle External Fuel Tank. The polyurethane foam found on the External Tank (ET) is manufactured by mixing liquid constituents and allowing them to react and expand upwards - a process which creates component cells that are generally elongated in the foam rise direction and gives rise to mechanical anisotropy. Similarly, the application of successive foam layers to the ET produces cohesive foam interfaces (knitlines) which may lead to local variations in mechanical properties. This study reports the fracture toughness of BX-265, NCFI 24-124, and PDL-1034 closed-cell polyurethane foam as a function of ambient and cryogenic temperatures and knitline/cellular orientation at ambient pressure.

  3. The Evolution of Nondestructive Evaluation Methods for the Space Shuttle External Tank Thermal Protection System

    Science.gov (United States)

    Walker, James L.; Richter, Joel D.

    2006-01-01

    Three nondestructive evaluation methods are being developed to identify defects in the foam thermal protection system (TPS) of the Space Shuttle External Tank (ET). Shearography is being developed to identify shallow delaminations, shallow voids and crush damage in the foam while terahertz imaging and backscatter radiography are being developed to identify voids and cracks in thick foam regions. The basic theory of operation along with factors affecting the results of these methods will be described. Also, the evolution of these methods from lab tools to implementation on the ET will be discussed. Results from both test panels and flight tank inspections will be provided to show the range in defect sizes and types that can be readily detected.

  4. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. Thermal spraying of corrosion protection layers in biogas plants; Erzeugung von Korrosionsschutzschichten fuer Bioenergieanlagen mittels Thermischen Spritzens

    Energy Technology Data Exchange (ETDEWEB)

    Crimmann, P.; Dimaczek, G.; Faulstich, M. [ATZ Entwicklungszentrum, Sulzbach-Rosenberg (Germany)

    2004-07-01

    Corrosion in plants for the energetic conversion of biomass is a severe problem that often causes premature damage of components. Thermal spraying is a process for the creation of corrosion protection layer. An advantage of thermal spraying is that as well as each material can be used as layer material. First practical results demonstrated that thermal spraying has the potential to create coatings to protect components against high temperature corrosion as well as biocorrosion. Layer materials are for example nickel base alloys (high temperature corrosion) and titan alloys (biocorrosion). Further investigations are necessary in order to examine whether cost-efficient coatings also contribute to the corrosion protection (e.g. polymer materials against biocorrosion). (orig.)

  7. Blanket design for imploding liner systems

    International Nuclear Information System (INIS)

    Schaffer, M. J.

    1980-01-01

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

  8. APT target-blanket fabrication development

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, D.L.

    1997-06-13

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

  9. Final analysis and design of a thermal protection system for 8-foot HTST combustor

    Science.gov (United States)

    Moskowitz, S.

    1973-01-01

    The cylindrical shell combustor with T-bar supports in the 8-foot HTST at the NASA-Langley Research Center encountered vibratory fatigue cracking over a period of 50-250 tunnel tests within a limited range of the required operating envelope. A preliminary design study provided several suitable thermal protection system designs for the combustor, one of which was a two-pass regenerative type air-cooled omega-shaped segment liner. A final design layout of the omega segment liner was prepared and analyzed for steady-state and transient conditions. The design of a support system for the fuel spray bar assembly was also included. Detail drawings suitable for fabrication purposes were also prepared. Liner design problems defined during the preliminary study included (1) the ingress of gas into the attachment bulb section of the omega segment, (2) the large thermal gradient along the leg of the omega bulb attachment section and, (3) the local peak metal temperature at the radius between the liner ID and the leg of the bulb attachment. These were resolved during the final design task. Analyses of the final design of the omega segment liner indicated that all design goals were met and the design provided the capability of operating over the required test envelope with a life expectancy substantially above the goal of 1500 cycles.

  10. Thermal Protection System Mass Estimating Relationships For Blunt-Body, Earth Entry Spacecraft

    Science.gov (United States)

    Sepka, Steven A.; Samareh, Jamshid A.

    2015-01-01

    Mass estimating relationships (MERs) are developed to predict the amount of thermal protection system (TPS) necessary for safe Earth entry for blunt-body spacecraft using simple correlations that are non-ITAR and closely match estimates from NASA's highfidelity ablation modeling tool, the Fully Implicit Ablation and Thermal Analysis Program (FIAT). These MERs provide a first order estimate for rapid feasibility studies. There are 840 different trajectories considered in this study, and each TPS MER has a peak heating limit. MERs for the vehicle forebody include the ablators Phenolic Impregnated Carbon Ablator (PICA) and Carbon Phenolic atop Advanced Carbon-Carbon. For the aftbody, the materials are Silicone Impregnated Reusable Ceramic Ablator (SIRCA), Acusil II, SLA- 561V, and LI-900. The MERs are accurate to within 14% (at one standard deviation) of FIAT prediction, and the most any MER can under predict FIAT TPS thickness is 18.7%. This work focuses on the development of these MERs, the resulting equations, model limitations, and model accuracy.

  11. Ceramic Matrix Composite (CMC) Thermal Protection Systems (TPS) and Hot Structures for Hypersonic Vehicles

    Science.gov (United States)

    Glass, David E.

    2008-01-01

    Thermal protection systems (TPS) and hot structures are required for a range of hypersonic vehicles ranging from ballistic reentry to hypersonic cruise vehicles, both within Earth's atmosphere and non-Earth atmospheres. The focus of this paper is on air breathing hypersonic vehicles in the Earth's atmosphere. This includes single-stage to orbit (SSTO), two-stage to orbit (TSTO) accelerators, access to space vehicles, and hypersonic cruise vehicles. This paper will start out with a brief discussion of aerodynamic heating and thermal management techniques to address the high heating, followed by an overview of TPS for rocket-launched and air-breathing vehicles. The argument is presented that as we move from rocket-based vehicles to air-breathing vehicles, we need to move away from the insulated airplane approach used on the Space Shuttle Orbiter to a wide range of TPS and hot structure approaches. The primary portion of the paper will discuss issues and design options for CMC TPS and hot structure components, including leading edges, acreage TPS, and control surfaces. The current state-of-the-art will be briefly discussed for some of the components. The two primary technical challenges impacting the use of CMC TPS and hot structures for hypersonic vehicles are environmental durability and fabrication, and will be discussed briefly.

  12. Monitoring of Thermal Protection Systems and MMOD using Robust Self-Organizing Optical Fiber Sensing Networks

    Science.gov (United States)

    Richards, Lance

    2014-01-01

    The general aim of this work is to develop and demonstrate a prototype structural health monitoring system for thermal protection systems that incorporates piezoelectric acoustic emission (AE) sensors to detect the occurrence and location of damaging impacts, such as those from Micrometeoroid Orbital Debris (MMOD). The approach uses an optical fiber Bragg grating (FBG) sensor network to evaluate the effect of detected damage on the thermal conductivity of the TPS material. Following detection of an impact, the TPS would be exposed to a heat source, possibly the sun, and the temperature distribution on the inner surface in the vicinity of the impact measured by the FBG network. A similar procedure could also be carried out as a screening test immediately prior to re-entry. The implications of any detected anomalies in the measured temperature distribution will be evaluated for their significance in relation to the performance of the TPS during reentry. Such a robust TPS health monitoring system would ensure overall crew safety throughout the mission, especially during reentry.

  13. Key achievements in elementary R and Ds on water-cooled solid breeder blanket for ITER Test Blanket Module in JAERI

    International Nuclear Information System (INIS)

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

    2005-01-01

    This paper presents significant progress in research and development (R and D) of key elementary technologies on the water-cooled solid breeder blanket for the ITER test blanket modules (TBMs) in JAERI. Development of module fabrication technology, bonding technology of armors, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup, and tritium release behavior from Li 2 TiO 3 pebble bed under neutron pulsed operation condition are summarized. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 deg C followed by normalizing at 930 deg C after the Hot Isostatic Pressing (HIP) process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a solid state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it was found that the thermal fatigue lifetime of F82H can be predicted by using Manson-Coffin's law. As for R and Ds on a breeder material, Li 2 TiO 3 , effective thermal conductivity of Li 2 TiO 3 pebble was measured under compressive force simulating the ITER TBM environment. The increase in the effective thermal conductivity of the pebble bed was about 2.5 % at the compressive strain of 0.9 % at 400 deg C. Neutronic performance of the blanket module mockup has been carried out by the 14 MeV neutron irradiation. It was confirmed that the measured tritium production rate agreed with the calculated values within about 10% difference. Also, tritium release from a Li 2 TiO 3 pebble bed was measured under pulsed neutron irradiation conditions simulating the ITER operation. (author)

  14. The fusion blanket program at Chalk River

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-03-01

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

  15. Three-dimensional biomimetic head model as a platform for thermal testing of protective goggles for prevention of eye injuries.

    Science.gov (United States)

    Friedman, Rinat; Haimy, Ayelet; Gefen, Amit; Epstein, Yoram

    2018-04-22

    The rate of eye injury is steadily rising during military conflicts of the century, with thermal burns being the most common type of injury to the eyes. The present study focuses on assessing the heat resistance properties of military protective goggles using three-dimensional (3D) finite element head modeling fitted with the tested protective gear. A computational thermal impact was applied onto a 3D biomimetic human head model fitted with two goggle models - sports (Type 1) and square (Type 2). The resultant temperature of the eye tissues and the thermal injury thresholds were calculated by using the modeling, hence allowing to determine the protective efficacy of the goggles objectively, in a standardized, quantitative and cost-effective manner. Both types of goggles had a dramatic protective effect on the eyes. The specific goggle geometry had no notable effect on the level of protection to the inner tissues against the thermal insult. At the skin level goggles reduced temperatures by ~64% under the impact zone, with only a mild difference (10 °C) between the goggles. Little limitations on the shape and geometry of goggles were observed and any structure of goggles can provide an adequate protection against a thermal insult (per se) to inner cranial tissues, assuming the lenses are wide and thick enough to block direct skin contact of the heat insult. It was shown that our 3D biomimetic human head model provides a practical and cost-effective tool for determining the performance level of goggles with different attributed (i.e., shapes and thermal properties). Copyright © 2018 Elsevier Ltd. All rights reserved.

  16. Evaluation of heat transfer characteristics of a sphere-packed pipe for Flibe blanket

    International Nuclear Information System (INIS)

    Watanabe, Atsushi; Ebara, Shinji; Sagara, Akio; Hashizume, Hidetoshi

    2013-01-01

    A Flibe blanket has been proposed to be used in FFHR. Since Flibe has poor heat transfer performance, heat transfer promoter is required, and a sphere-packed pipe (SPP) has been proposed to enhance the heat transfer performance in the Flibe blanket. In this paper, the fluid flow and heat transfer characteristics in the SPP is evaluated numerically using a k–ε turbulent model for the flow field and an algebraic model for the thermal field. As a result, it was shown that bypass flows in the SPP play a significant role in heat transfer. Also it is thought that the turbulent energy can strongly affect heat transfer performance

  17. Evaluation of steam as a potential coolant for nonbreeding blanket designs

    International Nuclear Information System (INIS)

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1978-01-01

    A steam-cooled nonbreeding blanket design has been developed as an evolution of the Argonne Experimental Power Reactor (EPR) studies. This blanket concept complete with maintenance considerations is to function at temperatures up to 650 0 C utilizing nickel-based alloys such as Inconel 625. Thermo-mechanical analyses were carried out in conjunction with thermal hydraulic analysis to determine coolant chennel arrangements that permit delivery of superheated steam at 500 0 C directly to a modern fossil plant-type turbine. A dual-cycle system combining a pressurized water circuit coupled with a superheated steam circuit can produce turbine plant conversion efficiencies approaching 41.5%

  18. Size limitations for microwave cavity to simulate heating of blanket material in fusion reactor

    International Nuclear Information System (INIS)

    Wolf, D.

    1987-01-01

    The power profile in the blanket material of a nuclear fusion reactor can be simulated by using microwaves at 200 MHz. Using these microwaves, ceramic breeder materials can be thermally tested to determine their acceptability as blanket materials without entering a nuclear fusion environment. A resonating cavity design is employed which can achieve uniform cross sectional heating in the plane transverse to the neutron flux. As the sample size increases in height and width, higher order modes, above the dominant mode, are propagated and destroy the approximation to the heating produced in a fusion reactor. The limits at which these modes develop are determined in the paper

  19. Cost study of the ESPRESSO blanket for a Tandem Mirror Reactor

    International Nuclear Information System (INIS)

    Raffray, A.R.; Hoffman, M.A.; Gaskins, T.

    1986-02-01

    A detailed cost study of the ESPRESSO blanket concept for the Tandem Mirror Fusion Reactor (TMR) has been performed to complement the thermal-hydraulic parametric study and to help narrow down the choice of parameters for the final design. The ESPRESSO blanket consists of a number of structurally independent ring modules. Each ring module is made up of a number of mutually pressure-supporting canisters containing arrays of breeder tubes. Two separate helium coolant flows are used: a main flow to cool the tube bank and a cooler first wall flow

  20. Workshop on cold-blanket research

    International Nuclear Information System (INIS)

    1977-05-01

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

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

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2004-07-01

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

  2. Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Aktaa, J., E-mail: jarir.aktaa@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Kecskés, S.; Pereslavtsev, P.; Fischer, U.; Boccaccini, L.V. [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-10-15

    Highlights: • First non-linear structural analysis for the European Helium Cooled Pebble Bed Blanket Module taking into account high dose irradiation. • Most critical areas were identified and analyzed with regard to the effect of irradiation on predicted damage at these areas. • Despite the extensive computing time 100 cycles were simulated by using the sub-modelling technique investigating damage at most critical area. • The results show a positive effect of irradiation on calculated damage which is mainly attributed to the irradiation induced hardening. - Abstract: For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design. For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of

  3. Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

    International Nuclear Information System (INIS)

    Aktaa, J.; Kecskés, S.; Pereslavtsev, P.; Fischer, U.; Boccaccini, L.V.

    2014-01-01

    Highlights: • First non-linear structural analysis for the European Helium Cooled Pebble Bed Blanket Module taking into account high dose irradiation. • Most critical areas were identified and analyzed with regard to the effect of irradiation on predicted damage at these areas. • Despite the extensive computing time 100 cycles were simulated by using the sub-modelling technique investigating damage at most critical area. • The results show a positive effect of irradiation on calculated damage which is mainly attributed to the irradiation induced hardening. - Abstract: For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design. For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of

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

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, Kenneth Mitchell [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1994-04-01

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

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

    OpenAIRE

    畠中, 龍太; 宮北, 健; 杉田, 寛之; Saitoh, Masanori; Hirai, Tomoyuki; Hatakenaka, Ryuta; Miyakita, Takeshi; Sugita, Hiroyuki; Saitoh, Masanori; Hirai, Tomoyuki

    2014-01-01

    New types of MLI blanket have been developed to achieve high thermal performance while maintaining production and assembly workability equivalent to the conventional type. Tag-pins, which are widely used in commercial applications to hook price tags to products, are used to fix the films in place and the pin material is changed to polyetheretherketone (PEEK) for use in space. Thermal performance is measured by using a boil-off calorimeter, in which a rectangular liquid nitrogen tank is used t...

  6. First wall and blanket stresses induced by cyclic fusion core operations

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.; Kostoff, R.N.

    1981-01-01

    An analysis is made of cyclic thermal loads and stresses for the complete range of operating conditions. Two critical components were examined; the solid wall adjacent to the fusion plasma (first wall) and the fuel elements in the high power density region of the blanket. Simple closed form expressions were derived for temperature increases and thermal stresses that may be evaluated conveniently and rapidly and the values compared for different systems

  7. The power-sharing formula for a seed/blanket core-resolution of a paradox

    International Nuclear Information System (INIS)

    Radkowsky, A.; Segev, M.; Galperin, A.

    1986-01-01

    The ''classical'' formula for the sharing of power between a seed and blanket was based on the two-group diffusion theory model and gave good agreement with experiments conducted in the original Shippingport program and with transport theory. Recently an extensive series of calculations on seed/blanket assemblies showed that the power sharing deviates widely from the classical formula but paradoxically is in good agreement with the one-group formula, which neglects the back leakage of thermal neutrons from the blanket to the seed. The power-sharing formula has now been rederived, and the paradox is resolved by taking into account epithermal absorptions in the seeds. The diffusion theory model is important as a guide to formulating innovative concepts for improved core designs

  8. Design and R and D activities on ceramic breeder blanket for fusion experimental reactors in JAERI

    International Nuclear Information System (INIS)

    Kurasawa, T.; Takatsu, H.; Sato, S.; Nakahira, M.; Furuya, K.; Hashimoto, T.; Kawamura, H.; Kuroda, T.; Tsunematsu, T.; Seki, M.

    1995-01-01

    Design and R and D activities on ceramic breeder blanket of a fusion experimental reactor have been progressed in JAERI. A layered pebble bed type ceramic breeder blanket with water cooling is a prime candidate concept. Design activities have been concentrated on improvement of the design by conducting detailed analyses and also by fabrication procedure consideration based on the current technologies. A wide variety of R and Ds have also been conducted in accordance with the design activities. Development of fabrication technology of the blanket box structure and its mechanical testing, elementary testing on thermal performances of the pebble bed, and engineering-oriented material tests of breeder and beryllium pebbles are the main achievements during the last two years. (orig.)

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  10. Thermostructural design of the first wall/blanket for the TITAN-RFP fusion reactor

    International Nuclear Information System (INIS)

    Orient, G.E.; Blanchard, J.P.; Ghoniem, N.M.

    1987-01-01

    The mass power density, which is defined as the average power per unit mass within the magnet boundary, is a rough and general measure of economic competitiveness. Conn et al. (1985) have identified a target value of 100 kW(e)/tonne as a reasonable threshold for 'compact' commercial fusion systems. In pursuit of this goal, Hagenson et al. (1984) and Najmabadi et al. (1987) have pointed out the inherent characteristics of the RFP toroidal confinement concept which allow it to exceed this target value. It is inevitable that the compactness of the fusion power core will introduce a unique set of design issues. The special design concerns stem from high thermal surface fluxes, high bulk energy deposition by neutrons, and a relatively short blanket structural lifetime. In the TITAN-RFP, study Najmabadi et al. (1987) investigate a number of blanket (B) and first wall (FW) options suitable for high power density fusion reactors. Final choices were made for two designs: A high pressure aqueous blanket and a vanadium/lithium self-cooled blanket. The first design utilizes a pressurized aqueous loop containing a lithium compound dissolved in water, while the second design is based upon a self-cooled lithium-vanadium blanket. In this paper, we consider the beginning-of-life (BOL) thermostructural design and analysis of only the second concept. (orig./GL)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    2000-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Li Chunjing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China)], E-mail: lcj@ipp.ac.cn; Huang Qunying; Wu Qingsheng; Liu Shaojun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Lei Yucheng [Jiangsu University, Zhenjiang, Jiangsu, 212013 (China); Muroga, Takeo; Nagasaka, Takuya [National Institute for Fusion Science, Toki, Jifu, 509-5292 (Japan); Zhang Jianxun [Xi' an Jiaotong University, Xi' an, Shanxi, 710049 (China); Li Jinglong [Northwestern Polytechnical University, Xi' an, Shanxi, 710072 (China)

    2009-06-15

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

  14. Thermal Protection System Cavity Heating for Simplified and Actual Geometries Using Computational Fluid Dynamics Simulations with Unstructured Grids

    Science.gov (United States)

    McCloud, Peter L.

    2010-01-01

    Thermal Protection System (TPS) Cavity Heating is predicted using Computational Fluid Dynamics (CFD) on unstructured grids for both simplified cavities and actual cavity geometries. Validation was performed using comparisons to wind tunnel experimental results and CFD predictions using structured grids. Full-scale predictions were made for simplified and actual geometry configurations on the Space Shuttle Orbiter in a mission support timeframe.

  15. To a question on thermal protection of constructional elements of vacuum-plasma devices

    International Nuclear Information System (INIS)

    Borisko, V.N.; Borisko, S.V.; Zinovev, D.V.; Lapshin, V.I.; Tselujko, A.F.

    2005-01-01

    The progress in development of vacuum-plasma devices is connected with the design and creation of constructional elements from materials, which have a high erosion resistance and can maintain the large specific flux of energy per effective area. Recently as the materials of such constructional elements it was offered to use the reversible sorbents of hydrogen of Zr-V system, which have high-rates of sorption-desorption and large thermal effect of the hydride phases decomposition. In the paper an experimental research of the thermal conditions features of the metal-hydride electrodes, which subjected of the energy loads in the vacuum-plasma devices, are given. The simulation of the energy loads on the electrodes was carried out with the help of gas discharge plasma as there is an possibility to vary the energy spectrum of the bombarding particles and to gather a necessary radiation dose to the material surface. For comparative examinations of various materials under the irradiation by high-energy heavy particles it is the most convenient to use the Penning discharge. In this case, the cathodes made of different materials are under the identical conditions even at the change of working discharge modes. Therefore in the device on the basis of the Penning discharge the cathodes of metal-hydride and stainless steel were set. It was detected, that the increase of the temperature gradient of metal-hydride cathode goes down with the increase of discharge current value. The dependence of operating temperatures difference of cathodes from exposure time has shown that the temperature of the metal-hydride cathode is sufficiently lower than the temperature of the stainless steel cathode. Such a softening of the thermal operation conditions of the metal hydride cathode is caused by thermal decomposition of hydride phases. Besides there is the energy flow dissipation of bombarding particles on the protective gas target formed by desorbed hydrogen. The considerable decrease of

  16. Test Blanket Working Group's recent activities

    International Nuclear Information System (INIS)

    Vetter, J.E.

    2001-01-01

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

  17. Strategy for the development of EU Test Blanket Systems instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Calderoni, P., E-mail: Pattrick.Calderoni@f4e.europa.eu; Ricapito, I.; Poitevin, Y.

    2013-10-15

    Highlights: • We developed a strategy for the development of instrumentation for EU ITER TBSs. • TBSs instrumentation functions: safety, operation and scientific mission. • Described activities are in support of ITER design review process. -- Abstract: The instrumentation of the HCLL and HCPB Test Blanket System is fundamental in ensuring that ITER safety and operational requirements are satisfied as well as in enabling the scientific mission of the TBM program. It carries out three essential functions: (i) safety, intended as compliance with ITER requirements toward public and workers protection; (ii) system control, intended as compliance with ITER operational requirements and investment protection; and (iii) scientific mission, intended as validating technology and predictive tools for blanket concepts relevant to fusion energy systems. This paper describes the strategy for instrumentation development by providing details of the following five steps to be implemented in procured activities in the short to mid-term (3–4 years): (i) provide mapping of sensors requirements based on critical review of preliminary design data; (ii) develop functional specifications for TBS sensors based on the analysis of operative conditions in the various ITER buildings in which they are located; (iii) assess availability of commercial sensors against developed specifications; (iv) develop prototypes when no available solution is identified; and (v) perform single effect tests for the most critical solicitations and post-test examination of commercial products and prototypes. Examples of technology assessment in two technical areas are included to reinforce and complement the strategy description.

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

    International Nuclear Information System (INIS)

    Bohm, Tim; Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul; Ulrickson, Michael; Bullock, James

    2015-01-01

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  20. Flutter Analysis of the Thermal Protection Layer on the NASA HIAD

    Science.gov (United States)

    Goldman, Benjamin D.; Dowell, Earl H.; Scott, Robert C.

    2013-01-01

    A combination of classical plate theory and a supersonic aerodynamic model is used to study the aeroelastic flutter behavior of a proposed thermal protection system (TPS) for the NASA HIAD. The analysis pertains to the rectangular configurations currently being tested in a NASA wind-tunnel facility, and may explain why oscillations of the articles could be observed. An analysis using a linear flat plate model indicated that flutter was possible well within the supersonic flow regime of the wind tunnel tests. A more complex nonlinear analysis of the TPS, taking into account any material curvature present due to the restraint system or substructure, indicated that significantly greater aerodynamic forcing is required for the onset of flutter. Chaotic and periodic limit cycle oscillations (LCOs) of the TPS are possible depending on how the curvature is imposed. When the pressure from the base substructure on the bottom of the TPS is used as the source of curvature, the flutter boundary increases rapidly and chaotic behavior is eliminated.

  1. Spatially distributed damage detection in CMC thermal protection materials using thin-film piezoelectric sensors

    Science.gov (United States)

    Kuhr, Samuel J.; Blackshire, James L.; Na, Jeong K.

    2009-03-01

    Thermal protection systems (TPS) of aerospace vehicles are subjected to impacts during in-flight use and vehicle refurbishment. The damage resulting from such impacts can produce localized regions that are unable to resist extreme temperatures. Therefore it is essential to have a reliable method to detect, locate, and quantify the damage occurring from such impacts. The objective of this research is to demonstrate a capability that could lead to detecting, locating and quantifying impact events for ceramic matrix composite (CMC) wrapped tile TPS via sensors embedded in the TPS material. Previous research had shown a correlation between impact energies, material damage state, and polyvinylidene fluoride (PVDF) sensor response for impact energies between 0.07 - 1.00 Joules, where impact events were located directly over the sensor positions1. In this effort, the effectiveness of a sensor array is evaluated for detecting and locating low energy impacts on a CMC wrapped TPS. The sensor array, which is adhered to the internal surface of the TPS tile, is used to detect low energy impact events that occur at different locations. The analysis includes an evaluation of signal amplitude levels, time-of-flight measurements, and signal frequency content. Multiple impacts are performed at each location to study the repeatability of each measurement.

  2. A new surface catalytic model for silica-based thermal protection material for hypersonic vehicles

    Directory of Open Access Journals (Sweden)

    Li Kai

    2015-10-01

    Full Text Available Silica-based materials are widely employed in the thermal protection system for hypersonic vehicles, and the investigation of their catalytic characteristics is crucially important for accurate aerothermal heating prediction. By analyzing the disadvantages of Norman’s high and low temperature models, this paper combines the two models and proposes an eight-reaction combined surface catalytic model to describe the catalysis between oxygen and silica surface. Given proper evaluation of the parameters according to many references, the recombination coefficient obtained shows good agreement with experimental data. The catalytic mechanisms between oxygen and silica surface are then analyzed. Results show that with the increase of the wall temperature, the dominant reaction contributing to catalytic coefficient varies from Langmuir–Hinshelwood (LH recombination (TW  1350 K. The surface coverage of chemisorption areas varies evidently with the dominant reactions in the high temperature (HT range, while the surface coverage of physisorption areas varies within quite low temperature (LT range (TW < 250 K. Recommended evaluation of partial parameters is also given.

  3. Investigation of Post-Flight Solid Rocket Booster Thermal Protection System

    Science.gov (United States)

    Nelson, Linda A.

    2006-01-01

    After every Shuttle mission, the Solid Rocket Boosters (SRBs) are recovered and observed for missing material. Most of the SRB is covered with a cork-based thermal protection material (MCC-l). After the most recent shuttle mission, STS-114, the forward section of the booster appeared to have been impacted during flight. The darkened fracture surfaces indicated that this might have occurred early in flight. The scope of the analysis included microscopic observations to assess the degree of heat effects and locate evidence of the impact source as well as chemical analysis of the fracture surfaces and recovered foreign material using Fourier Transform Infrared Spectroscopy and Scanning Electron Microscopy/Energy Dispersive Spectroscopy. The amount of heat effects and presence of soot products on the fracture surface indicated that the material was impacted prior to SRB re-entry into the atmosphere. Fragments of graphite fibers found on these fracture surfaces were traced to slag inside the Solid Rocket Motor (SRM) that forms during flight as the propellant is spent and is ejected throughout the descent of the SRB after separation. The direction of the impact mark matches with the likely trajectory of SRBs tumbling prior to re-entry.

  4. In-Space Repair of Reinforced Carbon-Carbon Thermal Protection System Structures

    Science.gov (United States)

    Singh, Mrityunjay

    2006-01-01

    Advanced repair and refurbishment technologies are critically needed for the thermal protection system of current space transportation system as well as for future Crew Exploration Vehicles (CEV). The damage to these components could be caused by impact during ground handling or due to falling of ice or other objects during launch. In addition, in-orbit damage includes micrometeoroid and orbital debris impact as well as different factors (weather, launch acoustics, shearing, etc.) during launch and re-entry. The GRC developed GRABER (Glenn Refractory Adhesive for Bonding and Exterior Repair) material has shown multiuse capability for repair of small cracks and damage in reinforced carbon-carbon (RCC) material. The concept consists of preparing an adhesive paste of desired ceramic with appropriate additives and then applying the paste to the damaged/cracked area of the RCC composites with adhesive delivery system. The adhesive paste cures at 100-120 C and transforms into a high temperature ceramic during simulated entry conditions. A number of plasma torch and ArcJet tests were carried out to evaluate the crack repair capability of GRABER materials for Reinforced Carbon-Carbon (RCC) composites. For the large area repair applications, integrated system for tile and leading edge repair (InSTALER) have been developed. In this presentation, critical in-space repair needs and technical challenges as well as various issues and complexities will be discussed along with the plasma performance and post test characterization of repaired RCC materials.

  5. Laser Shearography Inspection of TPS (Thermal Protection System) Cork on RSRM (Reusable Solid Rocket Motors)

    Science.gov (United States)

    Lingbloom, Mike; Plaia, Jim; Newman, John

    2006-01-01

    Laser Shearography is a viable inspection method for detection of de-bonds and voids within the external TPS (thermal protection system) on to the Space Shuttle RSRM (reusable solid rocket motors). Cork samples with thicknesses up to 1 inch were tested at the LTI (Laser Technology Incorporated) laboratory using vacuum-applied stress in a vacuum chamber. The testing proved that the technology could detect cork to steel un-bonds using vacuum stress techniques in the laboratory environment. The next logical step was to inspect the TPS on a RSRM. Although detailed post flight inspection has confirmed that ATK Thiokol's cork bonding technique provides a reliable cork to case bond, due to the Space Shuttle Columbia incident there is a great interest in verifying bond-lines on the external TPS. This interest provided and opportunity to inspect a RSRM motor with Laser Shearography. This paper will describe the laboratory testing and RSRM testing that has been performed to date. Descriptions of the test equipment setup and techniques for data collection and detailed results will be given. The data from the test show that Laser Shearography is an effective technology and readily adaptable to inspect a RSRM.

  6. In-Space Repair and Refurbishment of Thermal Protection System Structures for Reusable Launch Vehicles

    Science.gov (United States)

    Singh, M.

    2007-01-01

    Advanced repair and refurbishment technologies are critically needed for the thermal protection system of current space transportation systems as well as for future launch and crew return vehicles. There is a history of damage to these systems from impact during ground handling or ice during launch. In addition, there exists the potential for in-orbit damage from micrometeoroid and orbital debris impact as well as different factors (weather, launch acoustics, shearing, etc.) during launch and re-entry. The GRC developed GRABER (Glenn Refractory Adhesive for Bonding and Exterior Repair) material has shown multiuse capability for repair of small cracks and damage in reinforced carbon-carbon (RCC) material. The concept consists of preparing an adhesive paste of desired ceramic with appropriate additives and then applying the paste to the damaged/cracked area of the RCC composites with an adhesive delivery system. The adhesive paste cures at 100-120 C and transforms into a high temperature ceramic during reentry conditions. A number of plasma torch and ArcJet tests were carried out to evaluate the crack repair capability of GRABER materials for Reinforced Carbon-Carbon (RCC) composites. For the large area repair applications, Integrated Systems for Tile and Leading Edge Repair (InSTALER) have been developed and evaluated under various ArcJet testing conditions. In this presentation, performance of the repair materials as applied to RCC is discussed. Additionally, critical in-space repair needs and technical challenges are reviewed.

  7. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1994-11-01

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

  8. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

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

    1987-11-01

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

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

    International Nuclear Information System (INIS)

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

    1988-03-01

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

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

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Rampal, G.; Boccaccini, L.V.; Meyder, R.; Neuberger, H.; Laesser, R.; Poitevin, Y.; Zmitko, M.; Rigal, E.

    2006-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  12. LMFBR blanket physics project progress report No. 4

    International Nuclear Information System (INIS)

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

    1973-01-01

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

  13. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.

    1981-01-01

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

  14. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

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

  15. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1988-01-01

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

  17. Tritium inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Reiter, F.

    1990-01-01

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

  18. Phase change material thermal capacitor clothing

    Science.gov (United States)

    Buckley, Theresa M. (Inventor)

    2005-01-01

    An apparatus and method for metabolic cooling and insulation of a user in a cold environment. In its preferred embodiment the apparatus is a highly flexible composite material having a flexible matrix containing a phase change thermal storage material. The apparatus can be made to heat or cool the body or to act as a thermal buffer to protect the wearer from changing environmental conditions. The apparatus may also include an external thermal insulation layer and/or an internal thermal control layer to regulate the rate of heat exchange between the composite and the skin of the wearer. Other embodiments of the apparatus also provide 1) a path for evaporation or direct absorption of perspiration from the skin of the wearer for improved comfort and thermal control, 2) heat conductive pathways within the material for thermal equalization, 3) surface treatments for improved absorption or rejection of heat by the material, and 4) means for quickly regenerating the thermal storage capacity for reuse of the material. Applications of the composite materials are also described which take advantage of the composite's thermal characteristics. The examples described include a diver's wet suit, ski boot liners, thermal socks, gloves and a face mask for cold weather activities, and a metabolic heating or cooling blanket useful for treating hypothermia or fever patients in a medical setting and therapeutic heating or cooling orthopedic joint supports.

  19. Feasibility study of LiF-BeF2 and chloride salts as blanket coolants for fusion power reactors

    International Nuclear Information System (INIS)

    Imamura, Y.

    1977-09-01

    The feasibility of using molten salts, in particular, nonberyllium-bearing chloride salts, as blanket coolants for Tokamak fusion reactors has been examined for the nucleonic and thermal/hydraulic aspects. It is concluded that the chloride salts, i.e., LiCl--KCl, LiCl--PbCl 2 and LiCl--SnCl 2 , can be used as the blanket coolant for a static lithium metal blanket provided that large blanket thickness can be tolerated, along with the use of U-238 for neutron multiplication in the cases of LiCl--KCl or LiCl--SnCl 2 cooled blankets. However, to make the appraisal complete, the tritium recovery and corrosion problems must be examined extensively, based on data not yet at hand. As for LiF--BeF 2 , it is observed that although the salt mixture can be used for a single fluid blanket with satisfactory nuclear performance, careful attention should be paid to the cooling capability

  20. Conformal Ablative Thermal Protection System for Planetary and Human Exploration Missions

    Science.gov (United States)

    Beck, R.; Arnold, J.; Gasch, M.; Stackpole, M.; Wercinski, R.; Venkatapathy, E.; Fan, W.; Thornton, J; Szalai, C.

    2012-01-01

    The Office of Chief Technologist (OCT), NASA has identified the need for research and technology development in part from NASAs Strategic Goal 3.3 of the NASA Strategic Plan to develop and demonstrate the critical technologies that will make NASAs exploration, science, and discovery missions more affordable and more capable. Furthermore, the Game Changing Development Program (GCDP) is a primary avenue to achieve the Agencys 2011 strategic goal to Create the innovative new space technologies for our exploration, science, and economic future. In addition, recently released NASA Space Technology Roadmaps and Priorities, by the National Research Council (NRC) of the National Academy of Sciences stresses the need for NASA to invest in the very near term in specific EDL technologies. The report points out the following challenges (Page 2-38 of the pre-publication copy released on February 1, 2012): Mass to Surface: Develop the ability to deliver more payload to the destination. NASA's future missions will require ever-greater mass delivery capability in order to place scientifically significant instrument packages on distant bodies of interest, to facilitate sample returns from bodies of interest, and to enable human exploration of planets such as Mars. As the maximum mass that can be delivered to an entry interface is fixed for a given launch system and trajectory design, the mass delivered to the surface will require reductions in spacecraft structural mass more efficient, lighter thermal protection systems more efficient lighter propulsion systems and lighter, more efficient deceleration systems. Surface Access: Increase the ability to land at a variety of planetary locales and at a variety of times. Access to specific sites can be achieved via landing at a specific location(s) or transit from a single designated landing location, but it is currently infeasible to transit long distances and through extremely rugged terrain, requiring landing close to the site of

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

    Science.gov (United States)

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

    2006-02-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

    This paper presents the significant progress made in the research and development (R and D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li 2 TiO 3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 0 C followed by normalizing it at 930 0 C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R and D on the breeder material, Li 2 TiO 3 , the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li 2 TiO 3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li 2 TiO 3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation

  3. I/O Standard Based Thermal/Energy Efficient Green Communication For Wi-Fi Protected Access on FPGA

    DEFF Research Database (Denmark)

    Kumar, Tanesh; Pandey, Bishwajeet; Das, Teerath

    2014-01-01

    In this paper, we analyzed how does life and reliability of an integrated circuit is affected when it is operated in different regions under different temperatures. We have taken Fibonacci generator as our target circuit and LVCMOS as I/O standards. WPA and WPA2 (Wi-Fi Protected Access) key can...... be generated with Fibonacci generator. Here, thermal efficient green Fibonacci Generator is used to generate key for Wi-Fi Protected Access in order to make green communication possible under different room temperature. By analysis it is observed that at standard normal temperature (21degrees C), LVCMOS12 have...

  4. Fusion energy for alternate applications: the design of a high temperature falling bed as a long-lived blanket

    International Nuclear Information System (INIS)

    Harkness, S.D.; Stevens, H.C.; Hall, M.M.; Gohar, M.Y.A.; de Paz, J.F.

    1979-01-01

    The high temperature falling bed conceptual design work has consisted of a coordinated effort in neutronics, materials science, thermal hydraulics and mechanical design. The neutronics work has been based on a one-dimensional transport analysis and has been used to scope the implication of blanket dimensions, breeding materials, ceramic pebble material and coolant choice on both tritium breeding capabilities and energy deposition into the high temperature region of the blanket. The materials science effort has concentrated on defining the selection of a particular ceramic material. The thermal hydraulic analysis has been concerned with sizing the heat transfer system and defining the temperature gradients in the high temperature blanket. The mechanical design work has evaluated how such a system might be constructed from the point of view of maintainability and structural support

  5. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  6. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

    International Nuclear Information System (INIS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

    2006-10-01

    Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

  8. Heat Transfer and Failure Mode Analyses of Ultrahigh-Temperature Ceramic Thermal Protection System of Hypersonic Vehicles

    Directory of Open Access Journals (Sweden)

    Tianbao Cheng

    2014-01-01

    Full Text Available The transient temperature distribution of the ultrahigh-temperature ceramic (UHTC thermal protection system (TPS of hypersonic vehicles is calculated using finite volume method. Convective cooling enables a balance of heat increment and loss to be achieved. The temperature in the UHTC plate at the balance is approximately proportional to the surface heat flux and is approximately inversely proportional to the convective heat transfer coefficient. The failure modes of the UHTCs are presented by investigating the thermal stress field of the UHTC TPS under different thermal environments. The UHTCs which act as the thermal protection materials of hypersonic vehicles can fail because of the tensile stress at the lower surface, an area above the middle plane, and the upper surface as well as because of the compressive stress at the upper surface. However, the area between the lower surface and the middle plane and a small area near the upper surface are relatively safe. Neither the compressive stress nor the tensile stress will cause failure of these areas.

  9. Detection of impact damage on thermal protection systems using thin-film piezoelectric sensors for integrated structural health monitoring

    Science.gov (United States)

    Na, Jeong K.; Kuhr, Samuel J.; Jata, Kumar V.

    2008-03-01

    Thermal Protection Systems (TPS) can be subjected to impact damage during flight and/or during ground maintenance and/or repair. AFRL/RXLP is developing a reliable and robust on-board sensing/monitoring capability for next generation thermal protection systems to detect and assess impact damage. This study was focused on two classes of metallic thermal protection tiles to determine threshold for impact damage and develop sensing capability of the impacts. Sensors made of PVDF piezoelectric film were employed and tested to evaluate the detectability of impact signals and assess the onset or threshold of impact damage. Testing was performed over a range of impact energy levels, where the sensors were adhered to the back of the specimens. The PVDF signal levels were analyzed and compared to assess damage, where digital microscopy, visual inspection, and white light interferometry were used for damage verification. Based on the impact test results, an assessment of the impact damage thresholds for each type of metallic TPS system was made.

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

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Miki, Nobuharu; Akiba, Masato

    2003-06-01

    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 Li 2 TiO 3 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 Li 2 TiO 3 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 328

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

    International Nuclear Information System (INIS)

    Daigo Tsuru; Mikio Enoeda; Masato Akiba

    2006-01-01

    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

  12. Mechanical design and analysis for a EPR first wall/blanket/shield system

    International Nuclear Information System (INIS)

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1978-01-01

    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 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

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

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

    International Nuclear Information System (INIS)

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

    1995-05-01

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

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

    International Nuclear Information System (INIS)

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

    1980-09-01

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

  16. Thermal protection of electronic devices with the Nylon6/66-PEG nanofiber membranes

    OpenAIRE

    Li Ya; Li Xue-Weis; He Ji-Huan; Wang Ping

    2014-01-01

    Phase change materials for thermal energy storage have been widely applied to clothing insulation, electronic products of heat energy storage. The thermal storage potential of the nanofiber membranes was analyzed using the differential scanning calorimetry. Effect of microstructure of the membrane on energy storage was analyzed, and its applications to electronic devices were elucidated.

  17. Protection and thermal management of thermoelectric generator system using phase change materials: An experimental investigation

    DEFF Research Database (Denmark)

    Ahmadi Atouei, Saeed; Rezaniakolaei, Alireza; Ranjbar, A.A.

    2018-01-01

    In most thermoelectric systems the thermal boundary conditions are transient, and thermal manage-ment of the system is critical to improve electrical performance of the system. In this study, effect of using phase change materials (PCM) to control the hot and cold side temperatures...

  18. Thermal design of AOTV heatshields for a conical drag brake

    Science.gov (United States)

    Pitts, W. C.; Murbach, M. S.

    1985-01-01

    Results are presented from an on-going study of the thermal performance of thermal protection systems for a conical drag brake type AOTV. Three types of heatshield are considered: rigid ceramic insulation, flexible ceramic blankets, and ceramic cloths. The results for the rigid insulation apply to other types of AOTV as well. Charts are presented in parametric form so that they may be applied to a variety of missions and vehicle configurations. The parameters considered include: braking maneuver heat flux and total heat load, heatshield material and thickness, heatshield thermal mass and conductivity, absorptivity and emissivity of surfaces, thermal mass of support structure, and radiation transmission through thin heatshields. Results of temperature calculations presented show trends with and sensitivities to these parameters. The emphasis is on providing information that will be useful in estimating the minimum required mass of these heatshield materials.

  19. Analysis of risk and dose when using thermal protection on non-fissile and fissile-excepted UF{sub 6} 48-inch cylinder packages

    Energy Technology Data Exchange (ETDEWEB)

    Chambers, D.B.; Lowe, L.M. [SENES Consultants Ltd., Richmond Hill, ON (Canada); Elizabeth Darrough, M.; Jones, R.H.

    2004-07-01

    An industry consortium of owners of large (i.e., the 48-inch or 48X and 48Y) cylinders commissioned an independent study to evaluate the safety of using thermal protective covers on the cylinders and the likelihood that the cylinders would experience the regulations' hypothetical thermal accident. The study examined the demonstrable risks of the protective covers, i.e., increased dose to workers and the potential for accidents associated with the extra handling, vs. the theoretical risk of the UF{sub 6} cylinders' encountering the hypothetical fire, to evaluate the appropriateness of using the thermal protective covers.

  20. Analysis of risk and dose when using thermal protection on non-fissile and fissile-excepted UF6 48-inch cylinder packages

    International Nuclear Information System (INIS)

    Chambers, D.B.; Lowe, L.M.; Elizabeth Darrough, M.; Jones, R.H.

    2004-01-01

    An industry consortium of owners of large (i.e., the 48-inch or 48X and 48Y) cylinders commissioned an independent study to evaluate the safety of using thermal protective covers on the cylinders and the likelihood that the cylinders would experience the regulations' hypothetical thermal accident. The study examined the demonstrable risks of the protective covers, i.e., increased dose to workers and the potential for accidents associated with the extra handling, vs. the theoretical risk of the UF 6 cylinders' encountering the hypothetical fire, to evaluate the appropriateness of using the thermal protective covers

  1. Using one hybrid 3D-1D-3D approach for the conceptual design of WCCB blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Li, Jia [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2017-01-15

    Highlights: • The Hybrid 3D-1D-3D approach is used for radial building design of WCCB. • Nuclear heat obtained by this method agrees well with 3D neutronics results. • The final results of temperature and TBR satisfy with the requirements. • All the results show that this approach is high efficiency and high reliability. - Abstract: A hybrid 3D-1D-3D approach is proposed for the conceptual design of a blanket. Firstly, the neutron wall loading (NWL) of each blanket module is obtained through a neutronics calculation employing a 3D model, which contains the geometry outline of in-vacuum vessel components and the exact neutron source distribution. Secondly, a 1D cylindrical model with the blanket module containing a detailed radial building is adopted for the neutronics analysis, with the aim of calculating the tritium breeding ratio (TBR) and nuclear heating. Being normalized to the NWL, the nuclear heating is transferred to a 2D model for thermal-hydraulics analysis using the FLUENT code. Through a series analysis of nuclear-thermal iterations that considers the tritium breeding ratio (TBR) and thermal performance as optimization objectives, the optimized radial building of each module surrounding plasma can be obtained. Thirdly, the 3D structural design of each module is established by adding side walls, cover plates, stiffening plates, and other components based on the radial building. The 3D neutronics and thermal-hydraulics using the detailed blanket modules are re-analyzed. This approach has been successfully applied to the design of a water-cooled ceramic breeder blanket for the Chinese Fusion Engineering Test Reactor (CFETR). The radial building of each blanket module surrounding plasma is optimized. The global tritium breeding ratio (TBR) calculated by the 3D neutronics analysis is 1.21, and the temperature of all materials in the 3D blanket structure is below the upper limits. As indicated by the comparison of the 1D and 3D neutronics and thermal

  2. Fusion technology development: first wall/blanket system and component testing in existing nuclear facilities

    International Nuclear Information System (INIS)

    Hsu, P.Y.S.; Bohn, T.S.; Deis, G.A.; Judd, J.L.; Longhurst, G.R.; Miller, L.G.; Millsap, D.A.; Scott, A.J.; Wessol, D.E.

    1980-12-01

    A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment employing an existing nuclear facility, the Engineering Test Reactor at the Idaho National Engineering Laboratory, is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of an existing test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module. The principal reaction considered involves 3 He in the annulus as follows: n + 3 He → p + t + 0.75 MeV. Bulk heating in the test module is accomplished by neutron thermalization, gamma heating, and absorption reactions involving 6 Li in the blanket breeding region. The concept can be extended to modified core configurations that will accommodate test modules of different sizes and types. It makes possible development testing of first wall/blanket systems and other fusion components on a scale and in ways not otherwise available until actual high-power fusion reactors are built

  3. Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

    International Nuclear Information System (INIS)

    Barrera, G.; Branas, B.; Lucas, J.; Doncel, J.; Medrano, M.; Garcia, A.; Giancarli, L.; Ibarra, A.; Li Puma, A.; Maisonnier, D.; Sardain, P.

    2008-01-01

    The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 deg. C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 deg. C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions

  4. Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Malang, S.; Sawan, M.

    2005-01-01

    As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li 2 BeF 4 and the low melting point molten salts such as LiBeF 3 and LiNaBeF 4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC f /SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R and D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan

  5. Study of materials used for the thermal protection of the intake system for internal combustion engines

    Science.gov (United States)

    Birtok-Băneasă, C.; Raţiu, S.; Puţan, V.; Josan, A.

    2018-01-01

    The present paper focuses on calculation of thermal conductivity for a new materials developed by the authors, using the heat flux plate method. This experimental method consists in placing the sample of the new material in a calorimetric chamber and heating from underside. As the heat flux which passes through the sample material is constant and knowing the values of the temperatures for the both sides of sample, the sample material thermal conductivity is determined. Six types of different materials were tested. Based on the experimental data, the values of the thermal conductivity according to the material and the average temperature were calculated and plotted.

  6. On protection and operation of home consumption of large thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Pospisil, J.

    1988-01-01

    The operating reliability of the system of home consumption is affected not only by the operating reliability of the equipment but also by the system of home consumption protection. The existing system of protection in Czechoslovakia and elsewhere and new concepts in the GDR and in Poland are described. In order to improve protection redundancy in the home consumption system, distance protection can be used. The reliability of the home consumption power supply system can be improved by the use of low-ohm grounding of the home consumption transformer or by an additional transformer and a protection system with time-graded overcurrent protection devices to the zero component of the current. (M.D.). 2 figs., 8 refs

  7. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

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

  8. Experimental Studies on Shadow Shields for Thermal Protection of Cryogenic Tanks in Space

    National Research Council Canada - National Science Library

    Knoll, Richard

    1968-01-01

    ... (high-emissivity coatings on annular rings of shields) on thermal performance. The experimental data, in general, agreed closely with an analytical model which assumed diffuse surfaces with nonuniform radiosity...

  9. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  10. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

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

  11. Flow balancing in liquid metal blankets

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  12. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

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

  13. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  14. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  15. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Waganer, L.M.

    1985-01-01

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

  16. Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

    2017-05-15

    Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

  17. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

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

  18. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

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

  19. Helium Loop for the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

    2006-01-01

    In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

  20. Status of EC solid breeder blanket designs and R and D for demo fusion reactors

    International Nuclear Information System (INIS)

    Proust, E.; Anzidei, L.; Moons, F.

    1994-01-01

    Within the European Community Fusion Technology Program two solid breeder blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium purge gas, the martensitic steel MANET as structural material and beryllium as neutron multiplier. The configurations of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder materials are LiAlO 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li 4 SiO 4 and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behavior of the breeder ceramics and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation induced embrittlement of MANET, the mechanical effects caused by major plasma disruptions, and safety and reliability. The R and D work concentrate on these issues. The development of martensitic steels including MANET is part of a separate program. Breeder ceramics and beryllium irradiations have been so far performed for conditions which do not cover the peak values injected in the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. An effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling systems. First promising results have been obtained also in field of mechanical effects from plasma disruptions and safety and reliability, however further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects. (authors). 8 figs., 2 tabs., 53 refs

  1. Status of ITER blanket attachment design and related R and D

    International Nuclear Information System (INIS)

    Sadakov, S.; Khomiakov, S.; Calcagno, B.; Chappuis, Ph.; Dellopoulos, G.; Kolganov, V.; Merola, M.; Poddubnyi, I.; Raffray, R.; Raharijaona, J.J.; Ulrickson, M.; Zhmakin, A.

    2013-01-01

    Highlights: • ITER blanket attachment system went through a significant design upgrade and become basically compliant with specified design loads and required cyclic lifetime. • Upgrade of flexible supports allowed the doubling of cross sections of central bolts. Ceramic coatings were relocated to much larger areas on conical pairs screwed into shield blocks. • Key pads were relocated from keys of vacuum vessel into keyways of shield blocks and reshaped to enlarge areas of lateral interfaces with ceramic electro-insulating coatings. • Ceramic coatings are hidden between pads and enclosures in keyways with a purpose to minimize their wear rate, which depends on peak friction stress and cyclic sliding path. • Ceramic coatings to be verified by experiment, with several R and D aimed to collect statistically sufficient data on their reliability and durability in ITER relevant cyclic loading conditions. -- Abstract: Main function of the ITER blanket system [1–3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R and D is also briefly described

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

    International Nuclear Information System (INIS)

    Iida, H.; Minato, A.; Kitamura, K.

    1983-01-01

    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 (Li 2 O) 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)

  3. Phase change of First Wall in Water-Cooled Breeding Blankets of K-DEMO for Vertical

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon Woo; Lee, Jeong Hun; Cho, Hyoung Kyu; Park, Goon Cherl [Seoul National University, Seoul (Korea, Republic of); Im, Ki Hak [NFRI, Daejeon (Korea, Republic of)

    2016-05-15

    The purpose of this study is to simulate thermal-hydraulic behavior of a single blanket module when plasma disruption occurs. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. The thermal design, evaluation and validation have been performed in order to establish the conceptual design guidelines of the water-cooled breeding blanket for the K-DEMO reactor. As a part of the NFRI research, Seoul National University (SNU) is conducting transient thermal-hydraulic analysis to confirm the integrity of blanket system for plasma disruption events. Vertical displacement events (VDE) with high heat flux can cause melting and vaporization of plasma facing materials (PFCs) and also burnout of coolant channels. In order to simulate melting of first wall in blanket module when VDE occurs, one-dimensional heat conduction equations were solved numerically with modification of the specific heat of the first wall materials using effective heat capacity method. Temperature profiles in first wall for VDE are shown in fig 7 - 9. At first, temperature of tungsten rapidly raised and even exceeded its melting temperature. When VDE just ended at 0.1 second, 0.83 mm thick of tungsten melted. But the other materials including vanadium and RAFM didn't exceed their melting temperatures after 500 seconds.

  4. Phase change of First Wall in Water-Cooled Breeding Blankets of K-DEMO for Vertical

    International Nuclear Information System (INIS)

    Kim, Geon Woo; Lee, Jeong Hun; Cho, Hyoung Kyu; Park, Goon Cherl; Im, Ki Hak

    2016-01-01

    The purpose of this study is to simulate thermal-hydraulic behavior of a single blanket module when plasma disruption occurs. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. The thermal design, evaluation and validation have been performed in order to establish the conceptual design guidelines of the water-cooled breeding blanket for the K-DEMO reactor. As a part of the NFRI research, Seoul National University (SNU) is conducting transient thermal-hydraulic analysis to confirm the integrity of blanket system for plasma disruption events. Vertical displacement events (VDE) with high heat flux can cause melting and vaporization of plasma facing materials (PFCs) and also burnout of coolant channels. In order to simulate melting of first wall in blanket module when VDE occurs, one-dimensional heat conduction equations were solved numerically with modification of the specific heat of the first wall materials using effective heat capacity method. Temperature profiles in first wall for VDE are shown in fig 7 - 9. At first, temperature of tungsten rapidly raised and even exceeded its melting temperature. When VDE just ended at 0.1 second, 0.83 mm thick of tungsten melted. But the other materials including vanadium and RAFM didn't exceed their melting temperatures after 500 seconds

  5. Minimally Intrusive Embedded Sensors for High Mass to Mars Thermal Protection Systems

    Data.gov (United States)

    National Aeronautics and Space Administration — The safety and overall mission success of space vehicles rely heavily on the integrity of the TPS to protect valuable payload during atmospheric entry. Future...

  6. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

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

  7. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1980-01-01

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

  8. Cross-section uncertainty study of the NET shielding blanket

    International Nuclear Information System (INIS)

    Jaeger, J.F.

    1990-11-01

    The Next European Torus (NET) is foreseen as the next step in the European development towards the controlled use of thermonuclear fusion. Detail design of the shielding blanket protecting the peripherals, more especially the super-conducting coils, is well advanced. A cross-section uncertainty, i.e. a study of the expected inaccuracy due to the nuclear cross-section data, has been done for the neutron-gamma reactions in the insulation of the coils for such a design. As an extension of previous work on the NET shielding blanket (e.g. MCNP calculations), it was deemed necessary to estimate the accuracy attainable with transport codes in view of the uncertainties in microscopic cross-sections. The code used, SENSIBL, is based on perturbation theory and uses covariance files, COVFILS-2, for the cross-section data. This necessitates forward and adjoint flux calculations with a transport code (e.g. ONEDANT, TRISM) and folding the information contained in these coupled fluxes with the accuracy estimates of the evaluators of the ENDF/B-V files. Transport, P 5 S 12 , calculations were done with the ONEDANT code, for a shielding blanket design with 714 MW plasma fusion power. Several runs were done to obtain well converged forward and adjoint fluxes (ca. 1%). The forward and adjoint integral responses agree to 2%, which is consistent with the above accuracy. The n-γ response was chosen as it is typical of the general accuracy and is available for all materials considered. The present version of SENSIBL allows direct use of the geometric files of ONEDANT (or TRISM) which simplifies the input. Covariance data is not available at present in COVFILS-2 for all of the materials considered. Only H, C, N, O, Al, Si, Fe, Ni, and Pb could be considered, the big absentee being copper. The resulting uncertainty for the neutron-gamma reactions in the insulation of the coil was found to be 17%. Simulating copper by aluminium produces a negligible increase in the uncertainty, mainly

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

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  11. Design of the breeder units in the new HCPB modular blanket concept and material requirements

    International Nuclear Information System (INIS)

    Boccaccini, L.V.; Fischer, U.; Hermsmeyer, S.; Reimann, J.; Xu, Z.; Koehly, C.

    2004-01-01

    ; according to the experience from the old design no major problems are expected concerning stress levels and gap formation in the pebble beds; in fact, the situation should be more favourable due to the reduced dimensions (max. 20 cm) of the beds that should minimise ratcheting and particle flow phenomena. In respect to tritium extraction, the most favourable features of the HCPB concept (e.g. an overall low partial pressure of tritium in the beds that minimise permeation into the main coolant system) can be kept in the new design; a complication could be the necessity to provide each cell with a system of tubes to inlet the purge helium in the front part of the beds or to divide the purge flow for Be and CB. The modular design of the new HCPB blanket, that makes the breeder units almost independent on the structural design of the box, opens interesting possibilities in the development of these units. The present design can be optimised on the basis of the results of the present R and D programme on Be and CB. In addition, new requirements could appear to improve the design performances or manufacturing: i.e. in respect to filling procedures of the beds (use of pre-packed breeder units) or the necessity of insulating layer to thermally decouple the breeder units from the box or the selection of new materials with a better compatibility with Be and CB at high temperatures as protection of the steel structure. (author)

  12. Mechanisms Underpinning Degradation of Protective Oxides and Thermal Barrier Coatings in High Hydrogen Content (HHC) - Fueled Turbines

    Energy Technology Data Exchange (ETDEWEB)

    Mumm, Daniel

    2013-08-31

    The overarching goal of this research program has been to evaluate the potential impacts of coal-derived syngas and high-hydrogen content fuels on the degradation of turbine hot-section components through attack of protective oxides and thermal barrier coatings. The primary focus of this research program has been to explore mechanisms underpinning the observed degradation processes, and connections to the combustion environments and characteristic non-combustible constituents. Based on the mechanistic understanding of how these emerging fuel streams affect materials degradation, the ultimate goal of the program is to advance the goals of the Advanced Turbine Program by developing materials design protocols leading to turbine hot-section components with improved resistance to service lifetime degradation under advanced fuels exposures. This research program has been focused on studying how: (1) differing combustion environments – relative to traditional natural gas fired systems – affect both the growth rate of thermally grown oxide (TGO) layers and the stability of these oxides and of protective thermal barrier coatings (TBCs); and (2) how low levels of fuel impurities and characteristic non-combustibles interact with surface oxides, for instance through the development of molten deposits that lead to hot corrosion of protective TBC coatings. The overall program has been comprised of six inter-related themes, each comprising a research thrust over the program period, including: (i) evaluating the role of syngas and high hydrogen content (HHC) combustion environments in modifying component surface temperatures, heat transfer to the TBC coatings, and thermal gradients within these coatings; (ii) understanding the instability of TBC coatings in the syngas and high hydrogen environment with regards to decomposition, phase changes and sintering; (iii) characterizing ash deposition, molten phase development and infiltration, and associated corrosive

  13. Electromagnetic analysis of ITER shield blanket under VDE

    International Nuclear Information System (INIS)

    Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

    2010-01-01

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

  14. Minimum thickness blanket-shield for fusion reactors

    International Nuclear Information System (INIS)

    Karni, Y.; Greenspan, E.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.; Kernforschungszentrum Karlsruhe GmbH

    1994-10-01

    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.) [de

  17. Influence of start up and pulsed operation on tritium release and inventory of NET ceramic blanket

    International Nuclear Information System (INIS)

    Iseli, M.; Esser, B.

    1989-01-01

    A first estimate for the tritium release behaviour of a ceramic breeder blanket in pulsed operation is obtained by assuming a linear steady state temperature distribution and taking into account the time constant of the thermal behaviour. The release behaviour of the breeder exposed to consecutive periods of tritium generation is described with an analytical solution of the diffusion equation. The results are compared with a simple exponential approach valid for surfacte desorption controlled release. The exponential model is used to simulate a blanket with aluminate as breeder material, which takes longest to reach steady state. The simulation demonstrates that a significant fraction (>67%) of steady state can be achieved after a testing time of about one day. (author). 7 refs.; 8 figs.; 3 tabs

  18. Qualification of MHD effects in dual-coolant DEMO blanket and approaches to their modelling

    International Nuclear Information System (INIS)

    Mas de les Valls, E.; Batet, L.; Medina, V. de; Fradera, J.; Sedano, L.A.

    2011-01-01

    Design refinements of vertical insulated banana-shaped liquid metal channels are being considered as a progress of conceptual design of dual-coolant liquid metal blankets (DEMO specifications). Among them: (a) optimised channel geometry and (b) improvements on flow channel inserts. Progress of channel conceptual design is conducted in parallel with underlying physics of MHD models in diverse aspects: (1) MHD models, (2) MHD turbulence, (3) LM buoyancy effects, (4) three-dimensional flows, and (5) LM/FCI/wall electrical and thermal coupling; in order to progress on common liquid metal flow characterisation, pressure drop and three-dimensional flows. The analyses are assumed as extension of those previous carried out for the DCLL blankets for new design refinements. At the present stage of the conceptual design progress, a preliminary thermofluid MHD study is of crucial interest for further design improvements and future detailed modelling. The paper overviews the ongoing modelling studies, making model refinements explicit, and anticipates some modelling results.

  19. An Aeroelastic Evaluation of the Flexible Thermal Protection System for an Inatable Aerodynamic Decelerator

    Science.gov (United States)

    Goldman, Benjamin D.

    The purpose of this dissertation is to study the aeroelastic stability of a proposed flexible thermal protection system (FTPS) for the NASA Hypersonic Inflatable Aerodynamic Decelerator (HIAD). A flat, square FTPS coupon exhibits violent oscillations during experimental aerothermal testing in NASA's 8 Foot High Temperature Tunnel, leading to catastrophic failure. The behavior of the structural response suggested that aeroelastic flutter may be the primary instability mechanism, prompting further experimental investigation and theoretical model development. Using Von Karman's plate theory for the panel-like structure and piston theory aerodynamics, a set of aeroelastic models were developed and limit cycle oscillations (LCOs) were calculated at the tunnel flow conditions. Similarities in frequency content of the theoretical and experimental responses indicated that the observed FTPS oscillations were likely aeroelastic in nature, specifically LCO/flutter. While the coupon models can be used for comparison with tunnel tests, they cannot predict accurately the aeroelastic behavior of the FTPS in atmospheric flight. This is because the geometry of the flight vehicle is no longer a flat plate, but rather (approximately) a conical shell. In the second phase of this work, linearized Donnell conical shell theory and piston theory aerodynamics are used to calculate natural modes of vibration and flutter dynamic pressures for various structural models composed of one or more conical shells resting on several circumferential elastic supports. When the flight vehicle is approximated as a single conical shell without elastic supports, asymmetric flutter in many circumferential waves is observed. When the elastic supports are included, the shell flutters symmetrically in zero circumferential waves. Structural damping is found to be important in this case, as "hump-mode" flutter is possible. Aeroelastic models that consider the individual FTPS layers as separate shells exhibit

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

    International Nuclear Information System (INIS)

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

    1985-01-01

    Radiation-induced infections can be associated with changes in colonization potential of the intestine. Since the mucous blanket, which overlays the epithelium, is a major mucosal structure and is heavily colonized by microorganisms, we examined the status of the mucus after radiation and evaluated susceptibility to intestinal challenge with bacteria. A downward shift (2.5 X 10(8) cells/g to 5.3 X 10(5)) of total facultatively anaerobic bacteria of the ileum of C3HeB/FeJ mice was detected by 3 days post exposure to 10 Gy 60Co. Numbers of flora returned to normal by 11 days after radiation. Scanning electron microscopy was used to show that the loss of bacteria could be associated with major disruptions of the continuity of the mucous blanket. The pathogen Pseudomonas aeruginosa adhered to mouse mucous films used in in vitro assays. When irradiated mice were challenged orally with 1 X 10(5) P. aeruginosa on days 1, 2, or 3 after irradiation, a progressive increase in susceptibility was seen, but no animals died before Day 4 postirradiation. Sensitivity to subcutaneous (sc) challenge with Pseudomonas also increased by Day 3 and was probably due largely to the profound neutropenia observed. Immunoglobulin G (Gamimmune), which protected burned mice infected with Pseudomonas, was ineffectual in treatment of 7 or 10 Gy irradiated mice challenged either orally or sc with the organism. The ileal mucosal barrier was compromised after radiation in ways which could facilitate epithelial colonization, an event which combined with other immunological and physiological decrements in this model can compromise the effectiveness of therapeutic modalities

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  2. High dose neutron irradiation damage in beryllium as blanket material

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V.P. E-mail: fae@niiar.ru; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B. E-mail: vniinm.400@g23.relkom.ru

    2001-11-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10{sup 22} and 8.0x10{sup 22} cm{sup -2} (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10{sup 22} cm{sup -2} (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10{sup 22} cm{sup -2} (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

  3. High dose neutron irradiation damage in beryllium as blanket material

    International Nuclear Information System (INIS)

    Chakin, V.P.; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B.

    2001-01-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10 22 and 8.0x10 22 cm -2 (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10 22 cm -2 (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10 22 cm -2 (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket

  4. Manufacturing Technology of Ceramic Pebbles for Breeding Blanket

    Directory of Open Access Journals (Sweden)

    Rosa Lo Frano

    2018-05-01

    Full Text Available An open issue for the fusion power reactor is the choice of breeding blanket material. The possible use of Helium-Cooled Pebble Breeder ceramic material in the form of pebble beds is of great interest worldwide as demonstrated by the numerous studies and research on this subject. Lithium orthosilicate (Li4SiO4 is a promising breeding material investigated in this present study because the neutron capture of Li-6 allows the production of tritium, 6Li (n, t 4He. Furthermore, lithium orthosilicate has the advantages of low activation characteristics, low thermal expansion coefficient, high thermal conductivity, high density and stability. Even if they are far from the industrial standard, a variety of industrial processes have been proposed for making orthosilicate pebbles with diameters of 0.1–1 mm. However, some manufacturing problems have been observed, such as in the chemical stability (agglomeration phenomena. The aim of this study is to provide a new methodology for the production of pebbles based on the drip casting method, which was jointly developed by the DICI-University of Pisa and Industrie Bitossi. Using this new (and alternative manufacturing technology, in the field of fusion reactors, appropriately sized ceramic pebbles could be produced for use as tritium breeders.

  5. Temperature-base and soma thermal for Zinnia ‘Profusion Cherry’ potted grown in protected environment

    Directory of Open Access Journals (Sweden)

    Charleston Gonçalves

    2016-02-01

    Full Text Available The growing of consumer market demands introduction of new species of flowers and cultivars primarily for production under protected cultivation. The zinnia by the quickness of production can be regarded as an alternative, however demand studies by the lack of information in the literature. We evaluated the duration of different periods, the base temperature and thermal accumulation, expressed as degree-days for the potted zinnia ‘Profusion Cherry’, conducted under protected cultivation for different phenological subperiods. The test was conducted in a greenhouse covered with plastic and closed laterally with shading-net and the duration of subperiods were made to twenty times after sowing. The base temperature was determined by relative development and values-based temperature and thermal time in degree-days (DD. The results for the different phases were, respectively: first open flower-planting: 4.1 °C and GD 838, first open flower - 50% of flowers open: 3.0 °C and 184 GD and 50% of flowers open - senescence: 6.9 °C and 238 GD.

  6. Radiation protection Aspects Using the Thermal Waters from the Felix-1 Mai-Oradea Geothermal Deposit

    International Nuclear Information System (INIS)

    Jurcut, T.; Cosma, C.; Pop, I.

    2001-01-01

    Full text: The geothermal 'Felix-1 Mai-Oradea' deposit is situated in the western part of Romania and it is well known long years ago. The waters of this deposit are used in the medical treatments in the two resorts (Felix and 1 Mai) and for heating and swimming pools in Oradea town. The deposit depth (2500-3000 m) determines a high temperature (66-900 deg. C) of these waters also a mineral content of 200-1400 mg/l, the main components being Ca and Mg. First time, during some years, the thermal water was directly used in the central heating radiators from 'Nufarul' residential district. At present a heating switch installation is utilised. The high radium content of these thermal waters comparatively with Italian or Japanese thermal waters suggested us a study of radium deposition on the inner walls of the pipes also in the inner central heating radiators. Analysing these depositions using a high resolution Ge-Li detector, the radium-226 and small quantities of radium-224 and 223 isotopes were registered. Average radium-226 deposition was 1200 Bq/kg. (author)

  7. Development of polyisocyanurate pour foam formulation for space shuttle external tank thermal protection system

    Science.gov (United States)

    Harvey, James A.; Butler, John M.; Chartoff, Richard P.

    1988-01-01

    Four commercially available polyisocyanurate polyurethane spray-foam insulation formulations are used to coat the external tank of the space shuttle. There are several problems associated with these formulations. For example, some do not perform well as pourable closeout/repair systems. Some do not perform well at cryogenic temperatures (poor adhesion to aluminum at liquid nitrogen temperatures). Their thermal stability at elevated temperatures is not adequate. A major defect in all the systems is the lack of detailed chemical information. The formulations are simply supplied to NASA and Martin Marietta, the primary contractor, as components; Part A (isocyanate) and Part B (poly(s) and additives). Because of the lack of chemical information the performance behavior data for the current system, NASA sought the development of a non-proprietary room temperature curable foam insulation. Requirements for the developed system were that it should exhibit equal or better thermal stability both at elevated and cryogenic temperatures with better adhesion to aluminum as compared to the current system. Several formulations were developed that met these requirements, i.e., thermal stability, good pourability, and good bonding to aluminum.

  8. Effects of alpha-zirconium phosphate on thermal degradation and flame retardancy of transparent intumescent fire protective coating

    Energy Technology Data Exchange (ETDEWEB)

    Xing, Weiyi [State Key Laboratory of Fire Science, University of Science and Technology of China, 96 Jinzai Road, Hefei, Anhui 230026 (China); Zhang, Ping [State Key Laboratory Cultivation Base for Nonmetal Composites and Functional Materials, Southwest University of Science and Technology, 59 Qinglong Road, Mianyang 621010 (China); Song, Lei; Wang, Xin [State Key Laboratory of Fire Science, University of Science and Technology of China, 96 Jinzai Road, Hefei, Anhui 230026 (China); Hu, Yuan, E-mail: yuanhu@ustc.edu.cn [State Key Laboratory of Fire Science, University of Science and Technology of China, 96 Jinzai Road, Hefei, Anhui 230026 (China)

    2014-01-01

    Graphical abstract: - Highlights: • A transparent intumescent fire protective coating was obtained by UV-cured technology. • OZrP could enhance the thermal stability and anti-oxidation of the coating. • OZrP could reduce the combustion properties of the coatings. - Abstract: Organophilic alpha-zirconium phosphate (OZrP) was used to improve the thermal and fire retardant behaviors of the phenyl di(acryloyloxyethyl)phosphate (PDHA)-triglycidyl isocyanurate acrylate (TGICA)-2-phenoxyethyl acrylate (PHEA) (PDHA-TGICA-PHEA) coating. The morphology of nanocomposite coating was characterized by X-ray diffraction (XRD) and transmission electron microscopy (TEM). The effect of OZrP on the flame retardancy, thermal stability, fireproofing time and char formation of the coatings was investigated by microscale combustion calorimeter (MCC), thermogravimetric analysis (TGA), X-ray photoelectron spectroscopy (XPS), laser Raman spectroscopy (LRS) and scanning electric microscope (SEM). The results showed that by adding OZrP, the peak heat release rate and total heat of combustion were significantly reduced. The highest improvement was achieved with 0.5 wt% OZrP. XPS analysis indicated that the performance of anti-oxidation of the coating was improved with the addition of OZrP, and SEM images showed that a good synergistic effect was obtained through a ceramic-like layer produced by OZrP covered on the surface of char.

  9. Effects of alpha-zirconium phosphate on thermal degradation and flame retardancy of transparent intumescent fire protective coating

    International Nuclear Information System (INIS)

    Xing, Weiyi; Zhang, Ping; Song, Lei; Wang, Xin; Hu, Yuan

    2014-01-01

    Graphical abstract: - Highlights: • A transparent intumescent fire protective coating was obtained by UV-cured technology. • OZrP could enhance the thermal stability and anti-oxidation of the coating. • OZrP could reduce the combustion properties of the coatings. - Abstract: Organophilic alpha-zirconium phosphate (OZrP) was used to improve the thermal and fire retardant behaviors of the phenyl di(acryloyloxyethyl)phosphate (PDHA)-triglycidyl isocyanurate acrylate (TGICA)-2-phenoxyethyl acrylate (PHEA) (PDHA-TGICA-PHEA) coating. The morphology of nanocomposite coating was characterized by X-ray diffraction (XRD) and transmission electron microscopy (TEM). The effect of OZrP on the flame retardancy, thermal stability, fireproofing time and char formation of the coatings was investigated by microscale combustion calorimeter (MCC), thermogravimetric analysis (TGA), X-ray photoelectron spectroscopy (XPS), laser Raman spectroscopy (LRS) and scanning electric microscope (SEM). The results showed that by adding OZrP, the peak heat release rate and total heat of combustion were significantly reduced. The highest improvement was achieved with 0.5 wt% OZrP. XPS analysis indicated that the performance of anti-oxidation of the coating was improved with the addition of OZrP, and SEM images showed that a good synergistic effect was obtained through a ceramic-like layer produced by OZrP covered on the surface of char

  10. Thermal Shock Properties of Cladding with SiC{sub f}/SiC Composite Protective Films

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Donghee; Park, Kwangheon [Kyunghee University, Yongin (Korea, Republic of); Kim, Weonju; Park, Jiyeon; Kim, Daejong; Lee, Hyeon Geun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In general, Zr-4 alloy is used for such nuclear fuel cladding. Zr-4 possesses a very small thermal neutron absorption cross-section and has superior corrosion resistance in the normal operating conditions of a nuclear reactor. However, in the case of a critical accident such as a LOCA (loss-of-coolant accident) in the Fukushima disaster, the risk of hydrogen explosion becomes serious. That is, in the case of coolant leakage, a dramatic reaction between the nuclear fuel cladding and steam can cause a heating reaction accompanied by rapid high-temperature oxidation, while creating a huge amount of hydrogen. Hence, the search for an alternative material for nuclear fuel cladding is being actively undertaken. Ceramic-based nuclear fuel cladding is receiving much attention as a means of improving safety. SiC has excellent properties of resistance to high temperature and high exposure and superior mechanical properties, as well as a very small thermal neutron absorption cross-section (0.09 barns), which causes almost no decrease in mechanical strength or volume change following exposure. This experiment examined the thermal shock properties and microstructure of cladding that has SiCf/SiC composite protective film, using polycarbosilane preceramic polymer.

  11. Preliminary electromagnetic analysis of Helium Cooled Solid Blanket for CFETR by MAXWELL

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Cheng; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn

    2016-11-15

    Highlights: • A FEM model of the blanket and magnetic system was built. • Electromagnetic forces and moments of the typical blanket for ferromagnetic and non-ferromagnetic materials were computed and analyzed. • Maxwell forces and Lorentz forces were computed and compared. • Eddy current in the blanket was analyzed under MD condition. - Abstract: A Helium Cooled Solid Blanket (HCSB) for CFETR (Chinese Fusion Engineering Test Reactor) was designed by USTC. The structural and thermal-hydraulic analysis has been carried out, while electromagnetic analysis was not carefully researched. In this paper, a FEM (finite element method) model of the HCSB was developed and electromagnetic forces as well as moments was computed by a FEM software called MAXWELL integrated in ANSYS Workbench. In the geometrical model, flow channels and small connecting parts were neglected because of the extreme complication and the reasonable conservative assumption by neglecting these circumstantial details. As for electromagnetic (EM) analysis, Lorentz forces due to eddy currents caused by main disruption and Maxwell forces due to the magnetization of RAFM steel (i.e. EUROFER97) were computed. Since the unavailability of the details of the plasma in CFETR, when disruptions happen, the condition where a linear current quench of main disruption occurs was assumed. The maximum magnitude of the electromagnetic forces was 356.45 kN and the maximum value of the coupled electromagnetic moments was 1899.40 N m around the radial direction. It is feasible to couple electromagnetic analysis, structural analysis and thermal-hydraulic analysis in the future since MAXWELL has good channels to exchange data between different analytic parts.

  12. Thermal Stability and Oxidation Resistance of Nanocomposite TiC/a-C Protective Coatings

    NARCIS (Netherlands)

    Martinez-Martinez, Diego; Lopez-Cartes, Carlos; Gago, Raul; Fernandez, Asuncion; Carlos Sanchez-Lopez, Juan

    2009-01-01

    Nanocomposite films composed by small crystallites of hard phases embedded in an amorphous lubricant matrix have been extensively studied as protective coatings. These kinds of coatings have often to work in extreme environments, exposed to high temperatures (above 800-900 degrees C), and/or

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

    International Nuclear Information System (INIS)

    Finn, P.A.

    1985-01-01

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

  14. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  15. Structural materials for fusion reactor blanket systems

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  16. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

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

  17. Phase change thermal control materials, method and apparatus

    Science.gov (United States)

    Buckley, Theresa M. (Inventor)

    2001-01-01

    An apparatus and method for metabolic cooling and insulation of a user in a cold environment. In its preferred embodiment the apparatus is a highly flexible composite material having a flexible matrix containing a phase change thermal storage material. The apparatus can be made to heat or cool the body or to act as a thermal buffer to protect the wearer from changing environmental conditions. The apparatus may also include an external thermal insulation layer and/or an internal thermal control layer to regulate the rate of heat exchange between the composite and the skin of the wearer. Other embodiments of the apparatus also provide 1) a path for evaporation or direct absorption of perspiration from the skin of the wearer for improved comfort and thermal control, 2) heat conductive pathways within the material for thermal equalization, 3) surface treatments for improved absorption or rejection of heat by the material, and 4) means for quickly regenerating the thermal storage capacity for reuse of the material. Applications of the composite materials are also described which take advantage of the composite's thermal characteristics. The examples described include a diver's wet suit, ski boot liners, thermal socks, gloves and a face mask for cold weather activities, and a metabolic heating or cooling blanket useful for treating hypothermia or fever patients in a medical setting and therapeutic heating or cooling orthopedic joint supports.

  18. Flexible composite material with phase change thermal storage

    Science.gov (United States)

    Buckley, Theresa M. (Inventor)

    2001-01-01

    A highly flexible composite material having a flexible matrix containing a phase change thermal storage material. The composite material can be made to heat or cool the body or to act as a thermal buffer to protect the wearer from changing environmental conditions. The composite may also include an external thermal insulation layer and/or an internal thermal control layer to regulate the rate of heat exchange between the composite and the skin of the wearer. Other embodiments of the PCM composite also provide 1) a path for evaporation or direct absorption of perspiration from the skin of the wearer for improved comfort and thermal control, 2) heat conductive pathways within the material for thermal equalization, 3) surface treatments for improved absorption or rejection of heat by the material, and 4) means for quickly regenerating the thermal storage capacity for reuse of the material. Applications of the composite materials are also described which take advantage of the composite's thermal characteristics. The examples described include a diver's wet suit, ski boot liners, thermal socks, ,gloves and a face mask for cold weather activities, and a metabolic heating or cooling blanket useful for treating hypothermia or fever patients in a medical setting and therapeutic heating or cooling orthopedic joint supports.

  19. Thermal protection for hypervelocity flight in earth's atmosphere by use of radiation backscattering ablating materials

    Science.gov (United States)

    Howe, John T.; Yang, Lily

    1991-01-01

    A heat-shield-material response code predicting the transient performance of a material subject to the combined convective and radiative heating associated with the hypervelocity flight is developed. The code is dynamically interactive to the heating from a transient flow field, including the effects of material ablation on flow field behavior. It accomodates finite time variable material thickness, internal material phase change, wavelength-dependent radiative properties, and temperature-dependent thermal, physical, and radiative properties. The equations of radiative transfer are solved with the material and are coupled to the transfer energy equation containing the radiative flux divergence in addition to the usual energy terms.

  20. Methods and systems to thermally protect fuel nozzles in combustion systems

    Science.gov (United States)

    Helmick, David Andrew; Johnson, Thomas Edward; York, William David; Lacy, Benjamin Paul

    2013-12-17

    A method of assembling a gas turbine engine is provided. The method includes coupling a combustor in flow communication with a compressor such that the combustor receives at least some of the air discharged by the compressor. A fuel nozzle assembly is coupled to the combustor and includes at least one fuel nozzle that includes a plurality of interior surfaces, wherein a thermal barrier coating is applied across at least one of the plurality of interior surfaces to facilitate shielding the interior surfaces from combustion gases.

  1. Conformal Ablative Thermal Protection Systems (CA-TPS) for Venus and Saturn Backshells

    Science.gov (United States)

    Beck, R.; Gasch, M.; Stackpoole, M.; Wilder, M.; Boghozian, T.; Chavez-Garcia, J.; Prabhu, Dinesh; Kazemba, Cole D.; Venkatapathy, E.

    2016-01-01

    This poster provides an overview of the work performed to date on the Conformal Ablative TPS (CA-TPS) element of the TPSM project out of GCDP. Under this element, NASA is developing improved ablative TPS materials based on flexible felt for reinforcement rather than rigid reinforcements. By replacing the reinforcements with felt, the resulting materials have much higher strain-to-failure and are much lower in thermal conductivity than their rigid counterparts. These characteristics should allow for larger tile sizes, direct bonding to aeroshells and even lower weight TPS. The conformal phenolic impregnated carbon felt (C-PICA) is a candidate for backshell TPS for both Venus and Saturn entry vehicles.

  2. Behaviour of glass and thermal protective coatings on stainless steels in the nitrogen tetroxide based coolant

    International Nuclear Information System (INIS)

    Bakalin, Yu.I.; Dobrunova, V.M.; Doroshkevich, V.N.; Nesterenko, V.B.; Trubnikov, V.P.

    1985-01-01

    The technology of application of glass and enamel protective coatings on stainless steel has been examined, their testing in the medium of nitrogen tetroxide based coolant with different content of nitric acid has been carried out, the basic characteristics of the coatings after testing have been defined. Chromium-nickel austenitic 12kh18n10t steel, widely used in the nuclear power, have been chosen as a basic object of examination. The coatings have been tested in nitrogen oxide at P=12.0 MPa, temperature 310 deg C and 0.1% HNO 3 , and also in the medium of vat residue of the rectifying tower with nitric acid content up to 25 mass %. Tests of the coatings have demonstrated their sufficiently high stability, especially of those based on enamels A-20 and BK-5. These coatings are characterised by satisfactory performance and can be used for corrosion protection of the materials used in nuclear power

  3. Light Weight Ceramic Ablators for Mars Follow-on Mission Vehicle Thermal Protection System

    Science.gov (United States)

    Tran, Huy K.; Rasky, Daniel J.; Hsu, Ming-Ta; Turan, Ryan

    1994-01-01

    New Light Weight Ceramic Ablators (LCA) were produced by using ceramic and carbon fibrous substrates, impregnated with silicone and phenolic resins. The special infiltration techniques (patent pending) were developed to control the amount of organic resins in the highly porous fiber matrices so that the final densities of LCA's range from 0.22 to 0.24 g/cc. This paper presents the thermal and ablative performance of the Silicone Impregnated Reusable Ceramic Ablators (SIRCA) in simulated entry conditions for Mars-Pathfinder in the Ames 60 MW Interaction Heating Facility (I HF). Arc jet test results yielded no evidence of char erosion and mass loss at high stagnation pressures to 0.25 atm. Minimal silica melt was detected on surface char at a stagnation pressure of 0.31 atm. Four ceramic substrates were used in the production of SIRCA's to obtain the effective of boron oxide present in substrate so the thermal performance of SIRCA's. A sample of SIRCA was also exposed to the same heating condition for five cycles and no significant mass loss or recession was observed. Tensile testing established that the SIRCA tensile strength is about a factor of two higher than that of the virgin substrates. Thermogravimetric Analysis (TGA) of the char in nitrogen and air showed no evidence of free carbon in the char. Scanning Electron Microscopy of the post test sample showed that the char surface consists of a fibrous structure that was sealed with a thin layer of silicon oxide melt.

  4. SRB thermal protection systems materials test results in an arc-heated nitrogen environment

    Science.gov (United States)

    Wojciechowski, C. J.

    1979-01-01

    The external surface of the Solid Rocket Booster (SRB) will experience imposed thermal and shear environments due to aerodynamic heating and radiation heating during launch, staging and reentry. This report is concerned with the performance of the various TPS materials during the staging maneuver. During staging, the wash from the Space Shuttle Main Engine (SSME) exhust plumes impose severe, short duration, thermal environments on the SRB. Five different SRB TPS materials were tested in the 1 MW Arc Plasma Generator (APG) facility. The maximum simulated heating rate obtained in the APG facility was 248 Btu/sq ft./sec, however, the test duration was such that the total heat was more than simulated. Similarly, some local high shear stress levels of 0.04 psia were not simulated. Most of the SSME plume impingement area on the SRB experiences shear stress levels of 0.02 psia and lower. The shear stress levels on the test specimens were between 0.021 and 0.008 psia. The SSME plume stagnation conditions were also simulated.

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  6. Overview of the TFTB lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

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

  7. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  8. Accelerator driven heavy water blanket on circulating fuel

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  9. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2005-03-01

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

  10. Objectives and status of EUROfusion DEMO blanket studies

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  11. 18 CFR 284.303 - OCS blanket certificates.

    Science.gov (United States)

    2010-04-01

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

  12. Status of the EU test blanket systems safety studies

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  13. Status of the EU test blanket systems safety studies

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  14. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  15. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

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

  16. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  18. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

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

    1971-01-01

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

  19. Conformal Ablative Thermal Protection Systems (CA-TPS) for Venus and Saturn Backshells

    Science.gov (United States)

    Beck, R.; Gasch, M.; Stackpoole, M.; Wilder, M.; Boghozian, T.; Chavez-Garcia, J.; Prabhu, D.; Kazemba, C.; Venkatapathy, E.

    2015-01-01

    The new conformal ablator C-PICA, which was developed under STMD GCD, is an optimal candidate for use on the backshells for high velocity entry vehicles at both Venus and Saturn. The material has been tested at heat fluxes up to 400 Wcm2 in shear and over 1800 Wcm2 and 1.5 atm in stagnation with good results. C-PICA has similar density to PICA, but shows half the thermal penetration and similar recession at the same conditions, allowing for a lighter weight TPS to be flown. This poster for VEXAG will show the progress made in the development of the material and why it should be considered for use.

  20. Thermal protection and refurbishment of an old building. Lectures; Waermeschutz und Altbausanierung. Vortraege

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    Within the 22nd Hanseatic Reconstruction Symposium at the Baltic Seaside Heringsdorf/Usedom (Federal Republic of Germany) from 3rd to 5th November 2011, the following lectures were held: (1) Energetic refurbishment possibilities for building within existing properties by means of representative examples (F. Deitschum); (2) Constructional thermal insulation and indoor climate - for the good of the environment? (S. Groer); (3) Innovative insulating materials for the structural refurbishment? (O. Fechner); (4) Energetic half-timbering refurbishment (K. Lissner); (5) Wooden solar facades for existing buildings (U. Schwarz); (6) Timber beam bowls in a historic brickwork (U. Mueller); (7) Timber beam bowls and interior insulation (U. Ruisinger); (8) Innovative solutions for cavity filling insulations (A. Stefenelli); (9) Thermal insulating plaster - also for historical buildings (T. Stahl); (10) Experimental tension analysis of the structural behaviour of historical cross vaults (A.-J. Petereit); (10) Investigation of the increase of the flexural strength of stonework constructions with self-compacting steel fibre reinforced concrete (D. Haessler); (11) Dry and dense - the modified WTA leaflet 4-6, 'Subsequent sealing of components in contact with soil' - Content and innovations (R. Spirgatis); (12) What does the new standard DIN 68800 hold? (H. Willeitner); (13) News from the standard DIN 18195 waterproofing of buildings (H.-P. Sommer); (14) Liability of planning of the offering entrepreneur (H. Immoor); (15) Climate change and preservation of structures (W. Zillig); (16) Typical problems and deficiencies of the energetic refurbishment of old store buildings (H. Boehmer); (17) When do ex post horizontal sealings with injection agents make sense - Fundamentals for evaluation, planning and execution (F.-J. Hoelzen); (18) Drying up behaviour of stonework of different quality and at different variants of insulation (F. Antretter).

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

    Science.gov (United States)

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

    2015-01-01

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

  2. The Effects of Foam Thermal Protection System on the Damage Tolerance Characteristics of Composite Sandwich Structures for Launch Vehicles

    Science.gov (United States)

    Nettles, A. T.; Hodge, A. J.; Jackson, J. R.

    2011-01-01

    For any structure composed of laminated composite materials, impact damage is one of the greatest risks and therefore most widely tested responses. Typically, impact damage testing and analysis assumes that a solid object comes into contact with the bare surface of the laminate (the outer ply). However, most launch vehicle structures will have a thermal protection system (TPS) covering the structure for the majority of its life. Thus, the impact response of the material with the TPS covering is the impact scenario of interest. In this study, laminates representative of the composite interstage structure for the Ares I launch vehicle were impact tested with and without the planned TPS covering, which consists of polyurethane foam. Response variables examined include maximum load of impact, damage size as detected by nondestructive evaluation techniques, and damage morphology and compression after impact strength. Results show that there is little difference between TPS covered and bare specimens, except the residual strength data is higher for TPS covered specimens.

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

  4. Blanket coverage : small technology companies hope to mine some Athabasca riches of their own

    Energy Technology Data Exchange (ETDEWEB)

    Marsters, S.

    2006-09-15

    This article presented details of the Pyrogel 6350, an insulating fabric comprised of a nanotechnology-enabled flexible fabric with aerogel integrated into its matrix. Aerogel is the lightest solid known to science and is created by replacing the liquid phase in a gel with gas. Originally developed for National American Space Agency (NASA) spacesuits, the United States military uses aerogel-based blankets to provide infrared suppression around engine compartments and hot components. Designed specifically for the oilsands market, the Pyrogel 6350 combines extreme thermal performance in a flexible blanket form which is ideal for the insulation of process equipment, pipelines and vessels used in oil sands production. The Pyrogel 6350 is currently being installed on 3 kilometres of high-pressure steam lines at Devon Canada's steam-assisted gravity drainage (SAGD) Jackfish project. The fabric's R-value per inch is 4 to 6 times greater than conventional types of insulation. It was concluded that use of the aerogel blankets will result in a 3 inch reduction in pipe insulation thickness, and provide significant savings in installation costs. 2 figs.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

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

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

    International Nuclear Information System (INIS)

    Enoeda, Mikio; Akiba, Masato; Ohara, Yoshihiro

    2001-12-01

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

  8. A Thermal Physiological Comparison of Two HazMat Protective Ensembles With and Without Active Convective Cooling

    Science.gov (United States)

    Williamson, Rebecca; Carbo, Jorge; Luna, Bernadette; Webbon, Bruce W.

    1998-01-01

    Wearing impermeable garments for hazardous materials clean up can often present a health and safety problem for the wearer. Even short duration clean up activities can produce heat stress injuries in hazardous materials workers. It was hypothesized that an internal cooling system might increase worker productivity and decrease likelihood of heat stress injuries in typical HazMat operations. Two HazMat protective ensembles were compared during treadmill exercise. The different ensembles were created using two different suits: a Trelleborg VPS suit representative of current HazMat suits and a prototype suit developed by NASA engineers. The two life support systems used were a current technology Interspiro Spirolite breathing apparatus and a liquid air breathing system that also provided convective cooling. Twelve local members of a HazMat team served as test subjects. They were fully instrumented to allow a complete physiological comparison of their thermal responses to the different ensembles. Results showed that cooling from the liquid air system significantly decreased thermal stress. The results of the subjective evaluations of new design features in the prototype suit were also highly favorable. Incorporation of these new design features could lead to significant operational advantages in the future.

  9. Thermal Protection for Mars Sample Return Earth Entry Vehicle: A Grand Challenge for Design Methodology and Reliability Verification

    Science.gov (United States)

    Venkatapathy, Ethiraj; Gage, Peter; Wright, Michael J.

    2017-01-01

    Mars Sample Return is our Grand Challenge for the coming decade. TPS (Thermal Protection System) nominal performance is not the key challenge. The main difficulty for designers is the need to verify unprecedented reliability for the entry system: current guidelines for prevention of backward contamination require that the probability of spores larger than 1 micron diameter escaping into the Earth environment be lower than 1 million for the entire system, and the allocation to TPS would be more stringent than that. For reference, the reliability allocation for Orion TPS is closer to 11000, and the demonstrated reliability for previous human Earth return systems was closer to 1100. Improving reliability by more than 3 orders of magnitude is a grand challenge indeed. The TPS community must embrace the possibility of new architectures that are focused on reliability above thermal performance and mass efficiency. MSR (Mars Sample Return) EEV (Earth Entry Vehicle) will be hit with MMOD (Micrometeoroid and Orbital Debris) prior to reentry. A chute-less aero-shell design which allows for self-righting shape was baselined in prior MSR studies, with the assumption that a passive system will maximize EEV robustness. Hence the aero-shell along with the TPS has to take ground impact and not break apart. System verification will require testing to establish ablative performance and thermal failure but also testing of damage from MMOD, and structural performance at ground impact. Mission requirements will demand analysis, testing and verification that are focused on establishing reliability of the design. In this proposed talk, we will focus on the grand challenge of MSR EEV TPS and the need for innovative approaches to address challenges in modeling, testing, manufacturing and verification.

  10. Mount Protects Thin-Walled Glass or Ceramic Tubes from Large Thermal and Vibration Loads

    Science.gov (United States)

    Amato, Michael; Schmidt, Stephen; Marsh. James; Dahya, Kevin

    2011-01-01

    The design allows for the low-stress mounting of fragile objects, like thin walled glass, by using particular ways of compensating, isolating, or releasing the coefficient of thermal expansion (CTE) differences between the mounted object and the mount itself. This mount profile is lower than true full kinematic mounting. Also, this approach enables accurate positioning of the component for electrical and optical interfaces. It avoids the higher and unpredictable stress issues that often result from potting the object. The mount has been built and tested to space-flight specifications, and has been used for fiber-optic, optical, and electrical interfaces for a spaceflight mission. This mount design is often metal and is slightly larger than the object to be mounted. The objects are optical or optical/electrical, and optical and/or electrical interfaces are required from the top and bottom. This requires the mount to be open at both ends, and for the object s position to be controlled. Thin inside inserts at the top and bottom contact the housing at defined lips, or edges, and hold the fragile object in the mount. The inserts can be customized to mimic the outer surface of the object, which further reduces stress. The inserts have the opposite CTE of the housing material, partially compensating for the CTE difference that causes thermal stress. A spring washer is inserted at one end to compensate for more CTE difference and to hold the object against the location edge of the mount for any optical position requirements. The spring also ensures that any fiber-optic or optic interface, which often requires some pressure to ensure a good interface, does not overstress the fragile object. The insert thickness, material, and spring washer size can be traded against each other to optimize the mount and stresses for various thermal and vibration load ranges and other mounting requirements. The alternate design uses two separate, unique features to reduce stress and hold the

  11. Water intake and fish protection sytems for thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Kuz'min, D.O.; Lukashevich, V.S.

    1986-01-01

    Various designs of water intake and fish protection systems for TPP and NPP are considered. Water intake systems are divided into shore and outside shore types. There are two main modifications of the latter - opened and closed. The closed systems are more complex for construction and maintenance, but their negative influence on environment is considerably weaker. In disigning of water intake systems basic efforts are directed at optimization of a water intake device disposition, development of reliable repellents for fish, as well as devices for fish catch and return from the water intake region. A special attention is paid to the problem of preventing their icing. The conclusion of expedience of introducing into the water purification system reliable, soft mechanical barriers for fish equipped with means affecting its behaviour and preventing contacts of fish and water intake system elements was drawn

  12. 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.

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

    International Nuclear Information System (INIS)

    1995-01-01

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

  14. Thermal Protection of Carbon Fiber-Reinforced Composites by Ceramic Particles

    Directory of Open Access Journals (Sweden)

    Baljinder Kandola

    2016-06-01

    Full Text Available The thermal barrier efficiency of two types of ceramic particle, glass flakes and aluminum titanate, dispersed on the surface of carbon-fiber epoxy composites, has been evaluated using a cone calorimeter at 35 and 50 kW/m2, in addition to temperature gradients through the samples’ thicknesses, measured by inserting thermocouples on the exposed and back surfaces during the cone tests. Two techniques of dispersing ceramic particles on the surface have been employed, one where particles were dispersed on semi-cured laminate and the other where their dispersion in a phenolic resin was applied on the laminate surface, using the same method as used previously for glass fiber composites. The morphology and durability of the coatings to water absorption, peeling, impact and flexural tension were also studied and compared with those previously reported for glass-fiber epoxy composites. With both methods, uniform coatings could be achieved, which were durable to peeling or water absorption with a minimal adverse effect on the mechanical properties of composites. While all these properties were comparable to those previously observed for glass fiber composites, the ceramic particles have seen to be more effective on this less flammable, carbon fiber composite substrate.

  15. Protection of the skin against occupational and operational ultraviolet and thermal radiation

    International Nuclear Information System (INIS)

    Wiskemann, A.

    1980-01-01

    When irradiation with short wave ultraviolet (UVB) exceed the threshold doses, the eye as well as the skin react with an acute inflammation. After chronic exposure to both radiations the skin is altered as a farmers skin. Thermal visible and infrared radiation may produce a local combustion or a livedo or a general hyperthermia. Many possibilities of an occupational exposition to natural or artificial optical radiation are listed. Until now no exposure limits have been recommended in the Federal Republic of Germany. The biologic effective radiant exposure can be calculated from the spectral distribution of the irradiance. The resulting value should be clearly lower than the threshold doses for the UV-keratoconjunctivitis and for the UV-erythema of the skin. Artificial light sources have to be closed exept the useful radiation beam. When this is impossible and in case of natural radiation, the skin must be shielded by clothing and/or by sunscreen preparations. Photosensitizers as tar products have to be kept away from the skin. (orig.) 891 MG/orig. 892 HIS [de

  16. Protective role of vitamin E preconditioning of human dermal fibroblasts against thermal stress in vitro.

    Science.gov (United States)

    Butt, Hira; Mehmood, Azra; Ali, Muhammad; Tasneem, Saba; Anjum, Muhammad Sohail; Tarar, Moazzam N; Khan, Shaheen N; Riazuddin, Sheikh

    2017-09-01

    Oxidative microenvironment of burnt skin restricts the outcome of cell based therapies of thermal skin injuries. The aim of this study was to precondition human dermal fibroblasts with an antioxidant such as vitamin E to improve their survival and therapeutic abilities in heat induced oxidative in vitro environment. Fibroblasts were treated with 100μM vitamin E for 24h at 37°C followed by heat shock for 10min at 51°C in fresh serum free medium. Preconditioning with vitamin E reduced cell injury as demonstrated by decreased expression of annexin-V, cytochrome p450 (CYP450) mediated oxidative reactions, senescence and release of lactate dehydrogenase (LDH) accomplished by down-regulated expression of pro-apoptotic BAX gene. Vitamin E preconditioned cells exhibited remarkable improvement in cell viability, release of paracrine factors such as epidermal growth factor (EGF), basic fibroblast growth factor (bFGF), vascular endothelial growth factor (VEGF), stromal derived factor-1alpha (SDF-1α) and also showed significantly up-regulated levels of PCNA, VEGF, BCL-XL, FGF7, FGF23, FLNβ and Col7α genes presumably through activation of phosphatidylinositol 3-kinase (PI3-K)/Akt pathway. The results suggest that pretreatment of fibroblasts with vitamin E prior to transplantation in burnt skin speeds up the wound healing process by improving the antioxidant scavenging responses in oxidative environment of transplanted burn wounds. Copyright © 2017 Elsevier Inc. All rights reserved.

  17. Aerothermal Ground Testing of Flexible Thermal Protection Systems for Hypersonic Inflatable Aerodynamic Decelerators

    Science.gov (United States)

    Bruce, Walter E., III; Mesick, Nathaniel J.; Ferlemann, Paul G.; Siemers, Paul M., III; DelCorso, Joseph A.; Hughes, Stephen J.; Tobin, Steven A.; Kardell, Matthew P.

    2012-01-01

    Flexible TPS development involves ground testing and analysis necessary to characterize performance of the FTPS candidates prior to flight testing. This paper provides an overview of the analysis and ground testing efforts performed over the last year at the NASA Langley Research Center and in the Boeing Large-Core Arc Tunnel (LCAT). In the LCAT test series, material layups were subjected to aerothermal loads commensurate with peak re-entry conditions enveloping a range of HIAD mission trajectories. The FTPS layups were tested over a heat flux range from 20 to 50 W/cm with associated surface pressures of 3 to 8 kPa. To support the testing effort a significant redesign of the existing shear (wedge) model holder from previous testing efforts was undertaken to develop a new test technique for supporting and evaluating the FTPS in the high-temperature, arc jet flow. Since the FTPS test samples typically experience a geometry change during testing, computational fluid dynamic (CFD) models of the arc jet flow field and test model were developed to support the testing effort. The CFD results were used to help determine the test conditions experienced by the test samples as the surface geometry changes. This paper includes an overview of the Boeing LCAT facility, the general approach for testing FTPS, CFD analysis methodology and results, model holder design and test methodology, and selected thermal results of several FTPS layups.

  18. Dustless Process for Minor Actinide-Bearing Blanket Fabrication

    International Nuclear Information System (INIS)

    Caisso, M.; Lebreton, F.; Horlait, D.; Delahaye, Th.; Picart, S.; Martin, Ph.M.; Renard, C.; Roussel, P.; Neuville, D.R.; Belin, R.C.; Dardenne, K.; Rothe, J.; Ayral, A.

    2015-01-01

    U 1-x Am x O 2±δ mixed-oxides are considered promising compounds for americium heterogeneous transmutation in fast neutron reactor. At lab-scale, the fabrication of americium bearing blankets (AmBB) under the form of ceramic pellets, required for irradiation, follows a powder metallurgy route which generates highly contaminant fine particles. Considering scale-up, dustless processes that can avoid particle dispersion in the fabrication lines are thus recommended. With this aim, the development of an innovative route called calcined resin microsphere pelletizing (CRMP) process has been initiated. The general approach consists in synthesising mixed-oxide microsphere precursors from beads of ion exchange resin through an adaptation of the weak acid resin process (WAR), and their pelletizing before sintering. This study focuses on the microsphere synthesis and particularly on the mechanisms implied during the thermal conversion of metal loaded ion exchange resin in porous mixed-oxide microspheres. The results are discussed, in a first time, on the basis of the synthesis of oxide microspheres integrating uranium and americium surrogates (Ce and Gd respectively) before a transposition to the highly active materials in a second time. (authors)

  19. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

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

  20. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

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