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Sample records for liquid lithium blanket

  1. 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

  2. Chemical processing of liquid lithium fusion reactor blankets

    International Nuclear Information System (INIS)

    Weston, J.R.; Calaway, W.F.; Yonco, R.M.; Hines, J.B.; Maroni, V.A.

    1979-01-01

    A 50-gallon-capacity lithium loop constructed mostly from 304L stainless steel has been operated for over 6000 hours at temperatures in the range from 360 to 480 0 C. This facility, the Lithium Processing Test Loop (LPTL), is being used to develop processing and monitoring technology for liquid lithium fusion reactor blankets. Results of tests of a molten-salt extraction method for removing impurities from liquid lithium have yielded remarkably good distribution coefficients for several of the more common nonmetallic elements found in lithium systems. In particular, the equilibrium volumetric distribution coefficients, D/sub v/ (concentration per unit volume of impurity in salt/concentration per unit volume of impurity in lithium), for hydrogen, deuterium, nitrogen and carbon are approx. 3, approx. 4, > 10, approx. 2, respectively. Other studies conducted with a smaller loop system, the Lithium Mini-Test Loop (LMTL), have shown that zirconium getter-trapping can be effectively used to remove selected impurities from flowing lithium

  3. 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

  4. Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Mistrangelo, C., E-mail: chiara.mistrangelo@kit.edu; Bühler, L.

    2014-10-15

    Highlights: • MHD flows in model geometries of DCLL blanket manifolds. • Study of velocity, pressure distributions and flow partitioning in parallel ducts. • Flow partitioning affected by 3D MHD pressure drop and velocity distribution in the expanding zone. • Reduced pressure drop in a continuous expansion compared to a sudden expansion. - Abstract: An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure. Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules. In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.

  5. Evaluation of compatibility of flowing liquid lithium curtain for blanket with core plasma in fusion reactors

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua; Peng Lilin; Yan Jiancheng

    2003-01-01

    A global model analysis of the compatibility of flowing liquid lithium curtain for blanket with core plasma has been performed. The relationships between the surface temperature of lithium curtain and mean effective plasma charges, fuel dilution and produced fusion power have been obtained. Results show that under normal circumstances, the evaporation of liquid lithium does not affect Z eff seriously, but affects fuel dilution and fusion power sensitively. The authors have investigated the relationships between the flow velocity of liquid lithium and its surface temperature rise based on the conditions of the option II of the fusion experimental breeder (FEB-E) design with reversed shear configuration and fairly high power density. The authors concluded that the effects of evaporation from liquid lithium curtain for FEB-E on plasma are negligible even if the flow velocity of liquid lithium is as low as 0.5 m·s -1 . Finally, the sputtering yield of liquid lithium saturated by hydrogen isotopes is briefly discussed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  7. A conceptual composite blanket design for the Tokamak type of thermonuclear reactor incorporating thermoelectric pumping of liquid lithium

    International Nuclear Information System (INIS)

    Dutta Gupta, P.B.

    1981-01-01

    The conceptual liquid lithium blanket design for the tokamak type of thermonuclear reactor put forward is a modification of the initial simple but novel design concept enunciated earlier that exploits the availability of suitably oriented magnetic fields and temperature gradients within the blanket to pump the liquid as has been shown feasible by laboratory model experiments. The modular construction of the blanket cells is retained but the earlier simple back to back double spiralling channel module is replaced by a composite unit of three radially nested layer-structures to optimise heat and tritium extraction from the blanket. The layer-structure at the first wall generates liquid lithium circulation by thermoelectric magnetohydrodynamic forces and the segregated double spiralling channels serve as inlet-outlet driving devices. The outermost layer-structure is cooled by helium. Liquid lithium in the intermediate layer-structure is pumped at a very slow rate. The choice of the relative dimensional proportions of the three layer-structure and the channel cross-section, material property and the spiralling contour is of critical importance for the design. This paper presents the design data for a conceptual design of such a blanket with a 5000 MW (th) rating. (author)

  8. Comparison of lithium and the eutectic lead lithium alloy, two candidate liquid metal breeder materials for self-cooled blankets

    International Nuclear Information System (INIS)

    Malang, S.; Mattas, R.

    1994-06-01

    Liquid metals are attractive candidates for both near-term and long-term fusion applications. The subjects of this comparison are the differences between the two candidate liquid metal breeder materials Li and LiPb for use in breeding blankets in the areas of neutronics, magnetohydrodynamics, tritium control, compatibility with structural materials, heat extraction system, safety, and required R ampersand D program. Both candidates appear to be promising for use in self-cooled breeding blankets which have inherent simplicity with the liquid metal serving as both breeders and coolant. The remaining feasibility question for both breeder materials is the electrical insulation between liquid metal and duct walls. Different ceramic coatings are required for the two breeders, and their crucial issues, namely self-healing of insulator cracks and radiation induced electrical degradation are not yet demonstrated. Each liquid metal breeder has advantages and concerns associated with it, and further development is needed to resolve these concerns

  9. 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

  10. 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)

  11. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

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

    1992-06-01

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

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of this study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. The primary results of the study are as follows: a) the lithium-lead blanket achieves a higher TBR with a smaller blanket thickness relative to the lithium blanket; b) the lithium blanket generates more energy per fusion neutron relative to the lithium-lead blanket; c) among the possible reflector materials, the carbon reflector produces the highest TBR; d) the high-Z reflector materials (Mo, Cu, W, or steel) generate more energy per fusion neutron and produce smaller TBRs relative to the carbon reflector; e) lithium-6 enrichment is required for the lithium-lead blanket to reduce the total blanket thickness; and f) the energy deposition per fusion neutron reaches a saturation as the blanket thickness, the fraction of the high-Z material in the reflector, or the reflector zone thickness increases (this allows one to design the blanket for a specific TBR without reducing the energy production)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  14. 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)

  15. 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

  16. Structural effects on fusion reactor blankets due to liquid metals in magnetic fields

    International Nuclear Information System (INIS)

    Lehner, J.R.; Reich, M.; Powell, J.R.

    1976-01-01

    The transient stress distribution caused in the blanket structure when the plasma current suddenly switches off in a time short compared to the L/R decay time of the liquid metal blanket was studied. Poloidal field of the plasma will induce a current to flow in the liquid metal and blanket walls. Since the resistance of the liquid lithium will be much less than that of the metal walls, the current can be considered as flowing around the blanket near the cross section perimeter, but in the lithium

  17. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-12-01

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

  18. The impact of tritium solubility and diffusivity on inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Caorlin, M.; Gervasini, G.; Reiter, F.

    1988-01-01

    The authors reviewed hydrogen solubility and diffusivity data for liquid lithium-based compounds which are potential breeding blanket materials in NET-type fusion devices. These data have been used to assess tritium permeation and inventory in separately cooled NET blankets and in self cooled blankets with a vanadium first wall. The results for the separately cooled NET-liquid breeder show that tritium permeation is negligible for lithium, a serious problem for Pb-17Li and a critical one for Flibe. The total tritium inventory is lowest in lithium, high in Pb-17Li and very high in Flibe. The high tritium partial pressure for Flibe or Pb-17Li can be reduced in a self cooled blanket with a vanadium first wall. Permeation into the plasma reduces the blanket tritium inventory and permeation. Tritium recovery can be combined with the plasma exhaust

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

    Science.gov (United States)

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

    2017-11-01

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

  20. An alternative high breeding radio design concept with liquid breeder for the NET/INTOR blanket

    International Nuclear Information System (INIS)

    Avanzini, P.G.; Cardella, A.; Raia, G.; Rosatelli, F.; Farfaletti-Casali, F.

    1984-01-01

    A liquid lithium tubolar breeding blanket concept has been studied which could be applied to NET/INTOR or other next generation Tokamak reactors. A high breeding ratio can be achieved using a moderator medium, without enriching lithium in the Li6 percentage. Preliminary neutron and gamma flux and thermohydraulics calculations have shown the feasibility and efficiency of our concept. (author)

  1. 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

  2. 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

  3. 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

  4. Overview of the TFTR Lithium Blanket Module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

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

  5. Solubility of lithium deuteride in liquid lithium

    International Nuclear Information System (INIS)

    Veleckis, E.; Yonco, R.M.; Maroni, V.A.

    1977-01-01

    The solubility of LiD in liquid lithium between the eutectic and monotectic temperatures was measured using a direct sampling method. Solubilities were found to range from 0.0154 mol.% LiD at 199 0 C to 3.32 mol.% LiD at 498 0 C. The data were used in the derivation of an expression for the activity coefficient of LiD as a function of temperature and composition and an equation relating deuteride solubility and temperature, thus defining the liquidus curve. Similar equations were also derived for the Li-LiH system using the existing solubility data. Extrapolation of the liquidus curves yielded the eutectic concentrations (0.040 mol.% LiH and 0.035 mol.% LiD) and the freezing point depressions (0.23 0 C for Li-LiH and 0.20 0 C for Li-LiD) at the eutectic point. The results are compared with the literature data for hydrogen and deuterium. The implications of the relatively high solubility of hydrogen isotopes in lithium just above the melting point are discussed with respect to the cold trapping of tritium in fusion reactor blankets. (Auth.)

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

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

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

    1997-07-01

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

  10. Neutronics and activation analysis of lithium-based ternary alloys in IFE blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, Alejandra, E-mail: aleja311@berkeley.edu [University of California Berkeley, Berkeley, CA 94706 (United States); Kramer, Kevin [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA (United States); Meier, Wayne; DeMuth, James; Reyes, Susana [TerraPower, Bellevue, WA 98005 (United States); Fratoni, Massimiliano [University of California Berkeley, Berkeley, CA 94706 (United States)

    2016-06-15

    Highlights: • Monte Carlo calculations were performed on numerous lithium ternary alloys. • Elements with high neutron multiplication performed well with low absorbers. • Enriching lithium decreases minimum lithium concentration of alloys by 60% or more. • Alloys that performed well neutronically were selected for activation calculations. • Alloys activated, except LiBaBi, do not pose major environmental or safety concerns. - Abstract: An attractive feature of using liquid lithium as the breeder and coolant in fusion blankets is that it has very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns. The Lawrence Livermore National Laboratory is carrying an effort to develop a lithium-based ternary alloy that maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) and at the same time reduces overall flammability concerns. This study evaluates the neutronics performance of lithium-based alloys in the blanket of an inertial fusion energy chamber in order to inform such development. 3-D Monte Carlo calculations were performed to evaluate two main neutronics performance parameters for the blanket: tritium breeding ratio (TBR), and the fusion energy multiplication factor (EMF). It was found that elements that exhibit low absorption cross sections and higher q-values such as Pb, Sn, and Sr, perform well with those that have high neutron multiplication such as Pb and Bi. These elements meet TBR constrains ranging from 1.02 to 1.1. However, most alloys do not reach EMFs greater than 1.15. Additionally, it was found that enriching lithium with {sup 6}Li significantly increases the TBR and decreases the minimum lithium concentration by more than 60%. The amount of enrichment depends on how much total lithium is in the alloy to begin with. Alloys that performed well in the TBR

  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. Experimental study of gaseous lithium deuterides and lithium oxides. Implications for the use of lithium and Li2O as breeding materials in fusion reactor blankets

    International Nuclear Information System (INIS)

    Ihle, H.R.; Wu, C.H.; Kudo, H.

    1980-01-01

    In addition to LiH, which has been studied extensively by optical spectroscopy, the existence of a number of other stable lithium hydrides has been predicted theoretically. By analysis of the saturated vapour over dilute solutions of the hydrogen isotopes in lithium, using Knudsen effusion mass spectrometry, all lithium hydrides predicted to be stable were found. Solutions of deuterium in lithium were used predominantly because of practical advantages for mass spectrometric measurements. The heats of dissociation of LiD, Li 2 D, LiD 2 and Li 2 D 2 , and the binding energies of their singly charged positive ions were determined, and the constants of the gas/liquid equilibria were calculated. The existence of these lithium deuterides in the gas phase over solutions of deuterium in lithium leads to enrichment of deuterium in the gas above 1240 K. The enrichment factor, which increases exponentially with temperature and is independent of concentration for low concentrations of deuterium in the liquid, was determined by Rayleigh distillation experiments. It was found that it is thermodynamically possible to separate deuterium from lithium by distillation. One of the alternatives to the use of lithium in (D,T)-fusion reactors as tritium-breeding blanket material is to employ solid lithium oxide. This has a high melting point, a high lithium density and still favourable tritium-breeding properties. Because of its rather high volatility, an experimental study of the vaporization of Li 2 O was undertaken by mass spectrometry. It vaporizes to give lithium and oxygen, and LiO, Li 2 O, Li 3 O and Li 2 O 2 . The molecule Li 3 O was found as a new species. Heats of dissociation, binding energies of the various ions and the constants of the gas/solid equilibria were determined. The effect of using different materials for the Knudsen cells and the relative thermal stabilities of lithium-aluminium oxides were also studied. (author)

  13. 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

  14. 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

  15. Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

    International Nuclear Information System (INIS)

    Klix, A.; Ochiai, K.; Nishitani, T.; Takahashi, A.

    2004-01-01

    At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6 Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation

  16. The lithium blanket program at the LOTUS facility

    International Nuclear Information System (INIS)

    File, J.; Haldy, P.A.; Quanci, J.

    1987-01-01

    An experimental program of neutron transport studies of the lithium Blanket Module (LBM) carried out with the LOTUS point-neutron source at the Ecole Polytechnique Federale de Lausanne (EPTL), Switzerland has been concluded. The major objectives of this program are to perform a series of neutron transport and tritium breeding experiments to qualify the LBM for future experiments on toroidal fusion devices such as TFTR to perform neutron multiplier experiments on the LBM employing various materials in a removable slab geometry; and, to compare the experimental results of radiation dosimetry and tritium breeding with the calculations of two and three dimensional neutron transport codes. An overview of the results from the radiation dosimetry and tritium assay are presented and compared to the two and three dimensional neutron transport codes

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  18. 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

  19. 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

  20. FTU cooled liquid lithium upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Iafrati, M., E-mail: matteo.iafrati@enea.it [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Apicella, M.L.; Boncagni, L. [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Lyublinski, I. [JSC “RED STAR”, Moscow (Russian Federation); Mazzitelli, G. [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Vertkov, A. [JSC “RED STAR”, Moscow (Russian Federation)

    2017-04-15

    In the framework of the liquid lithium limiter experiment in Frascati a new auxiliary system was developed in order to provide a better control of the energy fluid vector. The cooled liquid lithium system (CLL) was installed for the first time at the end of 2013, it uses overheated water to heat the lithium and to extract, at the same time, the heat from the metal surface when it gets wet by the plasma. A first version of the system, developed and presented in previous papers, has been modified to optimize the heat flux measurement on the liquid lithium surface. The changes include a new power supply logic for the heating system, new sensors and new read-out electronics compatible with the implementation of a real time control system. The prototype was updated with the aim of achieving a low cost and versatile control system.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    International Nuclear Information System (INIS)

    Talbot, J.B.

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approx. 15 at. % lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility ( 0 C, the extraction process is not attractive

  3. A low-risk aqueous lithium salt blanket for engineering test reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-09-01

    A simple blanket concept is proposed based on 1-3 wt.% lithium dissolved as a salt in low temperature (80 degrees C) and low pressure (0.1 MPa) water. This concept can provide, for example, a 0.5 tritium breeding ratio with 60% steel structure and 70% coverage. The use of neutron multipliers, other structural materials (especially zirconium alloys), higher coverage and higher lithium salt concentrations allows tritium breeding ratios over unity if necessary. Other advantages of this concept include the simple shield-like geometry, substantial structural volume for mechanical strength, excellent heat transfer ability of water coolant, efficient neutron and gamma shielding through the combination of high-Z structure and low-Z water, and conventional tritium recovery and control technology. This concept could initially provide the shielding needs for an engineering test reactor and later, by the addition of lithium salt and tritium recovery systems, also provide tritium breeding. This staged operation and liquid breeder/coolant allows control over the tritium inventory in the device without machine disassembly. 14 refs

  4. 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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  6. Trace hydrogen extraction from liquid lithium tin alloy

    International Nuclear Information System (INIS)

    Xie Bo; Hu Rui; Xie Shuxian; Weng Kuiping

    2010-01-01

    In order to finish the design of tritium extraction system (TES) of fusion fission hybrid reactor (FFHR) tritium blanket, involving the dynamic mathematical model of liquid metal in contact with a gaseous atmosphere, approximate mathematical equation of tritium in lithium tin alloy was deduced. Moreover, carrying process used for trace hydrogen extraction from liquid lithium tin alloy was investigated with hydrogen being used to simulate tritium in the study. The study results indicate that carrying process is effective way for hydrogen extraction from liquid lithium tin alloy, and the best flow velocity of carrier gas is about 4 L/min under 1 kg alloy temperatures and carrying numbers are the main influencing factors of hydrogen number. Hydrogen extraction efficiency can reach 85% while the alloy sample is treated 6 times at 823 K. (authors)

  7. Design data, liquid distributors and condenser for a distillation column to enrich tritium in metallic lithium

    International Nuclear Information System (INIS)

    Barnert, E.

    1984-01-01

    Tritium, one fuel component of the fusion reactor is bred from the reactors blanket material lithium. Before extracting the tritium from, for instance, metallic lithium by permeation it has to be enriched in the lithium by high temperature distillation. In this report design data for a column for high temperature distillation are given. About the testing of two distributors for small liquid quantities and of a condenser is reported. (orig.) [de

  8. Deuterium retention in liquid lithium

    International Nuclear Information System (INIS)

    Baldwin, M.J.; Doerner, R.P.; Luckhardt, S.C.; Conn, R.W.

    2002-01-01

    Measurements of deuterium retention in samples of lithium exposed in the liquid state to deuterium plasma are reported. Retention was measured as a function of plasma ion dose in the range 6x10 19 -4x10 22 D atoms and exposure temperature between 523 and 673 K using thermal desorption spectrometry. The results are consistent with the full uptake of all deuterium ions incident on the liquid metal surface and are found to be independent of the temperature of the liquid lithium over the range explored. Full uptake, consistent with very low recycling, continues until the sample is volumetrically converted to lithium deuteride. This occurs for exposure temperatures where the gas pressure during exposure was both below and slightly above the corresponding decomposition pressure for LiD in Li. (author)

  9. Potential and problems of an aqueous lithium salt solution blanket for NET

    International Nuclear Information System (INIS)

    Kuechle, M.; Bojarsky, E.; Dorner, S.; Fischer, U.; Reimann, J.; Reiser, H.

    1987-07-01

    The report describes design studies on a water cooled in-vessel shield blanket for NET and its modification into an aqueous lithium salt blanket. The shield blankets are exchangable against breeding blankets and fulfill their shielding and heat removal functions. Emphasis is on simplicity and reliability. The water cooled shield is a large steel container in the shape of the blanket segment which is filled by water and containes a grid structure of poloidally arranged steel plates. The water flows several times in poloidal direction through the channels formed by the steel plates and is thereby heated up from 40degC to 70degC. When the water is replaced by an aqueous lithium salt solution the shield can be converted into a tritium breeding blanket without any design modification or invessel component replacement. When compared with other concepts this blanket has the advantage that the solution can replace water cooling also in the divertor and in segments dedicated to plasma heating and diagnostics, what increases the coverage considerably. Extensive three-dimensional neutronics calculations were done which, together with literature studies on candidate materials, corrosion, and tritium recovery led to a first assessment of the concept. There is an indication that no major corrosion problems are to be expected in the low temperature region envisaged. Tritium recovery capital costs were estimated to be in the 20 MECU to 50 MECU range and tritium breeding ratio is comparable to the best breeding blanket. (orig./GG) [de

  10. Modeling of liquid-metal corrosion/deposition in a fusion reactor blanket

    International Nuclear Information System (INIS)

    Malang, S.; Smith, D.L.

    1984-04-01

    A model has been developed for the investigation of the liquid-metal corrosion and the corrosion product transport in a liquid-metal-cooled fusion reactor blanket. The model describes the two-dimensional transport of wall material in the liquid-metal flow and is based on the following assumptions: (1) parallel flow in a straight circular tube; (2) transport of wall material perpendicular to the flow direction by diffusion and turbulent exchange; in flow direction by the flow motion only; (3) magnetic field causes uniform velocity profile with thin boundary layer and suppresses turbulent mass exchange; and (4) liquid metal at the interface is saturated with wall material. A computer code based on this model has been used to analyze the corrosion of ferritic steel by lithium lead and the deposition of wall material in the cooler part of a loop. Three cases have been investigated: (1) ANL forced convection corrosion experiment (without magnetic field); (2) corrosion in the MARS liquid-metal-cooled blanket (with magnetic field); and (3) deposition of wall material in the corrosion product cleanup system of the MARS blanket loop

  11. Experimental investigation of MHD pressure losses in a mock-up of a liquid metal blanket

    Science.gov (United States)

    Mistrangelo, C.; Bühler, L.; Brinkmann, H.-J.

    2018-03-01

    Experiments have been performed to investigate the influence of a magnetic field on liquid metal flows in a scaled mock-up of a helium cooled lead lithium (HCLL) blanket. During the experiments pressure differences between points on the mock-up have been recorded for various values of flow rate and magnitude of the imposed magnetic field. The main contributions to the total pressure drop in the test-section have been identified as a function of characteristic flow parameters. For sufficiently strong magnetic fields the non-dimensional pressure losses are practically independent on the flow rate, namely inertia forces become negligible. Previous experiments on MHD flows in a simplified test-section for a HCLL blanket showed that the main contributions to the total pressure drop in a blanket module originate from the flow in the distributing and collecting manifolds. The new experiments confirm that the largest pressure drops occur along manifolds and near the first wall of the blanket module, where the liquid metal passes through small openings in the stiffening plates separating two breeder units. Moreover, the experimental data shows that with the present manifold design the flow does not distribute homogeneously among the 8 stacked boxes that form the breeding zone.

  12. A vanadium alloy for the application in a liquid metal blanket of a fusion reactor

    Science.gov (United States)

    Borgstedt, H. U.; Grundmann, M.; Konys, J.; Perić, Z.

    1988-07-01

    The vanadium alloy V3Ti1Si has been corrosion tested in liquid lithium and the eutectic alloy Pb-17Li at 550°C. This alloy has a comparable corrosion resistance to the alloy V15Cr5Ti in lithium. In this molten metal it is superior to stainless steel AISI 316. In the Pb-17Li melt it is even superior to martensitic steels. The alloy has only a weak tendency to be dissolved. It is sensitive to an exchange of non-metallic elements, which causes the formation of a hardened surface layer. These chemical effects are influenced by the mass and surface ratios of the vanadium alloy to the molten metals and other structural materials. These ratios are unfavorable in the two test loops. The effects might be less pronounced in a vanadium alloy/liquid metal fusion reactor blanket.

  13. Beryllium and lithium resource requirements for solid blanket designs for fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.R.

    1975-01-01

    The lithium and beryllium requirements are analyzed for an economy of 10 6 MW(e) CTR 3 capacity using solid blanket fusion reactors. The total lithium inventory in fusion reactors is only approximately 0.2 percent of projected U. S. resources. The lithium inventory in the fusion reactors is almost entirely 6 Li, which must be extracted from natural lithium. Approximately 5 percent of natural lithium can be extracted as 6 Li. Thus the total feed of natural lithium required is approximately 20 times that actually used in fusion reactors, or approximately 4 percent of U. S. resources. Almost all of this feed is returned to the U. S. resource base after 6 Li is extracted, however. The beryllium requirements are on the order of 10 percent of projected U. S. resources. Further, the present cost of lithium and the cost of beryllium extraction could both be increased tenfold with only minor effects on CTR capital cost. Such an increase should substantially multiply the economically recoverable resources of lithium and beryllium. It is concluded that there are no lithium or beryllium resource limitations preventing large-scale implementation of solid blanket fusion reactors. (U.S.)

  14. Neutronic analysis of the European reference design of the water cooled lithium lead blanket for a DEMOnstration reactor

    International Nuclear Information System (INIS)

    Petrizzi, L.

    1994-01-01

    Water cooled lithium lead blankets, using liquid Pb-17Li eutectic both as breeder and neutron multiplier material, and martensitic steel as structural material, represent one of the four families under development in the European DEMO blanket programme. Two concepts were proposed, both reaching tritium breeding self-sufficiency: the 'box-shaped' and the 'cylindrical modules'. Also to this scope a new concept has been defined: 'the single box'. A neutronic analysis of the 'single box' is presented. A full 3-D model including the whole assembly and many of the reactor details (divertors, holes, gaps) has been defined, together with a 3-D neutron source. A tritium breeding ration (TBR) value of 1.19 confirms the tritium breeding self-sufficiency of the design. Selected power densities, calculated for the different materials and zones, are here presented. Some shielding capability considerations with respect to the toroidal field coil system are presented too. (author) 10 refs.; 3 figs.; 3 tabs

  15. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    International Nuclear Information System (INIS)

    Talbot, J.B.

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approximately 15 at. percent lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility (less than 10 ppb) at temperatures ranging from 500 to 700 0 C, the extraction process is not attractive

  16. Experimental system design of liquid lithium-lead alloy bubbler for DFLL-TBM

    International Nuclear Information System (INIS)

    Xie Bo; Li Junge; Xu Shaomei; Weng Kuiping

    2011-01-01

    The liquid lithium-lead alloy bubbler is a very important composition in the tritium unit of Chinese Dual-Functional Lithium Lead Test Blanket Module (DFLL-TBM). In order to complete the construction and run of the bubbler experimental system,overall design of the system, main circuit design and auxiliary system design have been proposed on the basis of theoretical calculations for the interaction of hydrogen isotope with lithium-lead alloy and experiment for hydrogen extraction from liquid lithium-lead alloy by bubbling with rotational jet nozzle. The key of this design is gas-liquid exchange packed column, to achieve the measurement and extraction of hydrogen isotopes from liquid lithium-lead alloy. (authors)

  17. Development of insulating coatings for liquid metal blankets

    International Nuclear Information System (INIS)

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

    1994-07-01

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

  18. Interactions of D-T neutrons in graphite and lithium blankets of fusion reactors

    International Nuclear Information System (INIS)

    Ofek, R.

    1986-05-01

    The present study deals with integral experiment and calculation of neutron energy spectra in bulks of graphite which is used as a reflector in blankets of fusion reactors, and lithium, the material of the blanket on which lithium is bred due to neutron interactions. The collimated beam configuration enables - due to the almost monoenergeticity and unidirectionality of the neutrons impinging on the target - to identify fine details in the measured spectra, and also facilitates the absolute normalization of the spectra. The measured and calculated spectra are generally in a good agreement and in a very good agreement at mesh points close to the system axis. A few conclusions may be drawn: a) the collimated beam source configuration is a sensitive tool for measuring neutron energy spectra with a high resolution, b) the method of unfolding proton-recoil spectra measured with a NE-213 scintillator should be improved, c) MCNP and DOT 4.2 may be used as complementary codes for neutron transport calculations of fusion blankets and deep-penetration problems, d) the updating of the cross-section libraries and checking by integral experiments is highly important for the design of fusion blankets. The present study may be regarded as an important course in the research and development of tools for the design of fusion blankets

  19. Rotating liquid blanket for a toroidal fusion reator

    International Nuclear Information System (INIS)

    Moir, R.W.

    1987-01-01

    A novel blanket concept is presented for toroidal geometry in which many of the limitations imposed by a first wall are avoided by not having a first wall in the usual sense. The blanket consists of a rapidly rotating, low-vapor-pressure liquid that has a sharp boundary with the vacuum region. Nozzles inject ja continuous layer of cool liquid on the inner surface. The noncentricity of the plasma is maintained so that the plasma scrape-off region intersects the rotating liqid in a localized region. This noncentricity allows sufficient space so that the scrape-off plasma layer will not bombard the nozzles, whch penetrate through the rotating liquid. This liquid ''first wall'' is bombarded by the plasma, resulting in heat deposition, sputtering, and evaporation during the short time before the exposed liquid is covered by fresh, cool liquid from the nozzles. The advantages of this reactor concept appear to be very high wall loadings (speculated to be over 10 MW/m 2 ) and long component lifetime, both crucial economic factors. The nozzles are designed for easy replacement. The reactor's disatvantage is its enormous potential for plasma contamination by impurities. (orig.)

  20. Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

    Energy Technology Data Exchange (ETDEWEB)

    Pitulski, R.H.

    1979-10-01

    A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U/sup 233/ in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U/sup 233/, Pu/sup 239/, and H/sup 3/ production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m/sup -2/) exposure period. Although the results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids.

  1. Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

    International Nuclear Information System (INIS)

    Pitulski, R.H.

    1979-10-01

    A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U 233 in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U 233 , Pu 239 , and H 3 production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m -2 ) exposure period. Although the results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids

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

    Science.gov (United States)

    Jolodosky, Alejandra

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

  3. Tritium isolation from lithium inorganic compounds applicable to thermonuclear reactor breeding blanket

    International Nuclear Information System (INIS)

    Vasil'ev, V.G.; Ershova, Z.V.; Nikiforov, A.S.

    1982-01-01

    Tritium separation from inorganic lithium compounds: Li 2 O, LiAlO 2 , Li 2 SiO 3 , Li 4 SiO 4 , LiF, LiBeF 3 , Li 2 BeF 4 irradiated with a beam of a gamma facility and a nuclear reactor, has been studied. In the first case the gas phase is absent. In the latter one- the tritium amount in the gas does not exceed 1-2% of its total amount in the salt. Based on the EPR spectra of irradiated salts the concentrations of paramagnetic centres are calculated. It is shown that during thermal annealing the main portion of tritium in the gas phase is in the form of oxide (HTO, T 2 O). Tritium is separated from lithium fluoroberyllates in the form of hydrogen (HT, T 2 ). The kinetics of tritium oxide isolation from irradiated lithium oxide aluminate, metha- and orthosilicates, lithium sulphate has been studied. The activation energies of tritium oxide separation process are presented. A supposition is made that chemical reaction of the HTO (T 2 O) or HT(T 2 ) or HF(TF) formation is a limiting stage. Clarification of the process stage limiting the rate of tritium recovery will permit to evaluate conditions for the optimum work of lithium material in the blanket, lithium zone to select the lithium element structure and temperature regime of irradiation

  4. Wetting properties of liquid lithium on lithium compounds

    Energy Technology Data Exchange (ETDEWEB)

    Krat, S.A., E-mail: stepan.krat@gmail.com [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States); National Research Nuclear University MEPhI, Moscow (Russian Federation); Popkov, A.S. [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States); National Research Nuclear University MEPhI, Moscow (Russian Federation); Gasparyan, Yu. M.; Pisarev, A.A. [National Research Nuclear University MEPhI, Moscow (Russian Federation); Fiflis, Peter; Szott, Matthew; Christenson, Michael; Kalathiparambil, Kishor; Ruzic, David N. [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States)

    2017-04-15

    Highlights: • Contact angles of liquid lithium and Li{sub 3}N, Li{sub 2}O, Li{sub 2}CO{sub 3} were measured. • Liquid lithium wets lithium compounds at relatively low temperatures: Li{sub 3}N at 257 °C, Li{sub 2}O at 259 °C, Li{sub 2}CO{sub 3} at 323 °C. • Li wets Li{sub 2}O and Li{sub 3}N better than previously measured fusion-relevant materials (W, Mo, Ta, TZM, stainless steel). • Li wets Li{sub 2}CO{sub 3} better than most previously measured fusion-relevant materials (W, Mo, Ta). - Abstract: Liquid metal plasma facing components (LMPFC) have shown a potential to supplant solid plasma facing components materials in the high heat flux regions of magnetic confinement fusion reactors due to the reduction or elimination of concerns over melting, wall damage, and erosion. To design a workable LMPFC, one must understand how liquid metal interacts with solid underlying structures. Wetting is an important factor in such interaction, several designs of LMPFC require liquid metal to wet the underlying solid structures. The wetting of lithium compounds (lithium nitride, oxide, and carbonate) by 200 °C liquid lithium at various surface temperature from 230 to 330 °C was studied by means of contact angle measurements. Wetting temperatures, defined as the temperature above which the contact angle is less than 90°, were measured. The wetting temperature was 257 °C for nitride, 259 °C for oxide, and 323 °C for carbonate. Surface tensions of solid lithium compounds were calculated from the contact angle measurements.

  5. 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

  6. 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)

  7. Compatibility of yttria (Y{sub 2}O{sub 3}) with liquid lithium

    Energy Technology Data Exchange (ETDEWEB)

    Mitsuyama, Takaaki; Yoneoka, Toshiaki; Terai, Takayuki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    Compatibility of Y{sub 2}O{sub 3} sintered specimens with liquid lithium was tested at 773K. No configuration change was observed with a slight increase of thickness for 1419 hr. Lithium-yttrium complex oxide (LiYO{sub 2}) was formed on the surface, and the inner part changed to gray or black nonstoichiometric Y{sub 2}O{sub 3-X} with lower electrical resistibility. It is concluded that Y{sub 2}O{sub 3} has a possibility as a ceramic coating material for liquid blankets if it can be made into a dense coating on the surface of piping materials. (author)

  8. Lithium Blanket Module dosimetry measurements at the LOTUS 14-MeV neutron source facility

    International Nuclear Information System (INIS)

    Tsang, F.Y.; Leo, W.R.; Sahraoui, C.; Wuthrich, S.; Harker, Y.D.

    1986-01-01

    This paper describes the measurements and results of the dosimeter material reaction rates inside the Lithium Blanket Module (LBM) after irradiation by the LOTUS 14-MeV neutron source at the Ecole Polytechnique Federale de Lausanne. The measurement program has been designed to utilize sets of passive dosimeter materials in the form of foils and wires. The dosimetry materials reaction thresholds and interaction response ranges chosen for this series of measurements encompass the entire neutron spectra along the full length of the LBM fuel rods

  9. Heat transfer in the lithium-cooled blanket of a pulsed fusion reactor

    International Nuclear Information System (INIS)

    Cort, G.E.; Krakowski, R.A.

    1978-01-01

    The transient temperature distribution in the lithium-cooled blanket of a pulsed fusion reactor has been calculated using a finite-element heat-conduction computer program. An auxiliary program was used to predict the coolant transient velocity in a network of parallel and series flow passages with constant driving pressure and varying magnetic field. The coolant velocity was calculated by a Runge-Kutta numerical integration of the conservation equations. The lithium coolant was part of the finite-element heat-conduction mesh with the velocity terms included in the total matrix. The matrix was solved implicitly at each time step for the nodal point temperatures. Slug flow was assumed in the coolant passages and the Boussinesq analogy was used to calculate turbulent heat transfer when the magnetic field was not present

  10. Tritium recovery from fusion blankets using solid lithium compounds. I. Design and minimization of tritium inventory

    International Nuclear Information System (INIS)

    Powell, J.R.

    1975-01-01

    Tritium blanket inventories of 100 curies/MW(e) are readily achievable, and inventories as low as 10 curies/MW(e) are possible for blankets with small lithium compound particulates (less than or equal to 50μ) at T greater than or equal to 800 0 C. Of the three release modes [A - to the main coolant (e.g., He) stream; B - to a separate processing circuit; and C - to the plasma region], mode A appears optimum for blankets using gas-cooled metallic structures (e.g., Al, stainless), while mode C appears optimum for high temperature refractory (e.g., C, SiC) structures. The greater structural complexity of mode B makes it less attractive than modes A and C. No recovery method is required for mode C release. With mode A release, tritium inventory in the coolant circuits ranges from 1 to 10 curies/MW(e), depending on processing parameters. Tritium leak rates to the environment during normal operation can be kept to less than or equal to 10 -3 curies/MW(e) per day with low permeability barriers. In general, a mixture of T 2 and T 2 O is present in the coolant stream. Three methods of tritium recovery are examined: (1) Conversion to T 2 followed by absorption in a metal hydride bed. (2) Conversion to T 2 followed by condensation at approximately 6 0 K. (3) Conversion to T 2 O followed by condensation at approximately 100 0 K

  11. Design of liquid lithium pumps for FMIT

    International Nuclear Information System (INIS)

    Adkins, H.E.

    1983-01-01

    In the Fusion Materials Irradiation Test (FMIT) facility, a jet of liquid lithium is bombarded by accelerated deuterons to generate high energy neutrons for materials testing. The lithium system will include two electromagnetic pumps, a 750 gpm main pump and a 10 gpm auxiliary pump. The larger pump was designed and built in 1982, following extensive testing of a similar pump in the Experimental Lithium System. Design of the auxiliary pump has been completed, but fabrication has not started. This paper discusses the design considerations leading to selection of the Annular Linear Induction Pump (ALIP) concept for these applications. Design parameters, fabrication procedures, and results of pump testing are also reviewed

  12. High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

    International Nuclear Information System (INIS)

    Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G.; Capece, A.; Koel, B.; Roszell, J.; Biewer, T. M.; Gray, T. K.; Kubota, S.; Beiersdorfer, P.

    2015-01-01

    The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started

  13. High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Capece, A.; Koel, B.; Roszell, J. [Princeton University, Princeton, New Jersey 08544 (United States); Biewer, T. M.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Kubota, S. [University of California at Los Angeles, Los Angeles, California 90095 (United States); Beiersdorfer, P. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

    2015-05-15

    The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

  14. Progress in design and development of series liquid lithium-lead expeirmental loops in China

    International Nuclear Information System (INIS)

    Wu Yican; Huang Qunying; Zhu Zhiqiang; Gao Sheng; Song Yong; Li Chunjing; Peng Lei; Liu Shaojun; Wu qingsheng; Liu Songlin; Chen Hongli; Bai Yunqing; Jin Ming; Lv Ruojun; Wang Weihua; Guo Zhihui; Chen Yaping; Ling Xinzhen; Zhang Maolian; Wang Yongliang; Wu Zhaoyang; Wang Hongyan

    2009-01-01

    Liquid LiPb (lithium-lead) experimental loops are the important platforms to investigate the key technologies of liquid LiPb breeder blankets for fusion reactors. Based on the development strategy for liquid LiPb breeder blankets, the technologies development of liquid LiPb experimental loops have been explored by the FDS Team for years, and a series of LiPb experimental loops named DRAGON have been designed and developed, which have independence intellectual property and multi-functional parameters. In this paper, the development route suggestion of Chinese LiPb experimental loops was elaborated, and some information for the senes experimental loops were introduced, such as the design principles, structural features, functions and related experimental researches, etc. (authors)

  15. 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

  16. New Monte Carlo results for the TFTR/Lithium Blanket Module system

    International Nuclear Information System (INIS)

    Engholm, B.A.

    1985-01-01

    Neutronics analysis results from Phase II of the TFTR Lithium Blanket Module (LBM) program are reported. Principal activities were analyses of new coverplate and protective plate designs; updating of the MCNP Monte Carlo model of TFTR/LBM; and performing new reference calculations for D-D and D-T plasmas. The new protective plate was found to reduce LBM responses by 20%. Updating the model included a new tally structure in which the LBM is divided into 92 volume elements corresponding to foil locations. A new version of the MCNP surface-source routine was used, along with the latest pointwise cross sections. All flux, tritium and foil responses are stored at NMFECC and are available for comparison with measurements, when the experimental program gets underway

  17. Fluidized-bed design for ICF reactor blankets using solid-lithium compounds

    International Nuclear Information System (INIS)

    Sucov, E.W.; Malick, F.S.; Green, L.; Hall, B.O.

    1983-01-01

    A fluidized-bed concept for blankets of dry or wetted first-wall ICF reactors using solid-lithium compounds is described. The reaction chamber is a right cylinder, 32 m high and 20 m in diameter; the blanket is composed of 36 steel tanks, 32 m high, which carry the sintered Li 2 O particles in the fluidizing helium gas. Each tank has a radial thickness of 2 m which generates a tritium breeding ration (TBR) of 1.27 and absorbs over 98% of the neutron energy; reducing the thickness to 1.2 m produces a TBR of 1.2 and energy absorption of 97% which satisfy the design goals. Calculations of tritium diffusion through the grains and heat removal from the grains showed that neither could be removed by the carrier gas; tritium and heat are therefore removed by removing the grains themselves by varying the helium flow rate. The particles are continuously fed into the bottom of the tanks at 300 0 C and removed at the top at 475 0 C. Tritium and heat extraction are easily and conveniently done outside the reactor

  18. 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

  19. EXOTIC: Development of ceramic tritium breeding materials for fusion reactor blankets. The behaviour of tritium in: lithium aluminate, lithium oxide, lithium silicates, lithium zirconates

    Energy Technology Data Exchange (ETDEWEB)

    Kwast, H [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Stijkel, H [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Muis, R [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Conrad, R [Commission of the European Communities, Petten (Netherlands). Joint Reseach Centre

    1995-12-01

    This report describes the results of six EXOTIC experiments comprising a total of 48 capsules. Samples of the candidate tritium breeding materials LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 4}SiO{sub 4}, Li{sub 6}Zr{sub 2}O{sub 7}, Li{sub 8}ZrO{sub 6}, Li{sub 2}O and Li{sub 2}SiO{sub 3} have been irradiated at different temperature levels and up to a maximum lithium burnup of about 3%. Tritium residence times of the various breeding materials have been determined from temperature transients performed during irradiation. After irradiation the tritium inventory has been determined from small samples of the various materials. From the out-of-pile tritium release experiments activation energies were determined. These activities have been performed at ECN within the framework of the European Fusion Technology Programme on Breeding Blankets. (orig.).

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

    International Nuclear Information System (INIS)

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

    1982-11-01

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

  1. Required momentum, heat, and mass transport experiments for liquid-metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Sze, D.K.; Abdou, M.A.

    1986-01-01

    Through the effects on fluid flow, many aspects of blanket behavior are affected by magnetohydrodynamic (MHD) effects, including pressure drop, heat transfer, mass transfer, and structural behavior. In this paper, a set of experiments is examined that could be performed in order to reduce the uncertainties in the highly related set of issues dealing with momentum, heat, and mass transport under the influence of a strong magnetic field (i.e., magnetic transport phenomena). By improving our basic understanding and by providing direct experimental data on blanket behavior, these experiments will lead to improved designs and an accurate assessment of the attractiveness of liquid-metal blankets

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

    International Nuclear Information System (INIS)

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

    1988-05-01

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

  3. Stable lithium electrodeposition in liquid and nanoporous solid electrolytes

    KAUST Repository

    Lu, Yingying; Tu, Zhengyuan; Archer, Lynden A.

    2014-01-01

    of these metals and their inability to form uniform electrodeposits on surfaces with inevitable defects. We report on electrodeposition of lithium in simple liquid electrolytes and in nanoporous solids infused with liquid electrolytes. We find that simple liquid

  4. Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket

    International Nuclear Information System (INIS)

    Spannagel, G.; Gierszewski, P.

    1991-01-01

    At the Karlsruhe Nuclear Research Center (KfK) a flexible tool is being developed to simulate the dynamics of tritium inventories. This tool can be applied to any tritium handling system, especially to the fuel cycle components of future nuclear fusion devices. This instrument of simulation will be validated in equipment to be operated at the Karlsruhe Tritium Laboratory. In this study tritium inventories in a NET/ITER type fuel cycle involving a lithium salt solution blanket are investigated. The salt solution blanket serves as an example because it offers technological properties which are attractive in modeling the process; the example does not impair the general validity of the tool. Usually, the operation strategy of complex structures will deteriorate due to failures of the subsystems involved. These failures together with the reduced availability ensuing from them will be simulated. The example of this study is restricted to reduced availabilities of two subsystems, namely the reactor and the tritium recovery system. For these subsystems the influence of statistically varying intervals of operation is considered. Strategies we selected which are representative of expected modes of operation. In the design of a fuel cycle, care will be taken that prescribed availabilities of the subsystems can be achieved; however, the description of reactor operation is a complex task since operation breaks down into several campaigns for which rules have been specified which enable determination of whether a campaign has been successful and can be stopped. Thus, it is difficult to predict the overall behavior prior to a simulation which includes stochastic elements. Using the example mentioned above the capabilities of the tool will be illustrated; besides the presentation of results of inventory simulation, the applicability of these data will be discussed. (orig.)

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

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

  6. Preliminary experimental study of liquid lithium water interaction

    International Nuclear Information System (INIS)

    You, X.M.; Tong, L.L.; Cao, X.W.

    2015-01-01

    Highlights: • Explosive reaction occurs when lithium temperature is over 300 °C. • The violence of liquid lithium water interaction increases with the initial temperature of liquid lithium. • The interaction is suppressed when the initial water temperature is above 70 °C. • Steam explosion is not ignorable in the risk assessment of liquid lithium water interaction. • Explosion strength of liquid lithium water interaction is evaluated by explosive yield. - Abstract: Liquid lithium is the best candidate for a material with low Z and low activation, and is one of the important choices for plasma facing materials in magnetic fusion devices. However, liquid lithium reacts violently with water under the conditions of loss of coolant accidents. The release of large heats and hydrogen could result in the dramatic increase of temperature and pressure. The lithium–water explosion has large effect on the safety of fusion devices, which is an important content for the safety assessment of fusion devices. As a preliminary investigation of liquid lithium water interaction, the test facility has been built and experiments have been conducted under different conditions. The initial temperature of lithium droplet ranged from 200 °C to 600 °C and water temperature was varied between 20 °C and 90 °C. Lithium droplets were released into the test section with excess water. The shape of lithium droplet and steam generated around the lithium were observed by the high speed camera. At the same time, the pressure and temperature in the test section were recorded during the violent interactions. The preliminary experimental results indicate that the initial temperature of lithium and water has an effect on the violence of liquid lithium water interaction.

  7. Preliminary experimental study of liquid lithium water interaction

    Energy Technology Data Exchange (ETDEWEB)

    You, X.M.; Tong, L.L.; Cao, X.W., E-mail: caoxuewu@sjtu.edu.cn

    2015-10-15

    Highlights: • Explosive reaction occurs when lithium temperature is over 300 °C. • The violence of liquid lithium water interaction increases with the initial temperature of liquid lithium. • The interaction is suppressed when the initial water temperature is above 70 °C. • Steam explosion is not ignorable in the risk assessment of liquid lithium water interaction. • Explosion strength of liquid lithium water interaction is evaluated by explosive yield. - Abstract: Liquid lithium is the best candidate for a material with low Z and low activation, and is one of the important choices for plasma facing materials in magnetic fusion devices. However, liquid lithium reacts violently with water under the conditions of loss of coolant accidents. The release of large heats and hydrogen could result in the dramatic increase of temperature and pressure. The lithium–water explosion has large effect on the safety of fusion devices, which is an important content for the safety assessment of fusion devices. As a preliminary investigation of liquid lithium water interaction, the test facility has been built and experiments have been conducted under different conditions. The initial temperature of lithium droplet ranged from 200 °C to 600 °C and water temperature was varied between 20 °C and 90 °C. Lithium droplets were released into the test section with excess water. The shape of lithium droplet and steam generated around the lithium were observed by the high speed camera. At the same time, the pressure and temperature in the test section were recorded during the violent interactions. The preliminary experimental results indicate that the initial temperature of lithium and water has an effect on the violence of liquid lithium water interaction.

  8. Liquid metal flows in insulating elements of self-cooled blankets

    International Nuclear Information System (INIS)

    Molokov, S.

    1995-01-01

    Liquid metal flows in insulating rectangular ducts in strong magnetic fields are considered with reference to poloidal concepts of self-cooled blankets. Although the major part of the flow in poloidal blanket concepts is close to being fully developed, manifolds, expansions, contractions, elbows, etc., which are necessary elements in blanket designs, cause three-dimensional effects. The present investigation demonstrates the flow pattern in basic insulating geometries for actual and more advanced liquid metal blanket concepts and discusses the ways to avoid pressure losses caused by flow redistribution. Flows in several geometries, such as symmetric and non-symmetric 180 turns with and without manifolds, sharp and linear expansions with and without manifolds, etc., have been considered. They demonstrate the attractiveness of poloidal concepts of liquid metal blankets, since they guarantee uniform conditions for heat transfer. If changes in the duct cross-section occur in the plane perpendicular to the magnetic field (ideally a coolant should always flow in the radial-poloidal plane), the disturbances are local and the slug velocity profile is reached roughly at a distance equivalent to one duct width from the manifolds, expansions, etc. The effects of inertia in these flows are unimportant for the determination of the pressure drop and velocity profiles in the core of the flow but may favour heat transfer characteristics via instabilities and strongly anisotropic turbulence. (orig.)

  9. Electrochemical Model for Ionic Liquid Electrolytes in Lithium Batteries

    International Nuclear Information System (INIS)

    Yoo, Kisoo; Deshpande, Anirudh; Banerjee, Soumik; Dutta, Prashanta

    2015-01-01

    ABSTRACT: Room temperature ionic liquids are considered as potential electrolytes for high performance and safe lithium batteries due to their very low vapor pressure and relatively wide electrochemical and thermal stability windows. Unlike organic electrolytes, ionic liquid electrolytes are molten salts at room temperature with dissociated cations and anions. These dissociated ions interfere with the transport of lithium ions in lithium battery. In this study, a mathematical model is developed for transport of ionic components to study the performance of ionic liquid based lithium batteries. The mathematical model is based on a univalent ternary electrolyte frequently encountered in ionic liquid electrolytes of lithium batteries. Owing to the very high concentration of components in ionic liquid, the transport of lithium ions is described by the mutual diffusion phenomena using Maxwell-Stefan diffusivities, which are obtained from atomistic simulation. The model is employed to study a lithium-ion battery where the electrolyte comprises ionic liquid with mppy + (N-methyl-N-propyl pyrrolidinium) cation and TFSI − (bis trifluoromethanesulfonyl imide) anion. For a moderate value of reaction rate constant, the electric performance results predicted by the model are in good agreement with experimental data. We also studied the effect of porosity and thickness of separator on the performance of lithium-ion battery using this model. Numerical results indicate that low rate of lithium ion transport causes lithium depleted zone in the porous cathode regions as the porosity decreases or the length of the separator increases. The lithium depleted region is responsible for lower specific capacity in lithium-ion cells. The model presented in this study can be used for design of optimal ionic liquid electrolytes for lithium-ion and lithium-air batteries

  10. 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.)

  11. Numeric implementation of a nucleation, growth and transport model for helium bubbles in lead-lithium HCLL breeding blanket channels: Theory and code development

    Energy Technology Data Exchange (ETDEWEB)

    Batet, L., E-mail: lluis.batet@upc.edu [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Physics and Nuclear Engineering (DFEN), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Fradera, J. [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Physics and Nuclear Engineering (DFEN), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Valls, E. Mas de les [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Heat Engines (DMMT), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Sedano, L.A. [EURATOM-CIEMAT Association, Fusion Technology Division, Av. Complutense 22, 28040 Madrid (Spain)

    2011-06-15

    Large helium (He) production rates in liquid metal breeding blankets of a DT fusion reactor might have a significant influence in the system design. Low He solubility together with high local concentrations may create the conditions for He cavitation, which would have an impact in the components performance. The paper states that such a possibility is not remote in a helium cooled lithium-lead breeding blanket design. A model based on the Classical Nucleation Theory (CNT) has been developed and implemented in order to have a specific tool able to simulate HCLL systems and identify the key parameters and sensitivities. The nucleation and growth model has been implemented in the open source CFD code OpenFOAM so that transport of dissolved atomic He and nucleated He bubbles can be simulated. At the current level of development it is assumed that void fraction is small enough not to affect either the hydrodynamics or the properties of the liquid metal; thus, bubbles can be represented by means of a passive scalar. He growth and transport has been implemented using the mean radius approach in order to save computational time. Limitations and capabilities of the model are shown by means of zero-dimensional simulation and sensitivity analysis under HCLL breeding unit conditions.

  12. Neutronic investigations on the application of lithium aluminates in the tritium breeding blanket of future fusion reactors

    International Nuclear Information System (INIS)

    Mohsin, A.

    1981-02-01

    A survey is given about the state of development work at the blanket. It shows that present designs aim at a fusion reactor with low tritium inventory. This aim can be achieved with a solid blanket. In this paper this concept is described and the selection of appropriate materials for the solid blanket is discussed. The lithium aluminates turned out to be the most suitable materials. Comparing the different lithium aluminates the compounds Li 5 AlO 4 and LiAlO 2 proved to be the most favourable. The improvement of the breeding ratio when using lead as neutron multiplier was investigated. Employing, for example, a lead zone of 15 cm thickness in front of a 60 cm thick breeding zone, the tritium breeding ratio is raised to 1.65 for Li 5 Al 4 and to 1.48 for LiAlO 2 - The originally higher breeding ratio of the Li 5 AlO 4 in contrary to the LiAlO 2 is compensated hereby. By this LiAlO 2 becomes a very interesting material for a solid blanket since it furthermore exhibits a higher melting point and higher phase transition temperature. For experimental check of the nuclear data of this material and the computational techniques used, a test model was designed and built. This blanket model was used for measuring the space-dependent tritium production rate, which could be compared to corresponding computations. The assembly was made of a lead zone as neutron multiplier, LiAlO 2 as breeding material, and polyethylene as neutron reflector. (orig.) [de

  13. Optimization of the first wall for the DEMO water cooled lithium lead blanket

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, Julien, E-mail: julien.aubert@cea.fr [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Aiello, Giacomo [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Bachmann, Christian [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Di Maio, Pietro Alessandro [Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, Rosario [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy); Li Puma, Antonella; Morin, Alexandre [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Tincani, Amelia [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy)

    2015-10-15

    Highlights: • This paper presents the optimization of the first wall of the water cooled lithium lead DEMO blanket with pressurized water reactor condition and circular channels in order to find the best geometry that can allow the maximum heat flux considering design criteria since an estimate of the engineering limit of the first wall heat load capacity is an essential input for the decision to implement limiters in DEMO. • An optimization study was carried out for the flat first wall design of the DEMO Water-Cooled Lithium Lead considering thermal and mechanical constraint functions, assuming T{sub inlet}/T{sub outlet} equal to 285 °C/325 °C, based on geometric design parameters. • It became clear that through the optimization the advantages of a waved First Wall are diminished. • The analysis shows that the maximum heat load could achieve 2.53 MW m{sup −2}, but considering assumptions such as a coolant velocity ≤8 m/s, pipe diameter ≥5 mm and a total first wall thickness ≤22 mm, heat flux is limited to 1.57 MW m{sup −2}. - Abstract: The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0 MW m{sup −2} with respect to the Eurofer temperature limit. An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet

  14. Effect of electromagnetic coupling on MHD flow in the manifold of fusion liquid metal blanket

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Meng, Zi; Feng, Jingchao; He, Qingyun

    2014-10-15

    In fusion liquid metal (LM) blanket, magnetohydrodynamics (MHD) effects will dominate the flow patterns and the heat transfer characteristics of the liquid metal flow. Manifold is a key component in LM blanket in charge of distributing or collecting the liquid metal coolant. In this region, the complex three dimensional MHD phenomena will be occurred, and the velocity, pressure and flow rate distributions may be dramatically influenced. One important aspect is the electromagnetic coupling effect resulting from an exchange of electric currents between two neighboring fluid domains that can lead to modifications of flow distribution and pressure drop compared to that in electrical separated channels. Understanding the electromagnetic coupling effect in manifold is necessary to optimize the liquid metal blanket design. In this work, a numerical study was carried out to investigate the effect of electromagnetic coupling on MHD flow in a manifold region. The typical manifold geometry in LM blanket was considered, a rectangular supply duct entering a rectangular expansion area, finally feeding into 3 rectangular parallel channels. This paper investigated the effect of electromagnetic coupling on MHD flow in a manifold region. Different electromagnetic coupling modes with different combinations of electrical conductivity of walls were studied numerically. The flow distribution and pressure drop of these modes have been evaluated.

  15. Heat deposition, damage, and tritium breeding characteristics in thick liquid wall blanket concepts

    International Nuclear Information System (INIS)

    Youssef, M.Z.; Abdou, M.A.

    2000-01-01

    The advanced power extraction (APEX) study aims at exploring new and innovative blanket concepts that can efficiently extract power from fusion devices with high neutron wall load. Among the concepts under investigation is the free liquid FW/liquid blanket concept in which a fast flowing liquid FW (∼2-3 cm) is followed by thick flowing blanket (B) of ∼40-50 cm thickness with minimal amount of structure. The liquid FW/B are contained inside the vacuum vessel (VV) with a shielding zone (S) located either behind the VV and outside the vacuum boundary (case A) or placed after the FW/B and inside the VV (case B). In this paper we investigate the nuclear characteristics of this concept in terms of: (1) attenuation capability of the liquid FW/B/S and protection of the VV and magnet against radiation damage; (2) profiles of tritium production rate and tritium breeding ratio (TBR) for several liquid candidates; and (3) profiles of heat deposition rate and power multiplication. The candidate liquid breeders considered are Li, Flibe, Li-Sn, and Li-Pb. Parameters varied are (1) FW/B thickness, L, (2) Li-6 enrichment and (3) thickness of the shield

  16. Corrosion resistance investigation of vanadium alloys in liquid lithium

    Energy Technology Data Exchange (ETDEWEB)

    Borovitskaya, I. V., E-mail: symp@imet.ac.ru [Russian Academy of Sciences, Baikov Institute of Metallurgy and Materials Science (Russian Federation); Lyublinskiy, I. E. [JSC Red Star (Russian Federation); Bondarenko, G. G. [National Research University Higher School of Economics (Russian Federation); Paramonova, V. V. [Russian Academy of Sciences, Baikov Institute of Metallurgy and Materials Science (Russian Federation); Korshunov, S. N.; Mansurova, A. N. [National Research Center Kurchatov Institute (Russian Federation); Lyakhovitskiy, M. M. [Russian Academy of Sciences, Baikov Institute of Metallurgy and Materials Science (Russian Federation); Zharkov, M. Yu. [JSC Red Star (Russian Federation)

    2016-12-15

    A major concern in using vanadium alloys for first wall/blanket systems in fusion reactors is their activity with regard to nonmetallic impurities in the coolants. This paper presents the results of studying the corrosion resistance in high-purity liquid lithium (with the nitrogen and carbon content of less than 10{sup –3} wt %) of vanadium and vanadium alloys (V–1.86Ga, V–3.4Ga–0.62Si, V–4.81Ti–4.82Cr) both in the initial state and preliminarily irradiated with Ar+ ions with energy of 20 keV to a dose of 10{sup 22} m{sup –2} at an irradiation temperature of ~400°C. The degree of corrosion was estimated by measuring the changes in the weight and microhardness. Corrosion tests were carried out under static isothermal conditions at a temperature of 600°C for 400 h. The identity of corrosion mechanisms of materials both irradiated with Ar ions and not irradiated, which consisted in an insignificant penetration of nitrogen into the materials and a substantial escape of oxygen from the materials, causing the formation of a zone with a reduced microhardness near the surface, was established. The influence of the corrosive action of lithium on the surface morphology of the materials under study was found, resulting in the manifestation of grain boundaries and slip lines on the sample surface, the latter being most clearly observed in the case of preliminary irradiation with Ar ions.

  17. Test module in NET for a self-cooled liquid metal blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Arheidt, K.; Fischer, U.

    1989-01-01

    The application of a self-cooled liquid metal blanket concept to the condition of a DEMO-reactor and its testing in NET is described. The neutronics analysis shows that tritium self-sufficiency can be achieved without beryllium multiplier if breeding blankets are arranged at both outboard and inboard side of the torus or, using beryllium as multiplier, with outboard breeding only. First estimates indicate that it should be possible to test all relevant features of the concept in one of the horizontal plug positions of NET. (author). 6 refs.; 7 figs.; 1 tab

  18. NSTX plasma operation with a Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Ellis, R.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Maingi, R.; McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M.; Paul, S.F. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); and others

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer NSTX 2010 experiments tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium molybdenum divertor surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. Black-Right-Pointing-Pointer Noteworthy improvements in plasma performance with the plasma strike point on the liquid lithium molybdenum divertor were obtained similar to those obtained previously with lithiated graphite. The role of lithium impurities in this result is discussed. Black-Right-Pointing-Pointer Inspection of the liquid lithium molybdenum divertor after the Campaign indicated mechanical damage to supports, and other hardware resulting from forces following plasma current disruptions. - Abstract: NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the

  19. Conceptual design of two helium cooled fusion blankets (ceramic and liquid breeder) for INTOR

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Taczanowski, S.

    1983-08-01

    Neutronic and heat transfer calculations have been performed for two helium cooled blankets for the INTOR design. The neutronic calculations show that the local tritium breeding ratios, both for the ceramic blanket (Li 2 SiO 3 ) and for the liquid blanket (Li 17 Pb 83 ) solutions, are 1.34 for natural tritium and about 1.45 using 30% Li 6 enrichment. The heat transfer calculations show that it is possible to cool the divertor section of the torus (heat flux = 1.7 MW/m 2 ) with helium with an inlet pressure of 52 bar and an inlet temperature of 40 0 C. The temperature of the back face of the divertor can be kept at 130 0 C. With helium with the same inlet conditions it is possible to cool the first wall as well (heat flux = 0.136 MW/m 2 ) and keep the back-face of this wall at a temperature of 120 0 C. For the ceramic blanket we use helium with 52 bar inlet pressure and 400 0 C inlet temperature to ensure sufficiently high temperatures in the breeder material. The maximum temperature in the pressure tubes containing the blanket is 450 0 C, while the maximum breeder particle temperature is 476 0 C. (orig./RW) [de

  20. 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)

  1. Response distributions of 6LiF and 7LiF thermoluminescence dosimeters in lithium blanket assemblies

    International Nuclear Information System (INIS)

    Maekawa, Hiroshi; Kusano, Jyo-ichi; Seki, Yasushi

    1976-11-01

    Measurement of the radiation-heating rate distribution in the fusion blanket is as important as measurement of the fission-rate distribution in a fission reactor. To obtain the information of radiation heating, the response (integral glow value) distributions in pseudo-spherical lithium assemblies with and without a graphite reflector were measured with 6 LiF and 7 LiF TLD's. The measured responses are normalized to values per source neutron. Experimental error is about 35%, and the error in positions of TLD's is about +- 3 mm. The experimental results are compared with those of calculation using RADHEAT code system and ENDF/B-III data file. (auth.)

  2. Lithium conducting ionic liquids based on lithium borate salts

    Energy Technology Data Exchange (ETDEWEB)

    Zygadlo-Monikowska, E.; Florjanczyk, Z.; Sluzewska, K.; Ostrowska, J.; Langwald, N.; Tomaszewska, A. [Warsaw University of Technology, Faculty of Chemistry, ul. Noakowskiego 3, 00-664 Warsaw (Poland)

    2010-09-15

    The simple reaction of trialkoxyborates with butyllithium resulted in the obtaining of new lithium borate salts: Li{l_brace}[CH{sub 3}(OCH{sub 2}CH{sub 2}){sub n}O]{sub 3}BC{sub 4}H{sub 9}{r_brace}, containing oxyethylene substituents (EO) of n=1, 2, 3 and 7. Salts of n {>=} 2 show properties of room temperature ionic liquid (RTIL) of low glass transition temperature, T{sub g} of the order from -70 to -80 C. The ionic conductivity of the salts depends on the number of EO units, the highest conductivity is shown by the salt with n = 3; in bulk its ambient temperature conductivity is 2 x 10{sup -5} S cm{sup -1} and in solution in cyclic propylene sulfite or EC/PC mixture, conductivity increases by an order of magnitude. Solid polymer electrolytes with borate salts over a wide concentration range, from 10 to 90 mol.% were obtained and characterized. Three types of polymeric matrices: poly(ethylene oxide) (PEO), poly(trimethylene carbonate) (PTMC) and two copolymers of acrylonitrile and butyl acrylate p(AN-BuA) were used in them as polymer matrices. It has been found that for systems of low salt concentration (10 mol.%) the best conducting properties were shown by solid polymer electrolytes with PEO, whereas for systems of high salt concentration, of the polymer-in-salt type, good results were achieved for PTMC as polymer matrix. (author)

  3. Operational Characteristics of Liquid Lithium Divertor in NSTX

    Science.gov (United States)

    Kaita, R.; Kugel, H.; Abrams, T.; Bell, M. G.; Bell, R. E.; Gerhardt, S.; Jaworski, M. A.; Kallman, J.; Leblanc, B.; Mansfield, D.; Mueller, D.; Paul, S.; Roquemore, A. L.; Scotti, F.; Skinner, C. H.; Timberlake, J.; Zakharov, L.; Maingi, R.; Nygren, R.; Raman, R.; Sabbagh, S.; Soukhanovskii, V.

    2010-11-01

    Lithium coatings on plasma-facing components (PFC's) have resulted in improved plasma performance on NSTX in deuterium H-mode plasmas with neutral beam heating.^ Salient results included improved electron confinement and ELM suppression. In CDX-U, the use of lithium-coated PFC's and a large-area liquid lithium limiter resulted in a six-fold increase in global energy confinement time. A Liquid Lithium Divertor (LLD) has been installed in NSTX for the 2010 run campaign. The LLD PFC consists of a thin film of lithium on a temperature-controlled substrate to keep the lithium liquefied between shots, and handle heat loads during plasmas. This capability was demonstrated when the LLD withstood a strike point on its surface during discharges with up to 4 MW of neutral beam heating.

  4. Control of nitrogen concentration in liquid lithium by iron-titanium alloy

    International Nuclear Information System (INIS)

    Hirakane, Shinji; Yoneoka, Toshiaki; Tanaka, Satoru

    2006-01-01

    Reducing the nitrogen concentration in liquid lithium is one of the most important steps in creating a liquid lithium blanket system. In this study, in order to verify the nitrogen gettering performance of Fe-Ti alloy, the variation in the nitrogen concentration in liquid lithium, into which Fe-10 at.% Ti or Fe-5 at.% Ti getter was immersed, was examined. The results confirmed a gettering performance of Fe-Ti alloy comparable to that of V-Ti alloy, although the effects were not durable in either the Fe-Ti or the V-Ti alloy. After the immersion test, the existing states of nitrogen absorbed in the gettering material were analyzed by means of XRD, XMA and XPS. TiN and some nitrogen dissolved in α-Fe without forming TiN were observed. It was indicated that nitrogen gettering is prevented not only by the surface nitrides, but also by the internal diffusion barriers originating from the absorbed nitrogen

  5. Magnetohydrodynamic (MHD) considerations for liquid metal blanket and a SiC/SiC composite structure

    International Nuclear Information System (INIS)

    Scholz, R.; Greeff, J. de; Vinche, C.

    1998-01-01

    The electrical conductivity was measured on SiC/SiC composite specimens, in the as-received conditions and after neutron irradiation, for temperatures between 20 deg. C and 1000 deg. C. The tests were aimed at estimating the magnitude of MHD effects in liquid metal blankets and a SiC/SiC composites structure. The electrical conductivity of the unirradiated samples increased continuously with temperature and ranged from 330 (Ω m) -1 at 20 deg. C to 550 (Ω m) -1 at 1000 deg.C. The irradiation reduced only slightly the magnitude of σ indicating the materials tested cannot be treated as an electrical insulator in a MHD analysis for liquid metal blankets. (authors)

  6. Magnetohydrodynamic (MHD) considerations for liquid metal blanket and a SiC/SiC composite structure

    Energy Technology Data Exchange (ETDEWEB)

    Scholz, R.; Greeff, J. de; Vinche, C. [Commission Europeenne Community, JRC, Vatican City State, Holy See (Italy)

    1998-07-01

    The electrical conductivity was measured on SiC/SiC composite specimens, in the as-received conditions and after neutron irradiation, for temperatures between 20 deg. C and 1000 deg. C. The tests were aimed at estimating the magnitude of MHD effects in liquid metal blankets and a SiC/SiC composites structure. The electrical conductivity of the unirradiated samples increased continuously with temperature and ranged from 330 ({omega} m){sup -1} at 20 deg. C to 550 ({omega} m){sup -1} at 1000 deg.C. The irradiation reduced only slightly the magnitude of {sigma} indicating the materials tested cannot be treated as an electrical insulator in a MHD analysis for liquid metal blankets. (authors)

  7. Progress in design and analysis of the net water cooled liquid breeder blanket

    International Nuclear Information System (INIS)

    Danner, W.; Rieger, M.; Verschuur, K.A.; Vieider, G.; Casini, G.; Chazalon, M.; Libin, B.; Farfaletti-Casali, F.; Piana, R.

    1987-01-01

    The NET liquid breeder blanket was subjected to a major design revision and integrated in the new NET-DN machine configuration. In this paper briefly the most essential design features are summarized and some results from thermohydraulics and 1D as well as 3D neutronics analyses are presented. It is concluded that the performance meets well the requirements of NET but that the concept needs substantial improvement if applied to a reactor

  8. Control of nitrogen concentration in liquid lithium by hot trapping

    International Nuclear Information System (INIS)

    Sakurai, Toshiharu; Yoneoka, Toshiaki; Tanaka, Satoru; Suzuki, Akihiro; Muroga, Takeo

    2002-01-01

    Nitrogen concentration in liquid lithium was controlled by the method of hot trapping. V-Ti alloy and chromium were used as nitrogen gettering materials. Chromium is known to form ternary nitride with lithium. Gettering experiments were conducted at 823 K for 0.8-2.2 Ms. Under high nitrogen concentration in liquid lithium, above 10 -2 mass%, nitrogen gettering effect of chromium was found to be larger than that of V-10at.% Ti alloy. Nitrogen gettering by chromium at 823 K reached a limit at about 6.5x10 -3 mass% of nitrogen concentration in liquid lithium. Instability of ternary nitride of chromium and lithium below this nitrogen concentration in liquid lithium was considered to be the reason for this limit. The composition of the ternary nitride that was formed in this study was considered to be Li 6 Cr(III) 3 N 5 . In addition, immersion experiments of yttrium with V-10at.% Ti alloy were performed. It was found that nitriding of yttrium in liquid lithium is controlled by nitrogen gettering effect of V-10at.% Ti alloy

  9. Heat-pulse flowmeter for a liquid breeder blanket

    International Nuclear Information System (INIS)

    Kondo, Masatoshi; Shibano, Kyohei; Tanaka, Teruya; Muroga, Takeo

    2013-01-01

    Liquid metals Li, Pb-17Li and Sn-20Li are candidate liquid breeders in fusion reactors. The development of a flowmeter that can be applied to high-temperature liquid metals is an important issue. A heat-pulse flowmeter is proposed in the present study. Its basic performance was investigated by means of a loop experiment with Pb-17Li and a numerical simulation. The temperature distribution in flowing Pb-17Li was obtained by local transient heating of the outer surface of a loop tube. The temperature distribution gradually changed and resembled the movement of a hot spot, which had a higher temperature than its surroundings. This hot spot moved along the flow and passed through the tips of the thermocouples. The change in temperature distribution with the movement of the hot spot was monitored by three thermocouples exposed to the Pb-17Li flow. The results of the loop experiments were numerically simulated by assuming a certain flow rate, and the temperature profile obtained in the loop experiment was in agreement with the simulation results. The time taken by the hot spot to pass through the tips of the thermocouples was measured and simulated, and the correlation between this time and the average flow velocity was evaluated. The results indicated the average flow velocity can be obtained using the heat-pulse flowmeter proposed in this study. (author)

  10. Corrosion of ferrous alloys in nitrogen contaminated liquid lithium

    International Nuclear Information System (INIS)

    Olson, D.L.; Bradley, W.L.

    1976-01-01

    Liquid lithium penetration of 304L stainless steel and Armco iron grain boundaries has been studied. The penetration kinetics for the 304L stainless steel was found to be diffusion controlled. The measured temperature dependent delay time has been associated with the initial formation of the corrosion product at the grain boundary. Nitrogen in the stainless steel or the liquid lithium has been found to accelerate the rate of attack without changing the apparent activation energy. Grain boundary grooving of Armco iron in liquid lithium indicates that the controlling mass transport is also through a corrosion product present as a surface film. Stresses as small as 12 MPa have been found to give rise to a fifty-fold increase in the rate of penetration of Armco iron by liquid lithium

  11. Liquid Crystals of Lithium Dodecylbenzenesulfonate for Electric Double Layer Capacitors

    International Nuclear Information System (INIS)

    Kuzmin, Andrey Vasil’evich; Yurtov, Evgeny V.

    2016-01-01

    Ionic lyotropic liquid crystals based on lithium dodecylbenzenesulfonate were used as electrolytes for electric double layer capacitors with carbon fibrous electrodes. The capacitors were tasted by cyclic voltammetry, galvanostatic charge and discharge, and impedance spectroscopy. The highest specific capacitance was achieved for electrical double layer capacitor equipped with ionic lyotropic liquid crystal of lithium dodecylbenzenesulfonate 35 wt% in water. The specific capacitance of capacitor was calculated from galvanostatic discharge curves – 15 F/g of carbon fibrous material

  12. Testing of Liquid Lithium Limiters in CDX-U

    Energy Technology Data Exchange (ETDEWEB)

    R. Majeski; R. Kaita; M. Boaz; P. Efthimion; T. Gray; B. Jones; D. Hoffman; H. Kugel; J. Menard; T. Munsat; A. Post-Zwicker; V. Soukhanovskii; J. Spaleta; G. Taylor; J. Timberlake; R. Woolley; L. Zakharov; M. Finkenthal; D. Stutman; G. Antar; R. Doerner; S. Luckhardt; R. Seraydarian; R. Maingi; M. Maiorano; S. Smith; D. Rodgers

    2004-07-30

    Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described.

  13. Testing of Liquid Lithium Limiters in CDX-U

    International Nuclear Information System (INIS)

    Majeski, R.; Kaita, R.; Boaz, M.; Efthimion, P.; Gray, T.; Jones, B.; Hoffman, D.; Kugel, H.; Menard, J.; Munsat, T.; Post-Zwicker, A.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Maingi, R.; Maiorano, M.; Smith, S.; Rodgers, D.

    2004-01-01

    Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described

  14. Testing of liquid lithium limiters in CDX-U

    International Nuclear Information System (INIS)

    Majeski, R.; Kaita, R.; Boaz, M.; Efthimion, P.; Gray, T.; Jones, B.; Hoffman, D.; Kugel, H.; Menard, J.; Munsat, T.; Post-Zwicker, A.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Maingi, R.; Maiorano, M.; Smith, S.; Rodgers, D.; Soukhanovskii, V.

    2004-01-01

    Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm 2 , subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now been performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments, the liquid lithium plasma-facing area was increased to 2000 cm 2 . Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described

  15. 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

  16. Initial three-dimensional neutronics calculations for the EU water cooled lithium-lead test blanket module for ITER-FEAT

    International Nuclear Information System (INIS)

    Jordanova, J.; Poitevin, Y.; Li Puma, A.; Kirov, N.

    2003-01-01

    The paper summarizes the main results of the initial three-dimensional radiation transport analysis of the EU water-cooled lithium-lead test blanket module performed using the Monte Carlo code MCNP. Estimates of tritium production rate, nuclear energy deposition and cumulative fluence effects such as radiation damage through atomic displacement and production of He and H are presented. (author)

  17. Extraction of tritium from liquid lithium by permeation

    International Nuclear Information System (INIS)

    Alire, R.M.

    1978-01-01

    This paper assesses a method for extracting tritium from liquid lithium for specific application to the conceptual laser fusion reactor that uses a continuous lithium ''waterfall.'' The tritium diffuses through a refractory metal that contains a getter and is then stored in a hydride-forming alloy. There are various uncertainties with this method including helium-4 extraction, unknown impurities that may accumulate in liquid lithium, the effects of these impurities on tritium separation, and the maintenance of tritium-contaminated equipment. Our study indicates that major tritium losses will occur during equipment maintenance rather than as a result of permeation losses through the primary vessel

  18. 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.

  19. Preparation and transport properties of novel lithium ionic liquids

    International Nuclear Information System (INIS)

    Shobukawa, Hitoshi; Tokuda, Hiroyuki; Tabata, Sei-Ichiro; Watanabe, Masayoshi

    2004-01-01

    Novel lithium salts of borates having two electron-withdrawing groups (either 1,1,1,3,3,3-hexafluoro-2-propoxy or pentafluorophenoxy group) and two methoxy-oligo(ethylene oxide) groups (number of repeating unit: n = 3, 4, 7.2) were prepared by successive substitution-reactions from LiBH 4 . The obtained lithium salts were clear and colorless liquids at room temperature. The density, thermal property, viscosity, and ionic conductivity were measured for the lithium ionic liquids. The pulsed-gradient spin-echo NMR (PGSE-NMR) method was used to independently determine self-diffusion coefficients of the lithium cation ( 7 Li NMR) and the anion ( 19 F NMR) in the bulk. The ionic conductivity of the new lithium salts was 10 -5 to 10 -4 S cm -1 at 30 deg. C, which was lower than that of typical ionic liquids by two orders of magnitude. However, the degree of self-dissociation of the lithium ionic liquids; the ratio of the molar conductivity determined by the complex impedance method to that calculated from the self-diffusion coefficients and the Nernst-Einstein equation, ranged from 0.1 to 0.4, which are comparable values to those of a highly dissociable salt in an aprotic polar solvent and of typical ionic liquids. The main reason for the meager conductivity was high viscosities of the lithium ionic liquids. It should be noted that the lithium ionic liquids have self-dissociation ability and conduct the ions in the absence of organic solvents

  20. 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

  1. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  2. Kinetics of liquid lithium reaction with oxygen-nitrogen mixtures

    International Nuclear Information System (INIS)

    Gil, T.K.; Kazimi, M.S.

    1986-01-01

    A series of experiments have been conducted in order to characterize the kinetics of lithium chemical reaction with a mixture of oxygen and nitrogen. Three mixed gas compositions were used; 80% N 2 and 20% O 2 , 90% N 2 and 10% O 2 , and 95% N 2 and 5% O 2 . The reaction rate was obtained as a function of lithium temperature and the oxygen fraction. Liquid lithium temperature varied from 400 to 1100 0 C. By varying the composition, the degree of inhibition of the lithium-nitrogen reaction rate due to the presence of oxygen was observed. The results indicate that the lithium-nitrogen reaction rate depended on both the fraction of oxygen present and lithium temperature. The lithium nitride layer formed from the reaction also had a significant inhibition effect on the lithium-nitrogen reaction rate while the lithium-oxygen reaction rate was not as greatly hindered. LITFIRE, a computer code which simulates temperature and pressure history in a containment building following lithium spills, was modified by including (1) an improved model for the lithium-nitrogen reaction rate and (2) a model for the lithium-CO 2 reaction. LITFIRE was used to simulate HEDL's LC-2 and LA-5 experiments, and the predicted temperatures and pressures were in a reasonable agreement. Furthermore, LITFIRE was applied to a prototypical fusion reactor containment in order to simulate the consequences of a lithium spill accident. The result indicated that if nitrogen was used as containment building gas during the accident, the consequences of the accident would be less severe than those with air. The pressure rise in the building was found to be reduced by 50% and the maximum temperature of the combustion zone was limited to 900 0 C instead of 1200 0 C in the case of air

  3. Plasma interaction with liquid lithium: Measurements of retention and erosion

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, M.J. E-mail: mbaldwin@ferp.ucsd.edu; Doerner, R.P.; Luckhardt, S.C.; Seraydarian, R.; Whyte, D.G.; Conn, R.W

    2002-11-01

    This paper reports on recent studies of high flux deuterium and helium plasma interaction with liquid lithium in the Pisces-B edge plasma simulator facility. Deuterium retention is explored as a function of plasma ion fluence in the range 6x10{sup 19}-4x10{sup 22} atoms cm{sup -2} and exposure temperatures of 523-673 K. The results are consistent with full uptake of the deuterium ions incident on the liquid metal surface, independent of the temperature of the liquid lithium. Full uptake continues until the sample is volumetrically converted to lithium deuteride. Helium retention is not observed for fluences up to 5x10{sup 21} He atoms cm{sup -2}. Measurements of the erosion of lithium are found to be consistent with physical sputtering for the lithium solid phase. However, a mechanism that provides an increased evaporative-like yield and is related to ion impact events on the surface, dominates during the liquid phase leading to an enhanced loss rate for liquid lithium that is greater than the expected loss rate due to evaporation at elevated temperatures. Further, the material loss rate is found to depend linearly on the incident ion flux, even at very high temperature.

  4. Optimization of the breeder zone cooling tubes of the DEMO Water-Cooled Lithium Lead breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A.; Arena, P.; Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Del Nevo, A. [ENEA Brasimone, Camugnano, BO (Italy); Forte, R. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy)

    2016-11-01

    Highlights: • Determination of an optimal configuration for the breeder zone cooling tubes. • Attention has been focused on the toroidal–radial breeder zone cooling tubes lay out. • A theoretical-computational approach based on the Finite Element Method (FEM) has been followed, adopting a qualified commercial FEM code. • Five different configurations have been investigated to optimize the breeder zone cooling tubes arrangement fulfilling all the rules prescribed by safety codes. - Abstract: The determination of an optimal configuration for the breeder zone (BZ) cooling tubes is one of the most important issues in the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket R&D activities, since BZ cooling tubes spatial distribution should ensure an efficient heat power removal from the breeder, avoiding hotspots occurrence in the thermal field. Within the framework of R&D activities supported by the HORIZON 2020 EUROfusion Consortium action on the DEMO WCLL breeding blanket design, a campaign of parametric analyses has been launched at the Department of Energy, Information Engineering and Mathematical Models of the University of Palermo (DEIM), in close cooperation with ENEA-Brasimone, in order to assess the potential influence of BZ cooling tubes number on the thermal performances of the DEMO WCLL outboard breeding blanket equatorial module under the nominal steady state operative conditions envisaged for it, optimizing their geometric configuration and taking also into account that a large number of cooling pipes can deteriorate the tritium breeding performances of the module. In particular, attention has been focused on the toroidal-radial option for the BZ tube bundles lay-out and a parametric study has been carried out taking into account different tube bundles arrangement within the module. The study has been carried out following a numerical approach, based on the finite element method (FEM), and adopting a qualified commercial FEM code. Results

  5. Recent Liquid Lithium Limiter Experiments in CDX-U

    International Nuclear Information System (INIS)

    Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J.; Timberlake, J.; Zakharov, L.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Soukhanovskii, V.; Maingi, R.; Finkenthal, M.; Stutman, D.; Rodgers, D.; Angelini, S.

    2005-01-01

    Recent experiments in the Current Drive eXperiment-Upgrade (CDX-U) provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R=34 cm, a=22 cm, B toroidal = 2 kG, I P =100 kA, T e (0) ∼ 100 eV, n e (0) ∼ 5 x 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium pool limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium pool limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium limited discharges are consistent with Z effective < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced

  6. Liquid lithium limiter experiments in CDX-U

    International Nuclear Information System (INIS)

    Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J.; Timberlake, J.; Zakharov, L.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Soukhanovskii, V.; Maingi, R.; Finkenthal, M.; Stutman, D.; Rodgers, D.

    2005-01-01

    Recent experiments in the Current Drive eXperiment - Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R=34 cm, a=22 cm, B toroidal 2 kG, I P =100 kA, T e (0)∼100 eV, n e (0)∼ 5 x 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium limited discharges are consistent with Z effective <1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced. (author)

  7. Liquid Lithium Limiter Experiments in CDX-U

    International Nuclear Information System (INIS)

    Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J.; Timberlake, J.; Zakharov, L.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Soukhanovskii, V.; Maingi, R.; Finkenthal, M.; Stutman, D.; Rodgers, D.

    2004-01-01

    Recent experiments in the Current Drive Experiment-Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, B toroidal = 2 kG, I P = 100 kA, T e (0) = 100 eV, n e (0) ∼ 5 x 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium-limited discharges are consistent with Z effective < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced

  8. 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.)

  9. Factors influencing the thermodynamic isotope effect of lithium in polyetherlithium liquid-liquid extraction systems

    International Nuclear Information System (INIS)

    Fu Lian; Fang Shengqiang; Yao Zhongqi; Gao Zhichang; Tan Ganzhu

    1989-01-01

    The published data up to now concerning polyether-lithium liquid-liquid extraction systems, can be summarized by the equation, ε p = (α-1)/[1 + 0.46(1-P)], where α denotes the isotope separation factor; P - the ratio of the lithium concentration in the organic phase to the initial concentration of crown ethers; ε p -the enrichment coefficient as P = 100%. Based on the changes in ε p , P, α and D(distribution ratio), the functions of factors such as polyether's structure, polyether's side group, polyether's concentration, organic solvent, negative ion of lithium salt and lithium salt's concentration, are discussed and reported

  10. Liquid Lithium Wall Experiments in CDX-U

    International Nuclear Information System (INIS)

    Doerner, R.; Kaita, R.; Majeski, R.; Luckhardt, S.

    1999-01-01

    The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. Sputtering and erosion tests are currently underway in the PISCES device at the University of California at San Diego (UCSD). To complement this effort, plasma interaction questions in a toroidal plasma geometry will be addressed by a proposed new groundbreaking experiment in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The CDX-U plasma is intensely heated and well diagnosed, and an extensive liquid lithium plasma-facing surface will be used for the first time with a toroidal plasma. Since CDX-U is a small ST, only approximately1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment

  11. Activation analysis and waste management for dual-cooled lithium lead breeder (DLL) blanket of the fusion power reactor FDS-II

    International Nuclear Information System (INIS)

    Chen Mingliang; Huang Qunying; Li Jingjing; Zeng Qin; Wu Yican

    2005-01-01

    The calculation and analysis on the activation levels of the different regions of dual-cooled lithium-lead (DLL) breeder blanket of FDS-II, including afterheat, dose rate, activity and biological hazard potential after shutdown, were carried out with the neutronics code system VisualBUS and multi-group working library HENDL1.0/MG. The safety and environment assessment of fusion power (SEAFP) strategy for the management of activated material is here applied to the DLL blanket, to define the suitable recycling (reuse of activated material) procedure and the possibility of clearance (declassification of the material with low activity level to non-active waste). (authors)

  12. Diagnostics for liquid lithium experiments in CDX-U

    International Nuclear Information System (INIS)

    Kaita, R.; Efthimion, P.; Hoffman, D.; Jones, B.; Kugel, H.; Majeski, R.; Munsat, T.; Raftopoulos, S.; Taylor, G.; Timberlake, J.; Soukhanovskii, V.; Stutman, D.; Iovea, M.; Finkenthal, M.; Doerner, R.; Luckhardt, S.; Maingi, R.; Causey, R.

    2000-01-01

    A flowing liquid lithium first wall or diverter target could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls in fusion reactors. To investigate the interaction of a spherical torus plasma with liquid lithium limiters, large area diverter targets, and walls, discharges will be established in the Current Drive Experiment-Upgrade (CDX-U) where the plasma-wall interactions are dominated by liquid lithium surfaces. Among the unique CDX-U lithium diagnostics is a multi-layer mirror (MLM) array, which will monitor the 135 (angstrom) LiIII line for core lithium concentrations. Additional spectroscopic diagnostics include a grazing incidence XUV spectrometer (STRS) and a filterscope system to monitor D α and various impurity lines local to the lithium limiter. Profile data will be obtained with a multichannel tangential bolometer and a multipoint Thomson scattering system configured to give enhanced edge resolution. Coupons on th e inner wall of the CDX-U vacuum vessel will be used for surface analysis. A 10,000 frame per second fast visible camera and an IR camera will also be available

  13. Mechanical Design of the NSTX Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

    2009-02-19

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  14. Stable lithium electrodeposition in liquid and nanoporous solid electrolytes

    KAUST Repository

    Lu, Yingying

    2014-08-10

    Rechargeable lithium, sodium and aluminium metal-based batteries are among the most versatile platforms for high-energy, cost-effective electrochemical energy storage. Non-uniform metal deposition and dendrite formation on the negative electrode during repeated cycles of charge and discharge are major hurdles to commercialization of energy-storage devices based on each of these chemistries. A long-held view is that unstable electrodeposition is a consequence of inherent characteristics of these metals and their inability to form uniform electrodeposits on surfaces with inevitable defects. We report on electrodeposition of lithium in simple liquid electrolytes and in nanoporous solids infused with liquid electrolytes. We find that simple liquid electrolytes reinforced with halogenated salt blends exhibit stable long-term cycling at room temperature, often with no signs of deposition instabilities over hundreds of cycles of charge and discharge and thousands of operating hours. We rationalize these observations with the help of surface energy data for the electrolyte/lithium interface and impedance analysis of the interface during different stages of cell operation. Our findings provide support for an important recent theoretical prediction that the surface mobility of lithium is significantly enhanced in the presence of lithium halide salts. Our results also show that a high electrolyte modulus is unnecessary for stable electrodeposition of lithium.

  15. Mechanical Design of the NSTX Liquid Lithium Divertor

    International Nuclear Information System (INIS)

    Ellis, R.; Kaita, R.; Kugel, H.; Paluzzi, G.; Viola, M.; Nygren, R.

    2009-01-01

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuum compatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  16. Irradiaiton facilities for testing solid and liquid blanket breeder materials with in-situ tritium release measurements in the HFR Petten

    International Nuclear Information System (INIS)

    Conrad, R.; Debarberis, L.

    1991-01-01

    Lithium-based tritium breeder materials for solid and liquid fusion reactor blanket concepts are being tested in the High Flux Reactor (HFR) Petten with in-situ tritium release measurements since 1985, within the European Fusion Technology Programme and the BEATRIX-I programme. Ceramic breeder materials are being tested in the EXOTIC and COMPLIMENT experimental programmes and the liquid breeder material, Pb-17Li, is being tested in the LIBRETTO experimental programme. The in-pile experiments are performed with irradiation facilities developed by the Joint Research Centre (JRC) Petten. The irradiation vehicles are multi-channel rigs. The sample holders consist of independent, fully instrumented and triple contained capsules. The out-of-pile experimental equipment consist of twelve independent circuits for on-line tritium release and tritium permeation measurements and eight independent circuits for temperature control. The experimental achievements obtained so far contribute to the selection of candidate tritium breeder materials for blanket concepts of near future machines like NET, ITER and DEMO. (orig.)

  17. Ionic Liquid-Nanoparticle Hybrid Electrolytes and their Application in Secondary Lithium-Metal Batteries

    KAUST Repository

    Lu, Yingying; Das, Shyamal K.; Moganty, Surya S.; Archer, Lynden A.

    2012-01-01

    Ionic liquid-tethered nanoparticle hybrid electrolytes comprised of silica nanoparticles densely grafted with imidazolium-based ionic liquid chains are shown to retard lithium dendrite growth in rechargeable batteries with metallic lithium anodes

  18. Interactions of liquid lithium with various atmospheres, concretes, and insulating materials; and filtration of lithium aerosols

    International Nuclear Information System (INIS)

    Jeppson, D.W.

    1979-06-01

    This report describes the facilities and experiments and presents test results of a program being conducted at the hanford Engineering Development Laboratory (HEDL) in support of the fusion reactor development effort. This experimental program is designed to characterize the interaction of liquid lithium with various atmospheres, concretes, and insulating materials. Lithium-atmosphere reaction tests were conducted in normal humidity air, pure nitrogen, and carbon dioxide. These tests are described and their results, such as maximum temperatures, aerosol generated, and reaction rates measured, are reported. Initial lithium temperatures for these tests ranged between 224 0 C and 843 0 C. A lithium-concrete reaction test, using 10 kg of lithium at 327 0 C, and lithium-insulating materials reaction tests, using a few grams of lithium at 350 0 C and 600 0 C, are also described and results are presented. In addition, a lithium-aerosol filter loading test was conducted to determine the mass loading capacity of a commercial high efficiency particulate air (HEPA) filter. The aerosol was characterized, and the loading-capacity-versus-pressure-buildup across the filter is reported

  19. Liquid Lithium Limiter Effects on Tokamak Plasmas and Plasma-Liquid Surface Interactions

    Energy Technology Data Exchange (ETDEWEB)

    R. Kaita; R. Majeski; R. Doerner; G. Antar; M. Baldwin; R. Conn; P. Efthimion; M. Finkenthal; D. Hoffman; B. Jones; S. Krashenninikov; H. Kugel; S. Luckhardt; R. Maingi; J. Menard; T. Munsat; D. Stutman; G. Taylor; J. Timberlake; V. Soukhanovskii; D. Whyte; R. Woolley; L. Zakharov

    2002-10-15

    We present results from the first experiments with a large area liquid lithium limiter in a magnetic fusion device, and its effect on improving plasma performance by reducing particle recycling. Using large area liquid metal surfaces in any major fusion device is unlikely before a test on a smaller scale. This has motivated its demonstration in the CDX-U spherical torus with a unique, fully toroidal lithium limiter. The highest current discharges were obtained with a liquid lithium limiter. There was a reduction in recycling, as indicated by a significant decrease in the deuterium-alpha emission and oxygen radiation. How these results might extrapolate to reactors is suggested in recycling/retention experiments with liquid lithium surfaces under high-flux deuterium and helium plasma bombardment in PISCES-B. Data on deuterium atoms retained in liquid lithium indicate retention of all incident ions until full volumetric conversion to lithium deuteride. The PISCES-B results also show a material loss mechanism that lowers the maximum operating temperature compared to that for the liquid surface equilibrium vapor pressure. This may restrict the lithium temperature in reactors.

  20. Liquid Lithium Limiter Effects on Tokamak Plasmas and Plasma-Liquid Surface Interactions

    International Nuclear Information System (INIS)

    Kaita, R.; Majeski, R.; Doerner, R.; Antar, G.; Baldwin, M.; Conn, R.; Efthimion, P.; Finkenthal, M.; Hoffman, D.; Jones, B.; Krashenninikov, S.; Kugel, H.; Luckhardt, S.; Maingi, R.; Menard, J.; Munsat, T.; Stutman, D.; Taylor, G.; Timberlake, J.; Soukhanovskii, V.; Whyte, D.; Woolley, R.; Zakharov, L.

    2002-01-01

    We present results from the first experiments with a large area liquid lithium limiter in a magnetic fusion device, and its effect on improving plasma performance by reducing particle recycling. Using large area liquid metal surfaces in any major fusion device is unlikely before a test on a smaller scale. This has motivated its demonstration in the CDX-U spherical torus with a unique, fully toroidal lithium limiter. The highest current discharges were obtained with a liquid lithium limiter. There was a reduction in recycling, as indicated by a significant decrease in the deuterium-alpha emission and oxygen radiation. How these results might extrapolate to reactors is suggested in recycling/retention experiments with liquid lithium surfaces under high-flux deuterium and helium plasma bombardment in PISCES-B. Data on deuterium atoms retained in liquid lithium indicate retention of all incident ions until full volumetric conversion to lithium deuteride. The PISCES-B results also show a material loss mechanism that lowers the maximum operating temperature compared to that for the liquid surface equilibrium vapor pressure. This may restrict the lithium temperature in reactors

  1. Liquid lithium limiter effects on tokamak plasmas and plasma-liquid surface interactions

    International Nuclear Information System (INIS)

    Kaita, R.; Majeski, R.; Doerner, R.

    2003-01-01

    We present results from the first experiments with a large area liquid lithium limiter in a magnetic fusion device, and its effect on improving plasma performance by reducing particle recycling. Using large area liquid metal surfaces in any major fusion device is unlikely before a test on a smaller scale. This has motivated its demonstration in the CDX-U spherical torus with a unique, fully toroidal lithium limiter. The highest current discharges were obtained with a liquid lithium limiter. There was a reduction in recycling, as indicated by a significant decrease in the deuterium-alpha emission and oxygen radiation. How these results might extrapolate to reactors is suggested in recycling/retention experiments with liquid lithium surfaces under high-flux deuterium and helium plasma bombardment in PISCES-B. Data on deuterium atoms retained in liquid lithium indicate retention of all incident ions until full volumetric conversion to lithium deuteride. The PISCES-B results also show a material loss mechanism that lowers the maximum operating temperature compared to that for the liquid surface equilibrium vapor pressure. This may restrict the lithium temperature in reactors. (author)

  2. Three-dimensional MHD [magnetohydrodynamic] flows in rectangular ducts of liquid-metal-cooled blankets

    International Nuclear Information System (INIS)

    Hua, T.Q.; Walker, J.S.; Picologlou, B.F.; Reed, C.B.

    1988-07-01

    Magnetohydrodynamic flows of liquid metals in rectangular ducts with thin conducting walls in the presence of strong nonuniform transverse magnetic fields are examined. The interaction parameter and Hartmann number are assumed to be large, whereas the magnetic Reynolds number is assumed to be small. Under these assumptions, viscous and inertial effects are confined in very thin boundary layers adjacent to the walls. A significant fraction of the fluid flow is concentrated in the boundary layers adjacent to the side walls which are parallel to the magnetic field. This paper describes the analysis and numerical methods for obtaining 3-D solutions for flow parameters outside these layers, without solving explicitly for the layers themselves. Numerical solutions are presented for cases which are relevant to the flows of liquid metals in fusion reactor blankets. Experimental results obtained from the ALEX experiments at Argonne National Laboratory are used to validate the numerical code. In general, the agreement is excellent. 5 refs., 14 figs

  3. Results of neutron irradiation of liquid lithium saturated with deuterium

    International Nuclear Information System (INIS)

    Tazhibayeva, Irina; Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Skakov, Mazhyn; Zaurbekova, Zhanna; Lyublinski, Igor; Vertkov, Alexey; Mazzitelli, Giuseppe

    2017-01-01

    Highlights: • The results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1M research reactor are described. • At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10 9 exp(−144/RT). • The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. - Abstract: This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10 −13 cm −2 s −1 . The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10 9 exp(-144/RT). The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

  4. The liquid lithium limiter control system on FTU

    Energy Technology Data Exchange (ETDEWEB)

    Bertocchi, A. [EURATOM-ENEA Association, Frascati Research Center, Via E. Fermi 45, 00044 Frascati (Rome) (Italy)], E-mail: bertocchi@frascati.enea.it; Di Donna, M [Department of Informatics, Systems and Productions, University of Rome Tor Vergata, Rome (Italy); Panella, M; Vitale, V [EURATOM-ENEA Association, Frascati Research Center, Via E. Fermi 45, 00044 Frascati (Rome) (Italy)

    2007-10-15

    In the second half of 2005, a liquid lithium limiter (LLL) with capillary porous system (CPS) configuration was installed to test on Tokamak FTU. The liquid lithium flows through capillaries from a reservoir to the side faced to the plasma to form a thin lithium film as wall coating. The system includes three stainless steel cases, which contain two thermocouples each one. A heating system brings the Li temperature about 200 deg. C to allow the liquid to flow. This temperature, monitored by thermocouples, needs to be controlled. To carry out this experimental procedure, some new features have been introduced in the existent control system based on Opto22{sup TM} modules and a CORBA/PHP/MySQL software architecture. The historical data storage to keep the lithium temperature evolution has been added. Two graphical tools - developed in MATLAB{sup TM} and Java environments, respectively, to monitor the lithium temperature coming from thermocouples - have been also implemented. The LLL control system allows to regulate the heater temperature in each unit to reach operational conditions, where the temperature adjustment can be performed either automatically through a specific control law or manually by the operator. During the plasma shot the system switches off the limiter power supply to prevent instruments damage. Moreover, in the same experimental context, a first approach to automatically obtain executable code - starting from control laws designed by Simulink{sup TM} tool - has been realized.

  5. The liquid lithium limiter control system on FTU

    International Nuclear Information System (INIS)

    Bertocchi, A.; Di Donna, M.; Panella, M.; Vitale, V.

    2007-01-01

    In the second half of 2005, a liquid lithium limiter (LLL) with capillary porous system (CPS) configuration was installed to test on Tokamak FTU. The liquid lithium flows through capillaries from a reservoir to the side faced to the plasma to form a thin lithium film as wall coating. The system includes three stainless steel cases, which contain two thermocouples each one. A heating system brings the Li temperature about 200 deg. C to allow the liquid to flow. This temperature, monitored by thermocouples, needs to be controlled. To carry out this experimental procedure, some new features have been introduced in the existent control system based on Opto22 TM modules and a CORBA/PHP/MySQL software architecture. The historical data storage to keep the lithium temperature evolution has been added. Two graphical tools - developed in MATLAB TM and Java environments, respectively, to monitor the lithium temperature coming from thermocouples - have been also implemented. The LLL control system allows to regulate the heater temperature in each unit to reach operational conditions, where the temperature adjustment can be performed either automatically through a specific control law or manually by the operator. During the plasma shot the system switches off the limiter power supply to prevent instruments damage. Moreover, in the same experimental context, a first approach to automatically obtain executable code - starting from control laws designed by Simulink TM tool - has been realized

  6. Experiments for liquid metal embrittlement of fusion reactor materials by liquid lithium

    International Nuclear Information System (INIS)

    Grundmann, M.; Borgstedt, H.U.

    1984-10-01

    The liquid metal embrittlement behaviour of two martensitic-ferritic steels [X22CrMoV121 (Nr. 1.4923) and X18CrMoVNb 121 (Nr. 1,4914)] and one austenite chromium-nickel-steel X5CrNi189 (Nr. 1.4301) was investigated. Tensile tests in liquid lithium at 200 and 250 0 C with two different strain rates on precorroded samples (1000 h at 550 0 C in lithium) were carried out. Reference values were gained from tensile tests in air (RT, 250 0 C). It is concluded that there is sufficient compatibility of the austenitic steel with liquid lithium. The use of the ferritic-martensitic steels in liquid lithium on the other hand, especially at temperatures of about 550 0 C, seems to be problematic. The experimental results led to a better understanding of LME, applying the theory of this material failure. (orig./IHOE) [de

  7. 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

  8. 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

  9. The Liquid Lithium Limiter control system on FTU

    International Nuclear Information System (INIS)

    Bertocchi, A.; Panella, M.; Vitale, V.; Sinibaldi, S.

    2006-01-01

    In the second half of 2005, a liquid lithium limiter (LLL) with capillary porous system configuration was installed for testing on the FTU tokamak. The liquid lithium flows through capillaries from a reservoir to the side facing the plasma to form a thin liquid lithium film. The system is composed of three stainless steel sections, which contain two thermocouples each. A heating system brings the Li temperature to about 200 o C allowing the liquid to flow. This temperature, monitored by thermocouples, needs to be controlled. [M. Apicella, G. Mazzitelli et al., First experiment with Lithium Limiter on FTU, 17 o International Conference on Plasma Surface Interaction in Controlled Fusion Devices, 22 - 26 May 2006, Hefei Anhui, China]. To carry out this experimental procedure, some new features have been introduced in the existent control system based on Opto22 TM modules and a CORBA/PHP/MySQL software architecture [A. Bertocchi, S. Podda, V. Vitale, Fusion Eng. Des. 74 (2005) 787-791]. The historical data storage to keep the lithium temperature evolution has been added. Two graphical tools - developed in MATLab and Java environments respectively to monitor the lithium temperature coming from thermocouples - have been also implemented. The control system allows regulating the heater temperature in each section of the LLL to reach operational conditions, where the temperature adjustment can be performed either automatically through a specific control law or manually by the operator. During plasma operations the system switches off the limiter power supply to prevent instruments damage. Moreover, in the same experimental context, a first approach to automatically obtain executable code - starting from control laws designed by Simulink TM tool - has been realized. (author)

  10. The Liquid Lithium Limiter control system on FTU

    Energy Technology Data Exchange (ETDEWEB)

    Bertocchi, A; Panella, M; Vitale, V [Associazione EURATOM- ENEA sulla Fusione, Via Enrico Fermi 45, I-00044 Frascati (RM) (Italy); Sinibaldi, S [Rome University ' ' Tor Vergata ' ' , Informatics, Systems and Production Dept., Via del Politecnico 1, 00133 Rome (Italy)

    2006-07-01

    In the second half of 2005, a liquid lithium limiter (LLL) with capillary porous system configuration was installed for testing on the FTU tokamak. The liquid lithium flows through capillaries from a reservoir to the side facing the plasma to form a thin liquid lithium film. The system is composed of three stainless steel sections, which contain two thermocouples each. A heating system brings the Li temperature to about 200 {sup o}C allowing the liquid to flow. This temperature, monitored by thermocouples, needs to be controlled. [M. Apicella, G. Mazzitelli et al., First experiment with Lithium Limiter on FTU, 17{sup o} International Conference on Plasma Surface Interaction in Controlled Fusion Devices, 22 - 26 May 2006, Hefei Anhui, China]. To carry out this experimental procedure, some new features have been introduced in the existent control system based on Opto22{sup TM} modules and a CORBA/PHP/MySQL software architecture [A. Bertocchi, S. Podda, V. Vitale, Fusion Eng. Des. 74 (2005) 787-791]. The historical data storage to keep the lithium temperature evolution has been added. Two graphical tools - developed in MATLab and Java environments respectively to monitor the lithium temperature coming from thermocouples - have been also implemented. The control system allows regulating the heater temperature in each section of the LLL to reach operational conditions, where the temperature adjustment can be performed either automatically through a specific control law or manually by the operator. During plasma operations the system switches off the limiter power supply to prevent instruments damage. Moreover, in the same experimental context, a first approach to automatically obtain executable code - starting from control laws designed by Simulink{sup TM} tool - has been realized. (author)

  11. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    Veleckis, E.; Yonco, R.M.; Maroni, V.A.

    1980-01-01

    The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li 3 N, Li 2 O, and Li 2 C 2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

  12. The solubility of carbon in low-nitrogen liquid lithium

    International Nuclear Information System (INIS)

    Yonco, R.M.; Homa, M.I.

    1986-01-01

    The solubility of carbon in liquid lithium containing 0 C and compared with the solubility in lithium containing proportional 2600 wppm nitrogen in that same temperature range. A direct sampling method was employed in which filtered samples of the saturated solution were taken at randomly selected temperatures. The entire sample was analyzed for carbon by the acetylene evolution method. The analytical method was examined critically and it was found that (1) all of the carbon in solution, including carbon introduced as lithium cyanamide is detected and (2) ethylene and ethane must also be measured and included with the acetylene to get complete recovery of the carbon content of the sample. The solubility of carbon in low-nitrogen lithium can be expressed by the equations ln S=6.731-8617T -1 and log Ssup(*)=7.459-3740T -1 , where S is the mole percent Li 2 C 2 and Ssup(*) is in weight parts per million carbon. The presence of proportional 2600 wppm nitrogen does not affect the solubility of carbon in lithium at temperatures above proportional 350 0 C, but at lower temperatures it increased the solubility by as much as an order of magnitude compared to the solubility in low-nitrogen lithium. (orig.)

  13. 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

  14. Hydrogen extraction from liquid lithium-lead alloy by gas-liquid contact method

    International Nuclear Information System (INIS)

    Xie Bo; Weng Kuiping; Hou Jianping; Yang Guangling; Zeng Jun

    2013-01-01

    Hydrogen extraction experiment from liquid lithium-lead alloy by gas-liquid contact method has been carried out in own liquid lithium-lead bubbler (LLLB). Experimental results show that, He is more suitable than Ar as carrier gas in the filler tower. The higher temperature the tower is, the greater hydrogen content the tower exports. Influence of carrier gas flow rate on the hydrogen content in the export is jagged, no obvious rule. Although the difference between experimental results and literature data, but it is feasible that hydrogen isotopes extraction experiment from liquid lithium-lead by gas-liquid contact method, and the higher extraction efficiency increases with the growth of the residence time of the alloy in tower. (authors)

  15. RELAP/SCDAPSIM/MOD4.0 modification for transient accident scenario of Test Blanket Modules in ITER involving helium flows into heavy liquid metal

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J.; Pérez, M.; Mas de les Valls, E.; Batet, L.; Sandeep, T.; Chaudhari, V.; Reventós, F.

    2015-07-01

    The Institute for Plasma Research (IPR), India, is currently involved in the design and development of its Test Blanket Module (TBM) for testing in ITER (International Thermo nuclear Experimental Reactor). The Indian TBM concept is a Lead-Lithium cooled Ceramic Breeder (LLCB), which utilizes lead-lithium eutectic alloy (LLE) as tritium breeder, neutron multiplier and coolant. The first wall facing the plasma is cooled by helium gas. In preparation of the regulatory safety files of ITER-TBM, a number of off-normal event sequences have been postulated. Thermal hydraulic safety analyses of the TBM system will be carried out with the system code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems during normal and accidental conditions. In order to analyze some of the postulated off-normal events, there is the need to simulate the mixing of Helium and Lead-Lithium fluids. The Technical University of Catalonia is cooperating with IPR to implement the necessary changes in the code to allow for the mixing of helium and liquid metal. In the present study, the RELAP/SCDAPSIM/MOD4 two-phase flow 6-equations structure has been modified to allow for the mixture of LLE in the liquid phase with dry Helium in the gas phase. Practically obtaining a two-fluid 6-equation model where each fluid is simulated with a set of energy, mass and momentum balance equations. A preliminary flow regime map for LLE and helium flow has been developed on the basis of numerical simulations with the OpenFOAM CFD toolkit. The new code modifications have been verified for vertical and horizontal configurations. (Author)

  16. Control of beryllium-7 in liquid lithium

    International Nuclear Information System (INIS)

    Anantatmula, R.P.; Brehm, W.F.; Baldwin, D.L.; Bevan, J.L.

    1978-12-01

    Radiation fields created by the production of 7 Be in lithium of the Fusion Materials Irradiation Test (FMIT) Facility can be sufficiently high to prevent contact maintenance of system components. Preliminary experiments have shown that 7 Be will adhere strongly to the FMIT piping and components and a good control method for 7 Be must be developed. The initial experiments have been conducted in static stainless steel capsules and a Modified Thermal Convection Loop (MTCL). The average lithium film thickness on stainless steel was found to be 11 μm in the temperature range 495 0 to 571 0 K from the capsule experiments. The diffusion coefficient for 7 Be in stainless steel at 543 0 K was calculated to be 5.31 x 10 -15 cm 2 /sec. The cold leg of the MTCL picked up much of the 7 Be activity released into the loop. The diffusion trap, located in the cold leg of the MTCL, was ineffective in removing 7 Be from lithium, at the very slow flow rates ( -4 m 3 /s) used in the MTCL. Pure iron has been shown to be superior to coblat and nickel as a getter material for 7 Be

  17. Deuterium trapping in liquid lithium irradiated by deuterium plasma

    International Nuclear Information System (INIS)

    Pisarev, A.; Moshkunov, K.; Vizgalov, I.; Gasparyan, Yu.

    2013-01-01

    Liquid lithium was irradiated by deuterium plasma to a low fluence of 10 22 –10 23 D/m 2 , cooled down to room temperature, and then slowly heated. The temperature and release rate were measured during heating. Two plateaus on the temperature–time dependence were observed at 180 °C and 660 °C. The first one corresponds to melting of Li and the second one – either to melting or to decomposition of solid LiD. Features of deuterium release in TDS were interpreted in terms of decomposition of lithium deuterides formed during plasma irradiation

  18. Electrical detection of liquid lithium leaks from pipe joints

    Energy Technology Data Exchange (ETDEWEB)

    Schwartz, J. A., E-mail: jschwart@pppl.gov; Jaworski, M. A.; Mehl, J.; Kaita, R.; Mozulay, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States)

    2014-11-15

    A test stand for flowing liquid lithium is under construction at Princeton Plasma Physics Laboratory. As liquid lithium reacts with atmospheric gases and water, an electrical interlock system for detecting leaks and safely shutting down the apparatus has been constructed. A defense in depth strategy is taken to minimize the risk and impact of potential leaks. Each demountable joint is diagnosed with a cylindrical copper shell electrically isolated from the loop. By monitoring the electrical resistance between the pipe and the copper shell, a leak of (conductive) liquid lithium can be detected. Any resistance of less than 2 kΩ trips a relay, shutting off power to the heaters and pump. The system has been successfully tested with liquid gallium as a surrogate liquid metal. The circuit features an extensible number of channels to allow for future expansion of the loop. To ease diagnosis of faults, the status of each channel is shown with an analog front panel LED, and monitored and logged digitally by LabVIEW.

  19. Study of MHD problems in liquid metal blankets of fusion reactors

    International Nuclear Information System (INIS)

    Michael, I.

    1984-12-01

    This study describes in a concise form the state of knowledge regarding MHD problems to be expected in case of use of liquid metal in the blankets of fusion reactors with magnetic confinement. MHD pressure losses and MHD friction coefficients in the straight channel, in bent sections and in case of variation of the channel cross section play a major role because the high MHD flow resistances call for high pumping powers. Influencing the velocity profile transverse to the main flow direction of the liquid metal by application of an external, strong magnetic field bears consequences on the release and transport of corrosion products in the liquid metal circuit and on the heat transfer. Possibilities of reducing the MHD effects are discussed. However, it becomes obvious that an account of the lack of experimental results there are still major gaps in the knowledge of MHD effects occurring in strong magnetic fields. These gaps can be greatly reduced by implementation of an experimental program as proposed in this report. (orig.) [de

  20. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    Science.gov (United States)

    Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

    2009-11-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  1. Homogeneous lithium electrodeposition with pyrrolidinium-based ionic liquid electrolytes.

    Science.gov (United States)

    Grande, Lorenzo; von Zamory, Jan; Koch, Stephan L; Kalhoff, Julian; Paillard, Elie; Passerini, Stefano

    2015-03-18

    In this study, we report on the electroplating and stripping of lithium in two ionic liquid (IL) based electrolytes, namely N-butyl-N-methylpyrrolidinium bis(fluorosulfonyl) imide (Pyr14FSI) and N-butyl-N-methylpyrrolidinium bis(trifluoromethanesulfonyl)imide (Pyr14TFSI), and mixtures thereof, both on nickel and lithium electrodes. An improved method to evaluate the Li cycling efficiency confirmed that homogeneous electroplating (and stripping) of Li is possible with TFSI-based ILs. Moreover, the presence of native surface features on lithium, directly observable via scanning electron microscope imaging, was used to demonstrate the enhanced electrolyte interphase (SEI)-forming ability, that is, fast cathodic reactivity of this class of electrolytes and the suppressed dendrite growth. Finally, the induced inhomogeneous deposition enabled us to witness the SEI cracking and revealed previously unreported bundled Li fibers below the pre-existing SEI and nonrod-shaped protuberances resulting from Li extrusion.

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

    International Nuclear Information System (INIS)

    Malang, S.; Reimann, J.; Sebening, H.; Barleon, L.; Bogusch, E.; Bojarsky, E.; Borgstedt, H.U.; Buehler, L.; Casal, V.; Deckers, H.; Feuerstein, H.; Fischer, U.; Frees, G.; Graebner, H.; John, H.; Jordan, T.; Kramer, W.; Krieg, R.; Lenhart, L.; Malang, S.; Meyder, R.; Norajitra, P.; Reimann, J.; Schwenk-Ferrero, A.; Schnauder, H.; Stieglitz, R.; Oschinski, J.; Wiegner, E.

    1991-12-01

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

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

    International Nuclear Information System (INIS)

    John, H.; Malang, S.; Sebening, H.

    1991-12-01

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

  4. Results of neutron irradiation of liquid lithium saturated with deuterium

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibayeva, Irina, E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Skakov, Mazhyn; Zaurbekova, Zhanna [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Lyublinski, Igor [JSC «Red Star», Moscow (Russian Federation); NRNU «MEPhI», Moscow (Russian Federation); Vertkov, Alexey [JSC «Red Star», Moscow (Russian Federation); Mazzitelli, Giuseppe [ENEA, RC Frascati, Frascati (Italy)

    2017-04-15

    Highlights: • The results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1M research reactor are described. • At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10{sup 9} exp(−144/RT). • The T{sub 2} molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. - Abstract: This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10{sup −13} cm{sup −2} s{sup −1}. The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10{sup 9} exp(-144/RT). The T{sub 2} molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

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

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.; Jankhah, M.H.

    1979-01-01

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

  6. On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it; Arena, P.; Bongiovì, G.; Chiovaro, P.; Forte, R.; Garitta, S.

    2016-11-01

    Highlights: • The geometric optimization of the DEMO WCLL blanket module first wall has been performed, maximizing the heat flux it may safely undergo. • Attention has been focused on the FW flat concept endowed with square cooling channels. • A theoretical-computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Four optimized FW configurations have been found to safely withstand a heat flux up to 2 MW/m{sup 2} fulfilling all the rules prescribed by safety codes. - Abstract: Within the framework of EUROfusion R&D activities a research campaign has been carried out at the University of Palermo in order to investigate the thermo-mechanical performances of the DEMO water-cooled lithium lead (WCLL) breeding blanket first wall (FW). The research campaign has been mainly focused on the optimization of the FW geometric configuration in order to maximize the heat flux it may safely withstand fulfilling all the thermal, hydraulic and mechanical requirements foreseen by safety codes. Attention has been focused on the FW flat concept endowed with square cooling channels and the potential influence of its four main geometrical parameters on its thermo-mechanical performances has been assessed performing a parametric analysis by means of a qualified commercial finite element method code. A set of 5929 different FW geometric configurations has been considered and the thermal performances of each one of them have been numerically assessed in case it undergoes 26 different values of heat flux on its plasma-facing surface. The resulting 154154 thermal analyses have allowed to select those cases fulfilling the adopted thermal-hydraulic requirements, whose thermo-mechanical performances have been numerically assessed under both normal operation and over-pressurization steady state loading scenarios to check whether they met the mechanical requirements prescribed by the pertaining SDC-IC safety rules. Four

  7. On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

    International Nuclear Information System (INIS)

    Di Maio, P.A.; Arena, P.; Bongiovì, G.; Chiovaro, P.; Forte, R.; Garitta, S.

    2016-01-01

    Highlights: • The geometric optimization of the DEMO WCLL blanket module first wall has been performed, maximizing the heat flux it may safely undergo. • Attention has been focused on the FW flat concept endowed with square cooling channels. • A theoretical-computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Four optimized FW configurations have been found to safely withstand a heat flux up to 2 MW/m"2 fulfilling all the rules prescribed by safety codes. - Abstract: Within the framework of EUROfusion R&D activities a research campaign has been carried out at the University of Palermo in order to investigate the thermo-mechanical performances of the DEMO water-cooled lithium lead (WCLL) breeding blanket first wall (FW). The research campaign has been mainly focused on the optimization of the FW geometric configuration in order to maximize the heat flux it may safely withstand fulfilling all the thermal, hydraulic and mechanical requirements foreseen by safety codes. Attention has been focused on the FW flat concept endowed with square cooling channels and the potential influence of its four main geometrical parameters on its thermo-mechanical performances has been assessed performing a parametric analysis by means of a qualified commercial finite element method code. A set of 5929 different FW geometric configurations has been considered and the thermal performances of each one of them have been numerically assessed in case it undergoes 26 different values of heat flux on its plasma-facing surface. The resulting 154154 thermal analyses have allowed to select those cases fulfilling the adopted thermal-hydraulic requirements, whose thermo-mechanical performances have been numerically assessed under both normal operation and over-pressurization steady state loading scenarios to check whether they met the mechanical requirements prescribed by the pertaining SDC-IC safety rules. Four

  8. Effect of Energetic Plasma Flux on Flowing Liquid Lithium Surfaces

    Science.gov (United States)

    Kalathiparambil, Kishor; Jung, Soonwook; Christenson, Michael; Fiflis, Peter; Xu, Wenyu; Szott, Mathew; Ruzic, David

    2014-10-01

    An operational liquid lithium system with steady state flow driven by thermo-electric magneto-hydrodynamic force and capable of constantly refreshing the plasma exposed surface have been demonstrated at U of I. To evaluate the system performance in reactor relevant conditions, specifically to understand the effect of disruptive plasma events on the performance of the liquid metal PFCs, the setup was integrated to a pulsed plasma generator. A coaxial plasma generator drives the plasma towards a theta pinch which preferentially heats the ions, simulating ELM like flux, and the plasma is further guided towards the target chamber which houses the flowing lithium system. The effect of the incident flux is examined using diagnostic tools including triple Langmuir probe, calorimeter, rogowski coils, Ion energy analyzers, and fast frame spectral image acquisition with specific optical filters. The plasma have been well characterized and a density of ~1021 m-3, with electron temperature ~10 - 20 eV is measured, and final plasma velocities of 34 - 74 kms-1 have been observed. Calorimetric measurements using planar molybdenum targets indicate a maximum plasma energy (with 6 kV plasma gun and 20 kV theta pinch) of 0.08 MJm-2 with plasma divergence effects resulting in marginal reduction of 40 +/- 23 J in plasma energy. Further results from the other diagnostic tools, using the flowing lithium targets and the planar targets coated with lithium will be presented. DOE DE-SC0008587.

  9. Spectroscopic measurements of lithium influx from an actively water-cooled liquid lithium limiter on FTU

    Energy Technology Data Exchange (ETDEWEB)

    Apruzzese, G.M., E-mail: gerarda.apruzzese@enea.it; Apicella, M.L.; Maddaluno, G.; Mazzitelli, G.; Viola, B.

    2017-04-15

    Since 2006, experiments using a liquid lithium limiter (LLL) were successfully performed on FTU, pointing out the problem of the quantity of lithium in the plasma, especially in conditions of strong evaporation due to the high temperature of limiter surface. In order to avoid the strong evaporation it is necessary to control the temperature by removing the heat from the limiter during the plasma exposure. To explore this issue a new actively cooled lithium limiter (CLL) has been installed and tested in FTU. Suitable monitors to detect the presence of lithium in the plasma are the spectroscopic diagnostics in the visible range that permit to measure the flux of lithium, coming from the limiter surface, through the brightness of the LiI spectral lines. For this aim an Optical Multichannel Analyser (OMA) spectrometer and a single wavelength impurities monitor have been used. The analysis of the Li influx signals has permitted to monitor the effects of interaction between the plasma and the limiter connected to the thermal load. Particular attention has been paid on the possible occurrence of sudden rise of the signals, which is an index of a strong interaction that could lead to a disruption. On the other hand, the appearance of significant signals gives useful indication if the interaction with the plasma has taken place.

  10. Analysis of the thermo-mechanical behaviour of the DEMO Water-Cooled Lithium Lead breeding blanket module under normal operation steady state conditions

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A.; Arena, P. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Aubert, J. [CEA Saclay, DEN/DANS/DM2S/SEMT, 91191 Gif sur Yvette Cedex (France); Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, R. [ENEA – C.R. Brasimone, 40032 Camugnano (Italy); Li Puma, A. [CEA Saclay, DEN/DANS/DM2S/SEMT, 91191 Gif sur Yvette Cedex (France); Tincani, A. [ENEA – C.R. Brasimone, 40032 Camugnano (Italy)

    2015-10-15

    Highlights: • A DEMO WCLL blanket module thermo-mechanical behaviour has been investigated. • Two models of the WCLL blanket module have been set-up adopting a code based on FEM. • The water flow domain in the module has been considered. • A set of uncoupled steady state thermo-mechanical analyses has been carried out. • Critical temperature is not overcome. Safety verifications are generally satisfied. - Abstract: Within the framework of DEMO R&D activities, a research cooperation has been launched between ENEA, the University of Palermo and CEA to investigate the thermo-mechanical behaviour of the outboard equatorial module of the DEMO1 Water-Cooled Lithium Lead (WCLL) blanket under normal operation steady state scenario. The research campaign has been carried out following a theoretical–computational approach based on the Finite Element Method (FEM) and adopting a qualified commercial FEM code. In particular, two different 3D FEM models (Model 1 and Model 2), reproducing respectively the central and the lateral poloidal–radial slices of the WCLL blanket module, have been set up. A particular attention has been paid to the modelling of water flow domain, within both the segment box channels and the breeder zone tubes, to simulate realistically the coolant-box thermal coupling. Results obtained are herewith reported and critically discussed.

  11. Liquid blanket MHD effects experimental results from LMEL facility at SWIP

    International Nuclear Information System (INIS)

    Xu Zengyu; Pan Chuanjie; Liu Yong; Pan Chuanhong; Reed, C.B.

    2007-01-01

    The self-cooled /helium-cooled liquid metal blanket concept is an attractive ITER and DEMO blanket candidate as it has low operating pressure, simplicity, and a convenient tritium breeding cycle. But MHD pressure drop remains a key issue, especially in ducts with flow channel inserts (FCI), where the reduction in MHD pressure drop is difficult to predict with existing tools, and there are no available experimental data to check current predictions. To understand well various kinds of MHD effects, it is important for us to analyze and understand FCI effects. In this paper, we present measurements of the MHD effects due to off normal power shutdown, two-dimensional effects due to channel velocity profiles, three-dimensional effects caused by manifolds, and surface/bulk instability effects as a result of insulator coating imperfections. These results were collected from the Liquid Metal Experimental Loop (LMEL) facility at Southwestern Institute of Physics, China and in collaboration with Argonne National Laboratory, US under an umbrella of the People's Republic of China/United States program of cooperation in magnetic fusion. Some results were observed for the first time, such as two dimensional effects and instabilities due to insulator coating imperfections. The experiments were conducted under the following conditions: a uniform magnetic field volume of 80 x 170 x 740 mm and a maximum value of magnetic field, B 0 , of 2 Tesla. The mean flow velocity v 0 was measured with an electromagnetic (EM) flow meter (error of 1.2%); a Liquid-metal Electro-magnetic Velocity Instrument (LEVI) was provided by Argonne National Laboratory. The flow was driven by two Electro-magnetic (EM) pumps (6.5+11.6 m3/h); the operating temperature was 85 centigrade degree due to self-heating by the EM pump and friction of the fluid against the loop piping. Experimental parameters were: Hartmann number, M, up to 3500, velocity v 0 up to 1.2 m/s under magnetic field, and B 0 =1.95 Tesla

  12. Supercritical CO2 Brayton power cycles for DEMO (demonstration power plant) fusion reactor based on dual coolant lithium lead blanket

    International Nuclear Information System (INIS)

    Linares, José Ignacio; Cantizano, Alexis; Moratilla, Beatriz Yolanda; Martín-Palacios, Víctor; Batet, Lluis

    2016-01-01

    This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles as alternative energy conversion systems for a future fusion reactor based on a DCLL (dual coolant lithium-lead) blanket, as prescribed by EUROfusion. The main issue dealt is the optimization of the integration of the different thermal sources with the power cycle in order to achieve the highest electricity production. The analysis includes the assessment of the pumping consumption in the heating and cooling loops, taking into account additional considerations as control issues and integration of thermal energy storage systems. An exergy analysis has been performed in order to understand the behavior of each layout. Up to ten scenarios have been analyzed assessing different locations for thermal sources heat exchangers. Neglecting the worst four scenarios, it is observed less than 2% of variation among the other six ones. One of the best six scenarios clearly stands out over the others due to the location of the thermal sources in a unique island, being this scenario compatible with the control criteria. In this proposal 34.6% of electric efficiency (before the self-consumptions of the reactor but including pumping consumptions and generator efficiency) is achieved. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of DCLL fusion reactor. • Integration of different available thermal sources has been analyzed considering ten scenarios. • Neglecting the four worst scenarios the electricity production varies less than 2%. • Control and energy storage integration issues have been considered in the analysis. • Discarding the vacuum vessel and joining the other sources in an island is proposed.

  13. Tritium breeding measurements in a lithium blanket module with Pb/Be multipliers at the LOTUS facility

    International Nuclear Information System (INIS)

    Azam, S.; Kumar, A.

    1987-01-01

    The lithium blanket module (LBM) was lent for a fixed duration in 1985 to Ecole Polytechnique Federale de Lausanne under an agreement with the Electric Power Research Institute and Princeton Plasma Physics Laboratory. The first tritium breeding measurements in the central rod of the LBM and their analysis have been reported previously. Some time ago, we carried out additional experiments wherein the Li 2 O sample disk, each having a theoretical density of ∼85% and dimensions of 17.8-mm diam x 0.9-mm thickness, were placed in four removable rods. In addition to the central rod, the other rods were at ∼6-, 18-, and 39-cm radial distances from the axis of the central one. The sample disks wee kept at every 3 cm inside each of these rods up to a length of 30 cm in the Li 2 O part of the LBM. The choice of the off-axis rods resulted from our interest in investigating the effect of room return on tritium breeding in the LBM. We chose two of the leading neutron multipliers: (a) a 5-cm-thick (∼100- x 110-cm) lead slab and (b) a 6-cm-thick (∼66- x 66-cm) beryllium slab. The experimental assembly, consisting of the multiplier followed by the LBM, was kept at 10 cm from the generator. A packet of three foils, zirconium, indium, and aluminum, was placed at the center of the flat face of the generator to monitor the source intensity during the 10-h operation for the experiments with each multiplier. The source intensity is deduced to be ∼1.9 x 10 12 n/s for both the experiments. 5 refs., 3 figs

  14. Investigation of wetting property between liquid lead lithium alloy and several structural materials for Chinese DEMO reactor

    Science.gov (United States)

    Lu, Wei; Wang, Weihua; Jiang, Haiyan; Zuo, Guizhong; Pan, Baoguo; Xu, Wei; Chu, Delin; Hu, Jiansheng; Qi, Junli

    2017-10-01

    The dual-cooled lead lithium (PbLi) blanket is considered as one of the main options for the Chinese demonstration reactor (DEMO). Liquid PbLi alloy is used as the breeder material and coolant. Reduced activation ferritic/martensitic (RAFM) steel, stainless steel and the silicon carbide ceramic matrix composite (SiCf) are selected as the substrate materials for different use. To investigate the wetting property and inter-facial interactions of PbLi/RAFM steel, PbLi/SS316L, PbLi/SiC and PbLi/SiCf couples, in this paper, the special vacuum experimental device is built, and the 'dispensed droplet' modification for the classic sessile droplet technique is made. Contact angles are measured between the liquid PbLi and the various candidate materials at blanket working temperature from 260 to 480 °C. X-ray photoelectron spectroscopy (XPS) is used to characterize the surface components of PbLi droplets and substrate materials, in order to study the element trans-port and corrosion mechanism. Results show that SiC composite (SiCf) and SiC ceramic show poor wetting properties with the liquid PbLi alloy. Surface roughness and testing temperature only provide tiny improvements on the wetting property below 480 °C. RAFM steel performs better wetting properties and corrosion residence when contacted with molten PbLi, while SS316L shows low corrosion residence above 420 °C for the decomposition of protective surface film mainly consisted of chromic sesquioxide. The results could provide meaningful compatibility database of liquid PbLi alloy and valuable reference in engineering design of candidate structural and functional materials for future fusion blanket.

  15. Assessment of alkali metal coolants for the ITER blanket

    International Nuclear Information System (INIS)

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

    1994-01-01

    The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

  16. Solutions of group IV elements in liquid lithium

    International Nuclear Information System (INIS)

    Dadd, A.T.; Hubberstey, P.; Roberts, P.G.

    1982-01-01

    The solubilities of tin (0.00 = 22 Sn 5 . A simple thermochemical cycle is used to demonstrate that, whereas carbon dissolves endothermically in both liquid lithium and liquid sodium, the heavier Group IV elements dissolve exothermically. A similar cycle is used to derive solvation enthalpies (for the neutral gaseous species) for all Group IV elements in the two solvents. The trend in solvation enthalpy: C > Si > Ge > Sn > Pb is indicative of a diminishing affinity of solvent for solute and is attributed to the increasing metallic character of the solute as the Group is descended. (author)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

    The blanket concept for the tandem mirror reactor described in this paper was developed to produce the medium temperature heat (approx. 850 to 950 K) for the General Atomic sulfur-iodine thermochemical process for producing hydrogen. This medium temperature heat from the blanket constitutes about 81% of the total power output of the fusion reactor

  18. Special topics reports for the reference tandem mirror fusion breeder: liquid metal MHD pressure drop effects in the packed bed blanket. Vol. 1

    International Nuclear Information System (INIS)

    McCarville, T.J.; Berwald, D.H.; Wong, C.P.C.

    1984-09-01

    Magnetohydrodynamic (MHD) effects which result from the use of liquid metal coolants in magnetic fusion reactors include the modification of flow profiles (including the suppression of turbulence) and increases in the primary loop pressure drop and the hydrostatic pressure at the first wall of the blanket. In the reference fission-suppressed tandem mirror fusion breeder design concept, flow profile modification is a relatively minor concern, but the MHD pressure drop in flowing the liquid lithium coolant through an annular packed bed of beryllium/thorium pebbles is directly related to the required first wall structure thickness. As such, it is a major concern which directly impacts fissile breeding efficiency. Consequently, an improved model for the packed bed pressure drop has been developed. By considering spacial averages of electric fields, currents, and fluid flow velocities the general equations have been reduced to simple expressions for the pressure drop. The averaging approach results in expressions for the pressure drop involving a constant which reflects details of the flow around the pebbles. Such details are difficult to assess analytically, and the constant may eventually have to be evaluated by experiment. However, an energy approach has been used in this study to bound the possible values of the constant, and thus the pressure drop. In anticipation that an experimental facility might be established to evaluate the undetermined constant as well as to address other uncertainties, a survey of existing facilities is presented

  19. A molecular dynamics study of lithium-containing aprotic heterocyclic ionic liquid electrolytes

    Science.gov (United States)

    Lourenço, Tuanan C.; Zhang, Yong; Costa, Luciano T.; Maginn, Edward J.

    2018-05-01

    Classical molecular dynamics simulations were performed on twelve different ionic liquids containing aprotic heterocyclic anions doped with Li+. These ionic liquids have been shown to be promising electrolytes for lithium ion batteries. Self-diffusivities, lithium transference numbers, densities, and free volumes were computed as a function of lithium concentration. The dynamics and free volume decreased with increasing lithium concentration, and the trends were rationalized by examining the changes to the liquid structure. Of those examined in the present work, it was found that (methyloxymethyl)triethylphosphonium triazolide ionic liquids have the overall best performance.

  20. Numerical Investigation of magnetohydrodynamic flow through Sudden expansion pipes in Liquid Metal Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Jingchao; He, Qingyun; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Ye, Minyou

    2016-11-01

    In fusion liquid metal blanket, sudden expansions and sudden contractions are very common geometries. Changing of the cross-section causes 3-D magnetohydrodynamic (MHD) effects, which will affect the flow pattern, current distribution and pressure drop. In this paper the numerical code based on OpenFOAM platform developed by University of Science and Technology of China was used to investigate and optimize the sudden expansion pipe. The code has been validated by the recommended benchmark cases including Shercliff, Hunt, ALEX experiments (rectangular duct and round pipe) and KIT experiment cases. The obtained numerical results agreed well with those of all the benchmark cases. Previous and valuable analytical and experimental works have been done by L. Buhler, et. el. Based on these works, in the present paper, further investigation of different expansion lengths between the upstream pipe and downstream pipe at high Hartmann number and Reynolds number were conducted. Besides, different expansion ratios with a specific expansion length were conducted. The numerical results showed that with the increasing of expansion length, the 3D MHD effects gradually weakened. Especially, the 3D pressure drop decreases with the increasing of expansion length. Whereas, the expansion ratio factor shows no obvious influences on the total MHD pressure drop but greatly influence the local pressure distribution. These numerical simulations can be used to evaluate the MHD flow inside the expansion and contraction pipes.

  1. Electromagnetic pumping of liquid lithium in inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Baker, R.S.; Blink, J.A.; Tessier, M.J.

    1983-01-01

    The basic operating principles and geometries of ten electromagnetic pumps are described. Two candidate pumps, the annular-linear-induction pump and the helical-rotor electromagnetic pump, are compared for possible use in a full-scale liquid-lithium inertial confinement fusion reactor. A parametric design study completed for the helical-rotor pump is shown to be valid when applied to an experimental sodium pump. Based upon the preliminary HYLIFE requirements for a lithium flow rate per pump of 8.08 m 3 /s at a head of 82.5 kPa, a complete set of 70 variables are specified for a helical-rotor pump with either a normally conducting or a superconducting winding. The two alternative designs are expected to perform with efficiencies of 50 and 60%, respectively

  2. A new facility for studying plasma interacting with flowing liquid lithium surface

    International Nuclear Information System (INIS)

    Cao, X.; Ou, W.; Tian, S.; Wang, C.; Zhu, Z.; Wang, J.; Gou, F.; Yang, D.; Chen, S.

    2014-01-01

    A new facility to study plasmas interacting with flowing liquid lithium surface was designed and is constructing in Sichuan University. The integrated setup includes the liquid lithium circulating part and linear high density plasma generator. The circulating part is consisted of main loop, on-line monitor system, lithium purification system and temperature programmed desorption system. In our group a linear high density plasma generator was built in 2012. Three coils were mounted along the vessel to produce an axial magnetic field inside. The magnetic field strength is up to 0.45 T and work continuously. Experiments on plasmas interacting with free flowing liquid lithium surface will be performed

  3. Advantages and Challenges of Radiative Liquid Lithium Divertor

    Science.gov (United States)

    Ono, Masayuki

    2017-10-01

    Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid Li divertor (RLLD) concept and its variant, the active liquid Li divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest Li-loop could provide a possible solution for the outstanding fusion reactor technology issues such as divertor heat flux mitigation and real time dust removal, while potentially improving the reactor plasma performance. Laboratory tests are also planned to investigate the Li-T recover efficiency and other relevant research topics of the RLLD. This work supported by DoE Contract No. DE-AC02-09CH11466.

  4. Response of NSTX liquid lithium divertor to high heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Abrams, T., E-mail: tabrams@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaworski, M.A. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Kallman, J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Foley, E.L. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kugel, H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Levinton, F. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2013-07-15

    Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ∼1.5 MW/m{sup 2} for 1–3 s. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the “bare” sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface.

  5. 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.)

  6. 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

  7. Magnetohydrodynamic pressure drop and flow balancing of liquid metal flow in a prototypic fusion blanket manifold

    Science.gov (United States)

    Rhodes, Tyler J.; Smolentsev, Sergey; Abdou, Mohamed

    2018-05-01

    Understanding magnetohydrodynamic (MHD) phenomena associated with the flow of electrically conducting fluids in complex geometry ducts subject to a strong magnetic field is required to effectively design liquid metal (LM) blankets for fusion reactors. Particularly, accurately predicting the 3D MHD pressure drop and flow distribution is important. To investigate these topics, we simulate a LM MHD flow through an electrically non-conducting prototypic manifold for a wide range of flow and geometry parameters using a 3D MHD solver, HyPerComp incompressible MHD solver for arbitrary geometry. The reference manifold geometry consists of a rectangular feeding duct which suddenly expands such that the duct thickness in the magnetic field direction abruptly increases by a factor rexp. Downstream of the sudden expansion, the LM is distributed into several parallel channels. As a first step in qualifying the flow, a magnitude of the curl of the induced Lorentz force was used to distinguish between inviscid, irrotational core flows and boundary and internal shear layers where inertia and/or viscous forces are important. Scaling laws have been obtained which characterize the 3D MHD pressure drop and flow balancing as a function of the flow parameters and the manifold geometry. Associated Hartmann and Reynolds numbers in the computations were ˜103 and ˜101-103, respectively, while rexp was varied from 4 to 12. An accurate model for the pressure drop was developed for the first time for inertial-electromagnetic and viscous-electromagnetic regimes based on 96 computed cases. Analysis shows that flow balance can be improved by lengthening the distance between the manifold inlet and the entrances of the parallel channels by utilizing the effect of flow transitioning to a quasi-two-dimensional state in the expansion region of the manifold.

  8. Conductivity of liquid lithium electrolytes with dispersed mesoporous silica particles

    International Nuclear Information System (INIS)

    Sann, K.; Roggenbuck, J.; Krawczyk, N.; Buschmann, H.; Luerßen, B.; Fröba, M.; Janek, J.

    2012-01-01

    Highlights: ► The conductivity of disperse lithium electrolytes with mesoporous fillers is studied. ► In contrast to other investigations in literature, no conductivity enhancement could be observed for standard battery electrolytes and typical mesoporous fillers in various combinations. ► Disperse electrolytes can become relevant in terms of battery safety. ► Dispersions of silicas and electrolyte with LiPF 6 as conducting salt are not stable, although the silicas were dried prior to preparation and the electrolyte water content was controlled. Surface modification of the fillers improved the stability. ► The observed conductivity decrease varied considerably for various fillers. - Abstract: The electrical conductivity of disperse electrolytes was systematically measured as a function of temperature (0 °C to 60 °C) and filler content for different types of fillers with a range of pore geometry, pore structure and specific surface area. As fillers mesoporous silicas SBA-15, MCM-41 and KIT-6 with pore ranges between 3 nm and 15 nm were dispersed in commercially available liquid lithium electrolytes. As electrolytes 1 M of lithium hexafluorophosphate (LiPF 6 ) in a mixture of ethylene carbonate (EC) and diethylene carbonate (DEC) at the ratio 3:7 (wt/wt) and the same solvent mixture with 0.96 M lithium bis(trifluoromethanesulfon)imide (LiTFSI) were used. No conductivity enhancement could be observed, but with respect to safety aspects the highly viscous disperse pastes might be useful. The conductivity decrease varied considerably for the different fillers.

  9. Ionic liquids and derived materials for lithium and sodium batteries.

    Science.gov (United States)

    Yang, Qiwei; Zhang, Zhaoqiang; Sun, Xiao-Guang; Hu, Yong-Sheng; Xing, Huabin; Dai, Sheng

    2018-03-21

    The ever-growing demand for advanced energy storage devices in portable electronics, electric vehicles and large scale power grids has triggered intensive research efforts over the past decade on lithium and sodium batteries. The key to improve their electrochemical performance and enhance the service safety lies in the development of advanced electrode, electrolyte, and auxiliary materials. Ionic liquids (ILs) are liquids consisting entirely of ions near room temperature, and are characterized by many unique properties such as ultralow volatility, high ionic conductivity, good thermal stability, low flammability, a wide electrochemical window, and tunable polarity and basicity/acidity. These properties create the possibilities of designing batteries with excellent safety, high energy/power density and long-term stability, and also provide better ways to synthesize known materials. IL-derived materials, such as poly(ionic liquids), ionogels and IL-tethered nanoparticles, retain most of the characteristics of ILs while being endowed with other favourable features, and thus they have received a great deal of attention as well. This review provides a comprehensive review of the various applications of ILs and derived materials in lithium and sodium batteries including Li/Na-ion, dual-ion, Li/Na-S and Li/Na-air (O 2 ) batteries, with a particular emphasis on recent advances in the literature. Their unique characteristics enable them to serve as advanced resources, medium, or ingredient for almost all the components of batteries, including electrodes, liquid electrolytes, solid electrolytes, artificial solid-electrolyte interphases, and current collectors. Some thoughts on the emerging challenges and opportunities are also presented in this review for further development.

  10. Effect of a novel amphipathic ionic liquid on lithium deposition in gel polymer electrolytes

    International Nuclear Information System (INIS)

    Choi, Nam-Soon; Koo, Bonjae; Yeon, Jin-Tak; Lee, Kyu Tae; Kim, Dong-Won

    2011-01-01

    Highlights: · Synthesis of a dimeric ionic liquid. · Gel polymer electrolytes providing uniform lithium deposit pathway. · An amphipathic ionic liquid locates at the interface between an electrolyte-rich phase and a polymer matrix in a gel polymer electrolyte. · The presence of PDMITFSI ionic liquid leads to the suppression of dendritic lithium formation on a lithium metal electrode. - Abstract: A novel dimeric ionic liquid based on imidazolium cation and bis(trifluoromethanesulfonyl) imide (TFSI) anion has been synthesized through a metathesis reaction. Its chemical shift values and thermal properties are identified via 1 H nuclear magnetic resonance (NMR) imaging and differential scanning calorimetry (DSC). The effect of the synthesized dimeric ionic liquid on the interfacial resistance of gel polymer electrolytes is described. Differences in the SEM images of lithium electrodes after lithium deposition with and without the 1,1'-pentyl-bis(2,3-dimethylimidazolium) bis(trifluoromethane-sulfonyl)imide (PDMITFSI) ionic liquid in gel polymer electrolytes are clearly discernible. This occurs because the PDMITFSI ionic liquid with hydrophobic moieties and polar groups modulates lithium deposit pathways onto the lithium metal anode. Moreover, high anodic stability for a gel polymer electrolyte with the PDMITFSI ionic liquid was clearly observed.

  11. Effects of Large Area Liquid Lithium Limiters on Spherical Torus Plasmas

    International Nuclear Information System (INIS)

    Kaita, R.; Majeski, R.; Boaz, M.; Efthimion, P.; Gettelfinger, G.; Gray, T.; Hoffman, D.; Jardin, S.; Kugel, H.; Marfuta, P.; Munsat, T.; Neumeyer, C.; Raftopoulos, S.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Delgado-Aparicio, L.; Seraydarian, R.P.; Antar, G.; Doerner, R.; Luckhardt, S.; Baldwin, M.; Conn, R.W.; Maingi, R.; Menon, M.; Causey, R.; Buchenauer, D.; Ulrickson, M.; Jones, B.; Rodgers, D.

    2004-01-01

    Use of a large-area liquid lithium surface as a first wall has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter

  12. Effects of large area liquid lithium limiters on spherical torus plasmas

    International Nuclear Information System (INIS)

    Kaita, R.; Majeski, R.; Boaz, M.; Efthimion, P.; Gettelfinger, G.; Gray, T.; Hoffman, D.; Jardin, S.; Kugel, H.; Marfuta, P.; Munsat, T.; Neumeyer, C.; Raftopoulos, S.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Delgado-Aparicio, L.; Seraydarian, R.P.; Antar, G.; Doerner, R.; Luckhardt, S.; Baldwin, M.; Conn, R.W.; Maingi, R.; Menon, M.; Causey, R.; Buchenauer, D.; Ulrickson, M.; Jones, B.; Rodgers, D.

    2005-01-01

    Use of a large-area liquid lithium surface as a limiter has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter

  13. European research and development programme for water-cooled lithium-lead blankets: present status and future work

    International Nuclear Information System (INIS)

    Giancarli, L.; Leroy, P.; Proust, E.; Raepsaet, X.

    1992-01-01

    The European R and D programme in support of the development of water-cooled Pb-17Li blankets for DEMO aims at improving the data base concerning tritium behaviour and compatibility between blanket materials. The four main areas of the experimental programme are structural material corrosion by Pb-17Li, tritium extraction and permeation control.=, Pb-17Li physico-chemistry, and water/Pb-17Li interaction. This paper describes the most significant results obtained to date in the various experiments performed in Europe and the future programme required to complete the data base by 1994. 28 refs

  14. Investigations on interactions between the flowing liquid lithium limiter and plasmas

    International Nuclear Information System (INIS)

    Ren, J.; Zuo, G.Z.; Hu, J.S.; Sun, Z.; Li, J.G.; Zakharov, L.E.; Ruzic, D.N.; Xu, W.Y.

    2016-01-01

    Two different designs of flowing liquid lithium limiter were first tested for power exhaust and particle removal in HT-7 in 2012 autumn campaign. During the experiments, the reliability and compatibility of the limiters within Tokamak were experimentally demonstrated, and some positive results were achieved. It was found that the flowing liquid lithium limiter was effective for suppressing H concentration and led to a low ratio of H/(H + D). O impurity was slightly decreased by using limiters as well as when using a Li coating. A significant increase of the wall retention ratio was also observed which resulted from the outstanding D particles pumping ability of flowing liquid lithium limiters. The strong interaction between plasma and lithium surface could cause lithium ejection into plasma and lead to disruptions. The stable plasmas produced by uniform Li flow were in favor of lithium control. While the limiters were applied with a uniform Li flow, the normal plasma was easy to be obtained, and the energy confinement time increased from ∼0.025 s to 0.04 s. Furthermore, it was encouraging to note that the application of flowing liquid lithium limiters could further improve the confinement of plasma by ∼10% on the basis of Li coating. These remarkable results will help for the following design of flowing liquid lithium limiter in EAST to improve the plasma operation.

  15. Improved liquid-lithium target for the FMIT facility

    International Nuclear Information System (INIS)

    Miles, R.R.; Greenwell, R.K.; Hassberger, J.A.; Ingham, J.G.

    1982-11-01

    An improved target for the Fusion Materials Irradiation Testing Facility was designed. The purpose of the target is to produce a high neutron flux (10 19 n/m 2 sec) for testing of candidate first wall materials for fusion reactors. The neutrons are produced through a Li(d,n) stripping reaction between accelerated deuterons (35 MeV, 0.1A) and a thin jet of flowing liquid lithium. The target consists of a high speed (approx. 17 m/s), free surface wall jet which is exposed to the high (10 -4 Pa) accelerator vacuum. The energy deposited by the deuteron beam in the lithium is sufficient to heat the jet internally to a maximum temperature of roughly 740 0 C, 430 0 C greater than the saturation temperature at the jet free surface. For this reason, the jet flows along a curved wall which provides the pressurization required to prevent sperheat internal to the jet. Supporting hardware for the jet and a drain line which controls the jet beyond the beam intercept region

  16. 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

  17. The design of a liquid lithium lens for a muon collider

    International Nuclear Information System (INIS)

    Balbekov, V.; Geer, S.; Hassanein, A.; Holtkamp, N.; Lebrun, P.; Neuffer, D.; Norem, J.; Palmer, R.; Reed, C.; Silvestrov, G.; Spentzouris, P.; Tollestrup, A.; Vsevolozhskaya, T. A.

    1999-01-01

    The last stage of ionization cooling for the muon collider requires a multistage liquid lithium lens. This system uses a large (approximately0.5 MA) pulsed current through liquid lithium to focus the beam while energy loss in the lithium removes momentum which is replaced by linacs. The beam optics are designed to maximize the 6 dimensional transmission from one lens to the next while minimizing emittance growth. The mechanical design of the lithium vessel is constrained by a pressure pulse due to the sudden ohmic heating, and the stress on the Be window. The authors describe beam optics, the liquid lithium pressure vessel, pumping, power supplies, as well as the overall optimization of the system

  18. VUV/XUV measurements of impurity emission in plasmas with liquid lithium surfaces on LTX

    International Nuclear Information System (INIS)

    Tritz, Kevin; Finkenthal, Michael; Stutman, Dan; Bell, Ronald E; Boyle, Dennis; Kaita, Robert; Kozub, Tom; Lucia, Matthew; Majeski, Richard; Merino, Enrique; Schmitt, John; Beiersdorfer, Peter; Clementson, Joel; Kubota, Shigeyuki

    2014-01-01

    The VUV/XUV spectrum has been measured on the Lithium Tokamak eXperiment (LTX) using a transmission grating imaging spectrometer (TGIS) coupled to a direct-detection x-ray charge-coupled device camera. TGIS data show significant changes in the ratios between the lithium and oxygen impurity line emission during discharges with varying lithium wall conditions. Lithium coatings that have been passivated by lengthy exposure to significant levels of impurities contribute to a large O/Li ratio measured during LTX plasma discharges. Furthermore, previous results have indicated that a passivated lithium film on the plasma facing components will function as a stronger impurity source when in the form of a hot liquid layer compared to a solid lithium layer. However, recent TGIS measurements of plasma discharges in LTX with hot stainless steel boundary shells and a fresh liquid lithium coating show lower O/Li impurity line ratios when compared to discharges with a solid lithium film on cool shells. These new measurements help elucidate the somewhat contradictory results of the effects of solid and liquid lithium on plasma confinement observed in previous experiments. (paper)

  19. Extraction of lithium from salt lake brine using room temperature ionic liquid in tributyl phosphate

    International Nuclear Information System (INIS)

    Shi, Chenglong; Jia, Yongzhong; Zhang, Chao; Liu, Hong; Jing, Yan

    2015-01-01

    Highlights: • We proposed a new system for Li recovery from salt lake brine by extraction using an ionic liquid. • Cation exchange was proposed to be the mechanism of extraction followed in ionic liquid. • This ionic liquid system shown considerable extraction ability for lithium and the single extraction efficiency of lithium reached 87.28% under the optimal conditions. - Abstract: Lithium is known as the energy metal and it is a key raw material for preparing lithium isotopes which have important applications in nuclear energy source. In this work, a typical room temperature ionic liquid (RTILs), 1-butyl-3-methyl-imidazolium hexafluorophosphate ([C 4 mim][PF 6 ]), was used as an alternative solvent to study liquid/liquid extraction of lithium from salt lake brine. In this system, the ionic liquid, NaClO 4 and tributyl phosphate (TBP) were used as extraction medium, co-extraction reagent and extractant respectively. The effects of solution pH value, phase ratio, ClO 4 − amount and other factors on lithium extraction efficiency had been investigated. Optimal extraction conditions of this system include the ratio of TBP/IL at 4/1 (v/v), O/A at 2:1, n(ClO 4 − )/n(Li + ) at 2:1, the equilibration time of 10 min and unadjusted pH. Under the optimal conditions, the single extraction efficiency of lithium was 87.28% which was much higher than the conventional extraction system. Total extraction efficiency of 99.12% was obtained by triple-stage countercurrent extraction. Study on the mechanism revealed that the use of ionic liquid increased the extraction yield of lithium through cation exchange in this system. Preliminary results indicated that the use of [C 4 mim][PF 6 ] as an alternate solvent to replace traditional organic solvents (VOCs) in liquid/liquid extraction was very promising

  20. 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

  1. Hydrophobic ionic liquids based on the 1-butyl-3-methylimidazolium cation for lithium/seawater batteries

    Science.gov (United States)

    Zhang, Yancheng; Urquidi-Macdonald, Mirna

    Two hydrophobic ionic liquids (room temperature molten salts) based on 1-butyl-3-methylimidazolium cation (BMI +), BMI +PF 6- and BMI +Tf 2N -, were used in developing a highly efficient lithium anode system for lithium/seawater batteries. The lithium anode system was composed of lithium metal/ionic liquid/Celgard membrane. Both BMI +PF 6-and BMI +Tf 2N - maintained high apparent anodic efficiency (up to 100%) under potentiostatic polarization (at +0.5 V versus open-circuit potential (OCP)) in a 3% NaCl solution. Eventually, traces of water contaminated the ionic liquid and a bilayer film (LiH and LiOH) on the lithium surface was formed, decreasing the rate of lithium anodic reaction and hence the discharge current density. BMI +Tf 2N - prevented traces of water from reaching the lithium metal surface longer than BMI +PF 6- (60 h versus 7 h). However, BMI +PF 6- was better than BMI +Tf 2N - in keeping a constant current density (˜0.2 mA cm -2) before the traces of water contaminated the lithium surface due to the non-reactivity of BMI +PF 6- with the lithium metal that kept the bare lithium surface. During the discharge process, BMI +PF 6- and BMI +Tf 2N - acted as ion transport media of Li +, Cl -, OH - and H 2O, but did not react with them because of the excellent chemical stability, high conductivity, and high hydrophobicity of these two ionic liquids. Both BMI +PF 6- and BMI +Tf 2N - gels were tentative approaches used to delay the traces of water coming in contact with the lithium surface.

  2. Ion transport properties of lithium ionic liquids and their ion gels

    International Nuclear Information System (INIS)

    Shobukawa, Hitoshi; Tokuda, Hiroyuki; Susan, Md. Abu Bin Hasan; Watanabe, Masayoshi

    2005-01-01

    A new series of lithium ionic liquids were prepared by introducing of two electron-withdrawing trifluoroacetyl groups in borate salts containing two methoxy-oligo(ethylene oxide) groups in the structures. Successive substitution reactions of oligo-ethylene glycol monomethyl ether and trifluroacetic acid from LiBH 4 yielded the lithium salts, which were clear and colorless liquids at room temperature. The fundamental physicochemical properties, such as density, thermal property, viscosity, ionic conductivity, self-diffusion coefficients, and electrochemical stability, were measured. The lithium ionic liquids had self-dissociation ability and conducted ions even in the absence of organic solvents. New polymer electrolytes, named 'ion gels', were prepared by radical cross-linking reactions of a poly(ethylene oxide-co-propylene oxide)tri-acrylate macromonomer in the presence the lithium ionic liquid. An increase in the glass transition temperatures (T g ) of the ion gels was very small even with increasing lithium ionic liquid concentration, and the T g 's were lower than that of the ionic liquid itself. The ionic conductivity of the ion gels surpassed that of the lithium ionic liquid in the bulk at certain compositions

  3. Hydrogen and helium recycling from stirred liquid lithium under steady state plasma bombardment

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, Yoshi, E-mail: hirooka.yoshihiko@nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); The Graduate School for Advanced Studies, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Zhou, Haishan [The Graduate School for Advanced Studies, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Ono, Masa [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2014-12-15

    For improved core performance via edge plasma-wall boundary control, solid and liquid lithium has been used as a plasma-facing material in a number of confinement experiments over the past several decades. Unfortunately, it is unavoidable that lithium is saturated in the surface region with implanted hydrogenic species as well as oxygen-containing impurities. For steady state operation, a flowing liquid lithium divertor with forced convection would probably be required. In the present work, the effects of liquid stirring to simulate forced convection have been investigated on the behavior of hydrogen and helium recycling from molten lithium at temperatures up to ∼350 °C. Data indicate that liquid stirring reactivates hydrogen pumping via surface de-saturation and/or uncovering impurity films, but can also induce helium release via surface temperature change.

  4. Knight shift of 23Na and 7Li nuclei in liquid sodium-lithium alloys

    International Nuclear Information System (INIS)

    Feitsma, P.D.

    1977-01-01

    The Knight shift of 23 Na and 7 Li nuclei in liquid sodium-lithium alloys has been measured. Some aspects of the theoretical interpretation of the Knight shift within the diffraction model, are clarified

  5. Some safety considerations of liquid lithium as a fusion breeder material

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Muhlestein, L.D.

    1986-01-01

    Test results and conclusions are presented for the reaction of steam with a high temperature lithium pool and for the reaction of high temperature lithium spray with a nitrogen atmosphere. The reactions are characterized and evaluated in regard to the potential for mobilization of radioactive species associated with the liquid breeder under postulated fusion reactor accident conditions. These evaluations include measured lithium temperature responses, atmosphere temperature and pressure responses, gas consumption and generation, aerosol quantities and particle size characterization, and potentially radioactive species releases. Conclusions are made as to the consequences of these safety considerations for the use of lithium as a fusion reactor breeder material

  6. Lithium test module on ITER: engineering design of the tritium recovery system

    International Nuclear Information System (INIS)

    Finn, P.A.

    1988-01-01

    The design presented is an overview of the tritium recovery system for a lithium module on an ITER type reactor. The design of a tritium recovery system for larger blanket units, sectors, etc. could use the information developed in this report. A goal of this design was to ensure that a reliable, integrated performance of the tritium recovery system could be demonstrated. An equally important goal was to measure and account for the tritium in the liquid lithium blanket module and its recovery system in order to validate the operation of the blanket module

  7. Ionic Liquid-Nanoparticle Hybrid Electrolytes and their Application in Secondary Lithium-Metal Batteries

    KAUST Repository

    Lu, Yingying

    2012-07-12

    Ionic liquid-tethered nanoparticle hybrid electrolytes comprised of silica nanoparticles densely grafted with imidazolium-based ionic liquid chains are shown to retard lithium dendrite growth in rechargeable batteries with metallic lithium anodes. The electrolytes are demonstrated in full cell studies using both high-energy Li/MoS2 and high-power Li/TiO2 secondary batteries. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  8. An induction-based magnetohydrodynamic 3D code for finite magnetic Reynolds number liquid-metal flows in fusion blankets

    International Nuclear Information System (INIS)

    Kawczynski, Charlie; Smolentsev, Sergey; Abdou, Mohamed

    2016-01-01

    Highlights: • A new induction-based magnetohydrodynamic code was developed using a finite difference method. • The code was benchmarked against purely hydrodynamic and MHD flows for low and finite magnetic Reynolds number. • Possible applications of the new code include liquid-metal MHD flows in the breeder blanket during unsteady events in the plasma. - Abstract: Most numerical analysis performed in the past for MHD flows in liquid-metal blankets were based on the assumption of low magnetic Reynolds number and involved numerical codes that utilized electric potential as the main electromagnetic variable. One limitation of this approach is that such codes cannot be applied to truly unsteady processes, for example, MHD flows of liquid-metal breeder/coolant during unsteady events in plasma, such as major plasma disruptions, edge-localized modes and vertical displacements, when changes in plasmas occur at millisecond timescales. Our newly developed code MOONS (Magnetohydrodynamic Object-Oriented Numerical Solver) uses the magnetic field as the main electromagnetic variable to relax the limitations of the low magnetic Reynolds number approximation for more realistic fusion reactor environments. The new code, written in Fortran, implements a 3D finite-difference method and is capable of simulating multi-material domains. The constrained transport method was implemented to evolve the magnetic field in time and assure that the magnetic field remains solenoidal within machine accuracy at every time step. Various verification tests have been performed including purely hydrodynamic flows and MHD flows at low and finite magnetic Reynolds numbers. Test results have demonstrated very good accuracy against known analytic solutions and other numerical data.

  9. An induction-based magnetohydrodynamic 3D code for finite magnetic Reynolds number liquid-metal flows in fusion blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kawczynski, Charlie; Smolentsev, Sergey, E-mail: sergey@fusion.ucla.edu; Abdou, Mohamed

    2016-11-01

    Highlights: • A new induction-based magnetohydrodynamic code was developed using a finite difference method. • The code was benchmarked against purely hydrodynamic and MHD flows for low and finite magnetic Reynolds number. • Possible applications of the new code include liquid-metal MHD flows in the breeder blanket during unsteady events in the plasma. - Abstract: Most numerical analysis performed in the past for MHD flows in liquid-metal blankets were based on the assumption of low magnetic Reynolds number and involved numerical codes that utilized electric potential as the main electromagnetic variable. One limitation of this approach is that such codes cannot be applied to truly unsteady processes, for example, MHD flows of liquid-metal breeder/coolant during unsteady events in plasma, such as major plasma disruptions, edge-localized modes and vertical displacements, when changes in plasmas occur at millisecond timescales. Our newly developed code MOONS (Magnetohydrodynamic Object-Oriented Numerical Solver) uses the magnetic field as the main electromagnetic variable to relax the limitations of the low magnetic Reynolds number approximation for more realistic fusion reactor environments. The new code, written in Fortran, implements a 3D finite-difference method and is capable of simulating multi-material domains. The constrained transport method was implemented to evolve the magnetic field in time and assure that the magnetic field remains solenoidal within machine accuracy at every time step. Various verification tests have been performed including purely hydrodynamic flows and MHD flows at low and finite magnetic Reynolds numbers. Test results have demonstrated very good accuracy against known analytic solutions and other numerical data.

  10. Diagnostics of high-speed liquid lithium jet for IFMIF/EVEDA lithium test loop

    International Nuclear Information System (INIS)

    Kanemura, Takuji; Kondo, Hiroo; Furukawa, Tomohiro; Sugiura, Hirokazu; Horiike, Hiroshi; Yamaoka, Nobuo; Ida, Mizuho; Nakamura, Kazuyuki; Matsushita, Izuru

    2011-01-01

    Regarding R and Ds on the International Fusion Materials Irradiation Facility (IFMIF), hydraulic stability of the liquid Li jet simulating the IFMIF Li target is planned to be validated using EVEDA Li Test Loop (ELTL). IFMIF is an accelerator-based deuteron-lithium (Li) neutron source for research and development of fusion reactor materials. The stable Li target is required in IFMIF to maintain the quality of the neutron fluence and integrity of the Li target itself. This paper presents diagnostics of the Li jet to be implemented in validation tests of the jet stability in ELTL, and those specifications and methodologies are introduced. In the tests, the following physical parameters need to be measured; thickness of the jet; surface structure (height, length/width and frequency of free-surface waves); local flow velocity at the free surface; and Li evaporation rate. With regard to measurement of jet thickness and the surface wave height, a contact-type liquid level sensor is to be used. As for measurement of wave velocity and visual understanding of detailed free-surface structure, a high-speed video camera is to be leveraged. With respect to Li evaporation measurement, weight change of specimens installed near the free surface and frequency change of a crystal quartz are utilized. (author)

  11. 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

  12. High-power liquid-lithium jet target for neutron production

    OpenAIRE

    Halfon, S.; Arenshtam, A.; Kijel, D.; Paul, M.; Berkovits, D.; Eliyahu, I.; Feinberg, G.; Friedman, M.; Hazenshprung, N.; Mardor, I.; Nagler, A.; Shimel, G.; Tessler, M.; Silverman, I.

    2013-01-01

    A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center. The lithium target, to be bombarded by the high-intensity proton beam of the Soreq Applied Research Accelerator Facility (SARAF), will constitute an intense source of neutrons produced by the 7Li(p,n)7Be reaction for nuclear astrophysics research and as a pilot setup for accelerator-based Boron Neutron Capture Therapy (BNCT). The liquid-lithium jet target acts both as ...

  13. Symmetric lithium-ion cell based on lithium vanadium fluorophosphate with ionic liquid electrolyte

    International Nuclear Information System (INIS)

    Plashnitsa, Larisa S.; Kobayashi, Eiji; Okada, Shigeto; Yamaki, Jun-ichi

    2011-01-01

    Lithium vanadium fluorophosphate, LiVPO 4 F, was utilized as both cathode and anode for fabrication of a symmetric lithium-ion LiVPO 4 F//LiVPO 4 F cell. The electrochemical evolution of the LiVPO 4 F//LiVPO 4 F cell with the commonly used organic electrolyte LiPF 6 /EC-DMC has shown that this cell works as a secondary battery, but exhibits poor durability at room temperature and absolutely does not work at increased operating temperatures. To improve the performance and safety of this symmetric battery, we substituted a non-flammable ionic liquid (IL) LiBF 4 /EMIBF 4 electrolyte for the organic electrolyte. The symmetric battery using the IL electrolyte was examined galvanostatically at different rates and operating temperatures within the voltage range of 0.01-2.8 V. It was demonstrated that the IL-based symmetric cell worked as a secondary battery with a Coulombic efficiency of 77% at 0.1 mA cm -2 and 25 o C. It was also found that the use of the IL electrolyte instead of the organic one resulted in the general reduction of the first discharge capacity by about 20-25% but provided much more stable behavior and a longer cycle life. Moreover, an increase of the discharge capacity of the IL-based symmetric battery up to 120 mA h g -1 was observed when the operating temperature was increased up to 80 o C at 0.1 mA cm -2 . The obtained electrochemical behavior of both symmetric batteries was confirmed by complex-impedance measurements at different temperatures and cycling states. The thermal stability of LiVPO 4 F with both the IL and organic electrolytes was also examined.

  14. Compositional depth profiles of the type 316 stainless steel undergone the corrosion in liquid lithium using laser-induced breakdown spectroscopy

    Science.gov (United States)

    Li, Ying; Ke, Chuan; Liu, Xiang; Gou, Fujun; Duan, Xuru; Zhao, Yong

    2017-12-01

    Liquid metal lithium cause severe corrosion on the surface of metal structure material that used in the blanket and first wall of fusion device. Fast and accurate compositional depth profile measurement for the boundary layer of the corroded specimen will reveal the clues for the understanding and evaluation of the liquid lithium corrosion process as well as the involved corrosion mechanism. In this work, the feasibility of laser-induced breakdown spectroscopy for the compositional depth profile analysis of type 316 stainless steel which was corroded by liquid lithium in certain conditions was demonstrated. High sensitivity of LIBS was revealed especially for the corrosion medium Li in addition to the matrix elements of Fe, Cr, Ni and Mn by the spectral analysis of the plasma emission. Compositional depth profile analysis for the concerned elements which related to corrosion was carried out on the surface of the corroded specimen. Based on the verified local thermodynamic equilibrium shot-by-shot along the depth profile, the matrix effect was evaluated as negligible by the extracted physical parameter of the plasmas generated by each laser pulse in the longitudinal depth profile. In addition, the emission line intensity ratios were introduced to further reduce the impact on the emission line intensity variations arise from the strong inhomogeneities on the corroded surface. Compositional depth profiles for the matrix elements of Fe, Cr, Ni, Mn and the corrosion medium Li were constructed with their measured relative emission line intensities. The distribution and correlations of the concerned elements in depth profile may indicate the clues to the complicated process of composition diffusion and mass transfer. The results obtained demonstrate the potentiality of LIBS as an effective technique to perform spectrochemical measurement in the research fields of liquid metal lithium corrosion.

  15. Extraction of lithium ion from alkaline aqueous media by a liquid surfactant membrane

    International Nuclear Information System (INIS)

    Kinugasa, Takumi; Ono, Yuri; Kawamura, Yuko; Watanabe, Kunio; Takeuchi, Hiroshi.

    1995-01-01

    Extraction of lithium ion from aqueous alkaline media by a liquid surfactant membrane was performed using a mixture of LIX54 and TOPO as the extractant. Stripping of lithium from the kerosene solution to the acid solution was suppressed with increasing content of polyamine (ECA) surfactant. The extraction rate of lithium by the liquid membrane could be interpreted taking account of an interfacial resistance due to ECA. It was confirmed that swelling of the (W/O) emulsion drops by water permeation through the liquid membrane is evaluated in terms of a change in osmotic pressure gradient between the external and internal aqueous phases during the lithium extraction. In the present operation, the extraction ratio of Li + from the external feed and the uptake into the internal phase reached as high as 95%. (author)

  16. Ferrous alloy metallurgy, liquid lithium corrosion and welding. Final report, April 1, 1973-March 31, 1984

    International Nuclear Information System (INIS)

    Olson, D.L.; Matlock, D.K.

    1984-01-01

    This research program consists of two parts: an evaluation of the corrosion behavior of ferrous alloys in liquid lithium, and a study of microstructure development and properties of dissimilar metal weldments. A ten-year overview of the research accomplishments made is presented. The effects of liquid lithium on both uniform corrosion and grain boundary penetration in ferrous alloys has been investigated as a function of time, temperature, base metal alloy content, and liquid lithium nitrogen content. Kinetic equations for the various corrosion processes have been developed and analyzed with respect to models for corrosion and corrosion product development. The effects of liquid lithium on mechanical properties, particularly fatigue, have been studied. Results have shown that in both austenitic stainless steels and ferritic steels, liquid lithium significantly reduces the mechanical integrity of all materials by inducing liquid metal embrittlement. A model for liquid metal embrittlement induced damage during fatigue was developed and shown to correlate with the experimental results. Microstructural development in austenitic weld metal, with particular emphasis on new grades with reduced chromium contents, has been investigated. The microstructures have been correlated with alloy content and the basics of a thermodynamic model for predicting weld metal microstructure has been developed. The high temperature mechanical behavior of dissimilar metal weldments (austenitic stainless steel to ferritic steel) has been investigated with the impression-creep test technique. Observed microstructural changes with position across the weldment are shown to correlate directly with creep behavior. A model based on deformation of composite materials was developed

  17. Code development for analysis of MHD pressure drop reduction in a liquid metal blanket using insulation technique based on a fully developed flow model

    International Nuclear Information System (INIS)

    Smolentsev, Sergey; Morley, Neil; Abdou, Mohamed

    2005-01-01

    The paper presents details of a new numerical code for analysis of a fully developed MHD flow in a channel of a liquid metal blanket using various insulation techniques. The code has specially been designed for channels with a 'sandwich' structure of several materials with different physical properties. The code includes a finite-volume formulation, automatically generated Hartmann number sensitive meshes, and effective convergence acceleration technique. Tests performed at Ha ∼ 10 4 have showed very good accuracy. As an illustration, two blanket flows have been considered: Pb-17Li flow in a channel with a silicon carbide flow channel insert, and Li flow in a channel with insulating coating

  18. Spherical Torus Plasma Interactions with Large-area Liquid Lithium Surfaces in CDX-U

    International Nuclear Information System (INIS)

    Kaita, R.; Majeski, R.; Boaz, M.; Efthimion, P.; Jones, B.; Hoffman, D.; Kugel, H.; Menard, J.; Munsat, T.; Post-Zwicker, A.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Antar, G.; Doerner, R.; Luckhardt, S.; Maingi, R.; Maiorano, M.; Smith, S.

    2002-01-01

    The Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego. Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance

  19. Spherical Torus Plasma Interactions with Large-area Liquid Lithium Surfaces in CDX-U

    Energy Technology Data Exchange (ETDEWEB)

    R. Kaita; R. Majeski; M. Boaz; P. Efthimion; B. Jones; D. Hoffman; H. Kugel; J. Menard; T. Munsat; A. Post-Zwicker; V. Soukhanovskii; J. Spaleta; G. Taylor; J. Timberlake; R. Woolley; L. Zakharov; M. Finkenthal; D. Stutman; G. Antar; R. Doerner; S. Luckhardt; R. Maingi; M. Maiorano; S. Smith

    2002-01-18

    The Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego. Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance.

  20. High-efficiency technology for lithium isotope separation using an ionic-liquid impregnated organic membrane

    International Nuclear Information System (INIS)

    Hoshino, Tsuyoshi; Terai, Takayuki

    2011-01-01

    The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 ( 6 Li) in tritium breeding materials. New lithium isotope separation technique using ionic-liquid impregnated organic membranes (Ionic-Liquid-i-OMs) have been developed. Lithium ions are able to move by electrodialysis through certain Ionic-Liquid-i-OMs between the cathode and the anode in lithium solutions. In this report, the effects of protection cover and membrane thickness on the durability of membrane and the efficiency of isotope separation were evaluated. In order to improve the durability of the Ionic-Liquid-i-OM, we developed highly-durable Ionic-Liquid-i-OM. Both surfaces of the Ionic-Liquid-i-OM were covered by a nafion 324 overcoat or a cation exchange membrane (SELEMION TM CMD) to prevent the outflow of the ionic liquid. It was observed that the durability of the Ionic-Liquid-i-OM was improved by a nafion 324 overcoat. On the other hand, the organic membrane selected was 1, 2 or 3 mm highly-porous Teflon film, in order to efficiently impregnate the ionic liquid. The 6 Li isotope separation factor by electrodialysis using highly-porous Teflon film of 3 mm thickness was larger than using that of 1 or 2 mm thickness.

  1. 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.

  2. Ionic Liquid-Doped Gel Polymer Electrolyte for Flexible Lithium-Ion Polymer Batteries

    Science.gov (United States)

    Zhang, Ruisi; Chen, Yuanfen; Montazami, Reza

    2015-01-01

    Application of gel polymer electrolytes (GPE) in lithium-ion polymer batteries can address many shortcomings associated with liquid electrolyte lithium-ion batteries. Due to their physical structure, GPEs exhibit lower ion conductivity compared to their liquid counterparts. In this work, we have investigated and report improved ion conductivity in GPEs doped with ionic liquid. Samples containing ionic liquid at a variety of volume percentages (vol %) were characterized for their electrochemical and ionic properties. It is concluded that excess ionic liquid can damage internal structure of the batteries and result in unwanted electrochemical reactions; however, samples containing 40–50 vol % ionic liquid exhibit superior ionic properties and lower internal resistance compared to those containing less or more ionic liquids.

  3. Ionic Liquid-Doped Gel Polymer Electrolyte for Flexible Lithium-Ion Polymer Batteries

    Directory of Open Access Journals (Sweden)

    Ruisi Zhang

    2015-05-01

    Full Text Available Application of gel polymer electrolytes (GPE in lithium-ion polymer batteries can address many shortcomings associated with liquid electrolyte lithium-ion batteries. Due to their physical structure, GPEs exhibit lower ion conductivity compared to their liquid counterparts. In this work, we have investigated and report improved ion conductivity in GPEs doped with ionic liquid. Samples containing ionic liquid at a variety of volume percentages (vol % were characterized for their electrochemical and ionic properties. It is concluded that excess ionic liquid can damage internal structure of the batteries and result in unwanted electrochemical reactions; however, samples containing 40–50 vol % ionic liquid exhibit superior ionic properties and lower internal resistance compared to those containing less or more ionic liquids.

  4. 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

  5. Water-cooled lithium-lead box-shaped blanket concept for Demo: thermo-mechanical optimization and manufacturing sequence proposal

    International Nuclear Information System (INIS)

    Baraer, L.; Dinot, N.; Giancarli, L.; Proust, E.; Salavy, J.F.; Severi, Y.; Quintric-Bossy, J.

    1992-01-01

    The development of the water-cooled lithium-lead box-shaped blanket concept for DEMO has now reached the stage of thermo-mechanical optimization. In the previous design phases the preliminary dimensioning of the cooling circuit has permitted to define the water proportions required in the breeder region and to demonstrate, after a minimization of steel proportion and thicknesses, that this concept could reach tritium breeding self-sufficiency. In the present analysis the location of the coolant pipes has been optimized for the whole equatorial plane cross-section of both inboard and outboard segments in order to maintain the maximum Pb-17Li/steel interface temperature below 480 deg C and to minimize the thermal gradients along the steel structures. The consequent thermo-mechanical analysis has shown that the thermal stresses always remain below the allowable limits. Segment fabricability and removal are the next design issues to be analyzed. Within this strategy, a first manufactury sequence for the outboard segment is proposed

  6. Experiments with liquid metal walls: Status of the lithium tokamak experiment

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, Robert, E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Berzak, Laura; Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Hammett, Gregory; Jacobson, Craig M.; Jones, Andrew; Kozub, Thomas; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Lucia, Matthew; Lundberg, Daniel; Majeski, Richard; Mansfield, Dennis; Menard, Jonathan; Spaleta, Jeffrey; Strickler, Trevor; Timberlake, John [Princeton Plasma Physics Laboratory, Princeton, NJ (United States)

    2010-11-15

    Abstarct: Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The lithium tokamak experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the current drive experiment-upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in ohmically heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy (CHERS). Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions.

  7. Experiments with Liquid Metal Walls: Status of the Lithium Tokamak Experiment

    International Nuclear Information System (INIS)

    Kaita, Robert; Berzak, Laura; Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Hammett, Gregory; Jacobson, Craig M.; Jones, Andrew; Kozub, Thomas; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Lucia, Matthew; Lundberg, Daniel; Majeski, Richard; Mansfield, Dennis; Menard, Jonathan; Spaleta, Jeffrey; Strickler, Trevor; Timberlak, John

    2010-01-01

    Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy. Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions.

  8. Verification of the hydraulic design of the FMIT liquid lithium target

    International Nuclear Information System (INIS)

    Miles, R.R.; Annese, C.E.; Ingham, J.G.

    1983-01-01

    A liquid lithium target is being developed to generate a neutron flux for material testing in a fusion-like environment. The target consists of a thin, high speed, curved wall jet of lithium which is formed by an asymmetric nozzle. A prototype target was designed using potential flow analysis and was tested in water. Measurements of jet thickness and velocity in water and thickness in lithium were compared with isothermal design predictions and were shown to match within 1% for thickness and 5% for jet velocity

  9. High-flux neutron source based on a liquid-lithium target

    Science.gov (United States)

    Halfon, S.; Feinberg, G.; Paul, M.; Arenshtam, A.; Berkovits, D.; Kijel, D.; Nagler, A.; Eliyahu, I.; Silverman, I.

    2013-04-01

    A prototype compact Liquid Lithium Target (LiLiT), able to constitute an accelerator-based intense neutron source, was built. The neutron source is intended for nuclear astrophysical research, boron neutron capture therapy (BNCT) in hospitals and material studies for fusion reactors. The LiLiT setup is presently being commissioned at Soreq Nuclear research Center (SNRC). The lithium target will produce neutrons through the 7Li(p,n)7Be reaction and it will overcome the major problem of removing the thermal power generated by a high-intensity proton beam, necessary for intense neutron flux for the above applications. The liquid-lithium loop of LiLiT is designed to generate a stable lithium jet at high velocity on a concave supporting wall with free surface toward the incident proton beam (up to 10 kW). During off-line tests, liquid lithium was flown through the loop and generated a stable jet at velocity higher than 5 m/s on the concave supporting wall. The target is now under extensive test program using a high-power electron-gun. Up to 2 kW electron beam was applied on the lithium flow at velocity of 4 m/s without any flow instabilities or excessive evaporation. High-intensity proton beam irradiation will take place at SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator currently in commissioning at SNRC.

  10. 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

  11. 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

  12. Materials science problems of blankets in Russian concept of fusion reactor

    International Nuclear Information System (INIS)

    Solonin, M.I.

    1998-01-01

    Structural materials, beryllium and tritium breeding materials proposed for blanket of Russian reactor DEMO and Test Modules for ITER are discussed. Main requirements for the materials are concerned with basis current designs of blankets and modules and possibility meet of ones for presence and developed alloys and materials discussed considered. Main properties and results of test of ferrite-martensite and vanadium alloys for DEMO and Test Modules are cited. Beryllium compositions used as component of first wall and neutron multiplier are discussed. Liquid lithium and ceramic (lithium orthosilicate) are treated as tritium breeding materials. Russian development of reactor experimental unit for tritium breeding zone using beryllium, lithium ceramic and ferrite-martensite alloys for structural materials is presented. (orig.)

  13. Extraction of lithium from salt lake brine using room temperature ionic liquid in tributyl phosphate

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Chenglong [Key Laboratory of Salt Lake Resources and Chemistry, Qinghai Institute of Salt Lakes, Chinese Academy of Sciences, 810008 Xining (China); University of Chinese Academy of Sciences, 100049 Beijing (China); Jia, Yongzhong [Key Laboratory of Salt Lake Resources and Chemistry, Qinghai Institute of Salt Lakes, Chinese Academy of Sciences, 810008 Xining (China); Zhang, Chao [Key Laboratory of Salt Lake Resources and Chemistry, Qinghai Institute of Salt Lakes, Chinese Academy of Sciences, 810008 Xining (China); University of Chinese Academy of Sciences, 100049 Beijing (China); Liu, Hong [Qinghai Salt Chemical Products Supervision and Inspection Center, 816000 Golmud (China); Jing, Yan, E-mail: 1580707906@qq.com [Key Laboratory of Salt Lake Resources and Chemistry, Qinghai Institute of Salt Lakes, Chinese Academy of Sciences, 810008 Xining (China)

    2015-01-15

    Highlights: • We proposed a new system for Li recovery from salt lake brine by extraction using an ionic liquid. • Cation exchange was proposed to be the mechanism of extraction followed in ionic liquid. • This ionic liquid system shown considerable extraction ability for lithium and the single extraction efficiency of lithium reached 87.28% under the optimal conditions. - Abstract: Lithium is known as the energy metal and it is a key raw material for preparing lithium isotopes which have important applications in nuclear energy source. In this work, a typical room temperature ionic liquid (RTILs), 1-butyl-3-methyl-imidazolium hexafluorophosphate ([C{sub 4}mim][PF{sub 6}]), was used as an alternative solvent to study liquid/liquid extraction of lithium from salt lake brine. In this system, the ionic liquid, NaClO{sub 4} and tributyl phosphate (TBP) were used as extraction medium, co-extraction reagent and extractant respectively. The effects of solution pH value, phase ratio, ClO{sub 4}{sup −} amount and other factors on lithium extraction efficiency had been investigated. Optimal extraction conditions of this system include the ratio of TBP/IL at 4/1 (v/v), O/A at 2:1, n(ClO{sub 4}{sup −})/n(Li{sup +}) at 2:1, the equilibration time of 10 min and unadjusted pH. Under the optimal conditions, the single extraction efficiency of lithium was 87.28% which was much higher than the conventional extraction system. Total extraction efficiency of 99.12% was obtained by triple-stage countercurrent extraction. Study on the mechanism revealed that the use of ionic liquid increased the extraction yield of lithium through cation exchange in this system. Preliminary results indicated that the use of [C{sub 4}mim][PF{sub 6}] as an alternate solvent to replace traditional organic solvents (VOCs) in liquid/liquid extraction was very promising.

  14. Lithium

    Science.gov (United States)

    Bradley, Dwight C.; Stillings, Lisa L.; Jaskula, Brian W.; Munk, LeeAnn; McCauley, Andrew D.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Lithium, the lightest of all metals, is used in air treatment, batteries, ceramics, glass, metallurgy, pharmaceuticals, and polymers. Rechargeable lithium-ion batteries are particularly important in efforts to reduce global warming because they make it possible to power cars and trucks from renewable sources of energy (for example, hydroelectric, solar, or wind) instead of by burning fossil fuels. Today, lithium is extracted from brines that are pumped from beneath arid sedimentary basins and extracted from granitic pegmatite ores. The leading producer of lithium from brine is Chile, and the leading producer of lithium from pegmatites is Australia. Other potential sources of lithium include clays, geothermal brines, oilfield brines, and zeolites. Worldwide resources of lithium are estimated to be more than 39 million metric tons, which is enough to meet projected demand to the year 2100. The United States is not a major producer at present but has significant lithium resources.

  15. Theoretical study and experimental detection of cavitation phenomena in Liquid Lithium Target Facility for IFMIF

    International Nuclear Information System (INIS)

    Orco, G. Dell; Horiike, H.; Ida, M.; Nakamura, H.

    2006-01-01

    In the IFMIF (International Fusion Materials Irradiation Facility) testing facility, the required high energy neutrons emission will be produced by reaction of two D + beams with a free surface liquid Lithium jet target flowing along concave back-wall at 20 m/s. The Lithium height in the experimental loop and its relevant static pressure, the high flow velocities and the presence of several devices for the flow control and the pressure reduction increase the risk of cavitation onset in the target system. Special attention has to be taken in the primary pump, in the flow straightener, in the nozzle and their interconnections where the local pressure decreases and/or velocity increases or flow separations could promote the emission of cavitation vapour bubbles. The successive bubble re-implosions, in the higher pressure liquid bulk, could activate material erosion and transportation of activated particulates. These bubbles, if emitted close to the free jet flow, could also procure hydraulic instability and disturbance of the neutron field in the D + beams-Lithium target zone. Therefore, the cavitation risk must be properly foreseen along the whole IFMIF Lithium target circuit and its occurrence at different operating condition should be also monitored by special instrumentation. ENEA, in close cooperation with JAEA, has demonstrated the capability to detect the onset of the cavitation noises in liquid Lithium, by using the ENEA patented accelerometric gauge called CASBA-2000, during hydraulic test campaigns carried-out at Osaka University Lithium facility on a straight mock-up of the IFMIF back plate target. Comparison with the Thoma' cavitation similitude criteria have also determined the critical threshold limit for the estimation of the onset. Theoretical study on the conditions of cavitations generation in the IFMIF Lithium Target Circuit were also launched between ENEA and JAEA aiming at analysing the risk of the cavitation occurrence in the Lithium flow by

  16. 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

  17. Thermal Aging of Anions in Ionic Liquids containing Lithium Salts by IC/ESI-MS

    International Nuclear Information System (INIS)

    Pyschik, Marcelina; Kraft, Vadim; Passerini, Stefano; Winter, Martin; Nowak, Sascha

    2014-01-01

    Highlights: • Thermal aging investigation of TFSI- and FSI- based ionic liquids and their mixtures with Li salts. • PYR 13 FSI shows thermal decomposition when mixed with LiPF 6 and LiClO 4 . • PYR 13 TFSI does not show any decomposition products with the electrolyte salts. • LiPF 6 dissolved in ionic liquids suffers of thermal aging as in conventional Li-ion battery electrolytes. - Abstract: The stability of 1-methyl-1-propylpyrrolidinium bis(trifluoromethanesulfonyl)imide (PYR 13 TFSI) and 1-methyl-1-propylpyrrolidinium bis(fluorosulfonyl)imide (PYR 13 FSI) ionic liquids at elevated temperatures (60 °C) is investigated by ion chromatography. Additionally, the influence of the electrolyte salts, lithium hexafluorophosphate (LiPF 6 ), lithium bis(trifluoromethanesulfonyl)imide (LiTFSI) and lithium perchlorate (LiClO 4 ), on the decomposition of both the ionic liquids was analysed over a long term stability study. It has been found out that TFSI has a much higher thermal stability than FSI. The addition of LiTFSI did not show any effect on the aging of both ionic liquid anions. However, PYR 13 FSI degraded when mixed with the electrolyte salts LiPF 6 and LiClO 4 , while PYR 13 TFSI did not. Finally, LiPF 6 forms the same hydrolysis products in the investigated ionic liquids as in the commonly used electrolytes based on organic solvents in lithium-ion batteries

  18. Design, calculation and experimental studies for liquid metal system main parameters in support of the liquid lithium fusion reactor

    International Nuclear Information System (INIS)

    Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.

    2001-01-01

    A new concept of a Liquid Lithium Fusion Reactor and the first experimental results were presented at the 16th IAEA Conference on Fusion Energy. During the past two years theoretical estimations have been made, and calculated and experimental results have been obtained in confirmation of this concept and supporting its progress. The main results of this work are given in the paper. (author)

  19. Design, calculation and experimental studies for liquid metal system main parameters in support of the liquid lithium fusion reactor

    International Nuclear Information System (INIS)

    Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.

    1999-01-01

    A new concept of a Liquid Lithium Fusion Reactor and the first experimental results were presented at the 16th IAEA Conference on Fusion Energy. During the past two years theoretical estimations have been made, and calculated and experimental results have been obtained in confirmation of this concept and supporting its progress. The main results of this work are given in the paper. (author)

  20. Supercritical CO2 Brayton power cycles for DEMO fusion reactor based on Helium Cooled Lithium Lead blanket

    International Nuclear Information System (INIS)

    Linares, José Ignacio; Herranz, Luis Enrique; Fernández, Iván; Cantizano, Alexis; Moratilla, Beatriz Yolanda

    2015-01-01

    Fusion energy is one of the most promising solutions to the world energy supply. This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles (S-CO 2 ) for low-temperature divertor fusion reactors cooled by helium (as defined by EFDA). Integration of three thermal sources (i.e., blanket, divertor and vacuum vessel) has been studied through proposing and analyzing a number of alternative layouts, achieving an improvement on power production higher than 5% over the baseline case, which entails to a gross efficiency (before self-consumptions) higher than 42%. In spite of this achievement, the assessment of power consumption for the circulating heat transfer fluids results in a penalty of 20% in the electricity production. Once the most suitable layout has been selected an optimization process has been conducted to adjust the key parameters to balance performance and size, achieving an electrical efficiency (electricity without taking into account auxiliary consumptions due to operation of the fusion reactor) higher than 33% and a reduction in overall size of heat exchangers of 1/3. Some relevant conclusions can be drawn from the present work: the potential of S-CO 2 cycles as suitable converters of thermal energy to power in fusion reactors; the significance of a suitable integration of thermal sources to maximize power output; the high penalty of pumping power; and the convenience of identifying the key components of the layout as a way to optimize the whole cycle performance. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of HCLL fusion reactor. • Low temperature sources have been successfully integrated with high temperature ones. • Optimization of thermal sources integration improves 5% the electricity production. • Assessment of pumping power with sources and sink loops results on 20% of gross power. • Matching of key parameters has conducted to 1/3 of reduction in heat

  1. Formation of electrically insulating coatings on aluminided vanadium-base alloys in liquid lithium

    International Nuclear Information System (INIS)

    Park, J.H.; Dragel, G.

    1993-01-01

    Aluminide coatings were produced on vanadium and vanadium-base alloys by exposure of the materials to liquid lithium that contained 3-5 at.% dissolved aluminum in sealed capsules at temperatures between 775 and 880 degrees C. Reaction of the aluminide layer with dissolved nitrogen in liquid lithium provides a means of developing an in-situ electrical insulator coating on the surface of the alloys. The electrical resistivity of A1N coatings on aluminided V and V-20 wt.% Ti was determined in-situ

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  3. Methods of tritium recovery from molten lithium

    International Nuclear Information System (INIS)

    Farookhi, R.; Rogers, J.E.

    1968-01-01

    It is important to keep the tritium inventory in a blanket of a thermonuclear reactor at a low level both to eliminate possible hydriding of structural components and to reduce inventory cost. Removing the tritium from a lithium blanket by fractional distillation, flash vaporization, and fractional crystallization was investigated. No definitive data are available either on the vapor-liquid equilibrium between lithium and tritium at low T 2 concentrations, or on the rate of formation and decomposition of lithium tritide. The final distinction between the recovery systems discussed in this report will depend on such data, but presently distillation appears to be the best alternate to the diffusion scheme proposed by A.P. Fraas. The capital cost of equipment necessary to remove tritium by distillation appears to be greater than 10 million dollars for a 5000 MW system, whereas the capital cost associated with the diffusion process has been estimated to be 4 million dollars

  4. Research and development of lithium isotope separation using an ionic-liquid impregnated organic membrane

    International Nuclear Information System (INIS)

    Hoshino, Tsuyoshi

    2013-01-01

    The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 ( 6 Li) in tritium breeding materials. However, natural Li contains only about 7.6 at.% 6 Li. In Japan, new lithium isotope separation technique using ionic-liquid impregnated organic membranes have been developed. The improvement in the durability of the ionic-liquid impregnated organic membrane is one of the main issues for stable, long-term operation of electrodialysis cells while maintaining good performance. Therefore, we developed highly-durable ionic-liquid impregnated organic membrane. Both ends of the ionic-liquid impregnated organic membrane were covered by a nafion 324 overcoat to prevent the outflow of the ionic liquid. The transmission of Lithium aqueous solution after 10 hours under the highly-durable ionic-liquid impregnated organic membrane is almost 13%. So this highly-durable ionic-liquid impregnated organic membrane for long operating of electrodialysis cells has been developed through successful prevention of ion liquid dissolution. (J.P.N.)

  5. 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.

  6. Model and simulation of a vacuum sieve tray for T extraction from liquid PbLi breeding blankets

    International Nuclear Information System (INIS)

    Mertens, M.A.J.; Demange, D.; Frances, L.

    2016-01-01

    Highlights: • A simulation tool was developed to analyse, optimise and scale up VST set-ups. • This tool predicts that efficiencies higher than 90% can be reached. • Upscaling to DEMO breeding blanket flow rates results in feasibly sized designs. - Abstract: Tritium self-sufficiency within a nuclear fusion reactor is necessary to demonstrate nuclear fusion as a viable source of energy. Tritium can be produced within liquid eutectic PbLi but then has to be extracted to be refuelled to the plasma. The vacuum sieve tray (VST) method is based on the extraction of tritium from millimetre-scaled oscillating PbLi droplets falling inside a vacuum chamber. A simulation tool was developed describing the fluid dynamics occurring along the PbLi flow and was used to study the influence of the different geometrical and operational parameters on the VST performance. The simulation predicts that extraction efficiencies over 90% can be easily reached according to theory and previous experimental results. The size of the VST extraction unit for a fusion reactor is estimated based on the findings from our single-nozzle model and assuming no T reabsorption. It is found to be in the feasible range. Nevertheless, two approaches are discussed which may further reduce this size by up to 90%. The simulation tool proved to be an easy and powerful way to analyse and optimise VST set-ups at any scale.

  7. Hydrogen extraction from liquid lithium-lead alloy by bubbling with rotational jet nozzle

    International Nuclear Information System (INIS)

    Xie Bo; Yang Tongzai; Guan Rui; Weng Kuiping

    2010-01-01

    The technology of tritium extraction from lithium-lead alloy has been simulated, hydrogen extraction from lithium-lead alloy by bubbling with rotational jet nozzle being used to simulate tritium in the study based on the introduction of fluid dynamics to establish algebraic model. The results show that the higher than lithium-lead melting temperature, the higher cumulative hydrogen extraction efficiency, and gas holdup of bubble column is little affected by the impeller diameter. Gas holdup when using small aperture is slightly higher when using large aperture only at a high helium flow rate, but the smaller the aperture, the greater the bubble surface area, and a marked increase in intensity of flow circulation for liquid lithium-lead with the increase of helium flow rate, hydrogen extraction rate increases too. Moreover, influence of the jet rotational velocity on hydrogen extraction is limited. (authors)

  8. Tritium permeation barriers in contact with liquid lithium-lead eutectic (Pb-17Li)

    International Nuclear Information System (INIS)

    Forcey, K.S.; Perujo, A.

    1995-01-01

    The permeation of deuterium through coated stainless steel tubes containing liquid lithium-lead eutectic (Pb-17Li) has been studied and compared to measurements through tubes without the lithium compound. The measurements form part of an investigation into the effect of a potential tritium breeder material on permeation barriers for fusion reactors. The coatings studied were CVD TiC and Al 2 O 3 and a pack aluminised layer. Without the lithium-lead, the CVD coatings reduced the permeation rate up to 1 order of magnitude, and the aluminised layer up to 2 orders of magnitude. A CVD layer was unaffected by Pb-17Li whilst in the case of the aluminised tube, the lithium-lead completely removed the permeation barrier, presumably by attacking the surface oxide. Furthermore, the aluminised sample presented a large number of cracks and poor adheren ce to the substrate. ((orig.))

  9. Two-dimensional cross-section sensitivity and uncertainty analysis of the LBM [Lithium Blanket Module] experiments at LOTUS

    International Nuclear Information System (INIS)

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

    1988-01-01

    In a recent common Los Alamos/PSI effort, a sensitivity and nuclear data uncertainty path for the modular code system AARE (Advanced Analysis for Reactor Engineering) was developed. This path includes the cross-section code TRAMIX, the one-dimensional finite difference S/sub N/-transport code ONEDANT, the two-dimensional finite element S/sub N/-transport code TRISM, and the one- and two-dimensional sensitivity and nuclear data uncertainty code SENSIBL. Within the framework of the present work a complete set of forward and adjoint two-dimensional TRISM calculations were performed both for the bare, as well as for the Pb- and Be-preceeded, LBM using MATXS8 libraries. Then a two-dimensional sensitivity and uncertainty analysis for all cases was performed. The goal of this analysis was the determination of the uncertainties of a calculated tritium production per source neutron from lithium along the central Li 2 O rod in the LBM. Considered were the contributions from 1 H, 6 Li, 7 Li, 9 Be, /sup nat/C, 14 N, 16 O, 23 Na, 27 Al, /sup nat/Si, /sup nat/Cr, /sup nat/Fe, /sup nat/Ni, and /sup nat/Pb. 22 refs., 1 fig., 3 tabs

  10. Liquid-liquid extraction to lithium isotope separation based on room-temperature ionic liquids containing 2,2'-binaphthyldiyl-17-crown-5

    International Nuclear Information System (INIS)

    Sun Xiaoli; Zhou Wen; Gu Lin; Qiu Dan; Ren Donghong; Gu Zhiguo; Li Zaijun

    2015-01-01

    A novel liquid-liquid extraction system was investigated for the selective separation of lithium isotopes using ionic liquids (ILs = C 8 mim + PF 6 - , C 8 mim + BF 4 - , and C 8 mim + NTf 2 - ) as extraction solvent and 2,2'-binaphthyldiyl-17-crown-5 (BN-17-5) as extractant. The effects of the concentration of lithium salt, counter anion of lithium salt, initial pH of aqueous phase, extraction temperature, and time on the lithium isotopes separation were discussed. Under optimized conditions, the maximum single-stage separation factor α of 6 Li/ 7 Li obtained in the present study was 1.046 ± 0.002, indicating the lighter isotope 6 Li was enriched in IL phase while the heavier isotope 7 Li was concentrated in the solution phase. The formation of 1:1 complex Li(BN-17-5) + in the IL phase was determined on the basis of slope analysis method. The large value of the free energy change (-ΔG° = 92.89 J mol -1 ) indicated the high separation capability of the Li isotopes by BN-17-5/IL system. Lithium in Li(BN-17-5) + complex was stripped by 1 mol L -1 HCl solution. The extraction system offers high efficiency, simplicity, and green application prospect to lithium isotope separation. (author)

  11. High-power liquid-lithium target prototype for accelerator-based boron neutron capture therapy.

    Science.gov (United States)

    Halfon, S; Paul, M; Arenshtam, A; Berkovits, D; Bisyakoev, M; Eliyahu, I; Feinberg, G; Hazenshprung, N; Kijel, D; Nagler, A; Silverman, I

    2011-12-01

    A prototype of a compact Liquid-Lithium Target (LiLiT), which will possibly constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals, was built. The LiLiT setup is presently being commissioned at Soreq Nuclear Research Center (SNRC). The liquid-lithium target will produce neutrons through the (7)Li(p,n)(7)Be reaction and it will overcome the major problem of removing the thermal power generated using a high-intensity proton beam (>10 kW), necessary for sufficient neutron flux. In off-line circulation tests, the liquid-lithium loop generated a stable lithium jet at high velocity, on a concave supporting wall; the concept will first be tested using a high-power electron beam impinging on the lithium jet. High intensity proton beam irradiation (1.91-2.5 MeV, 2-4 mA) will take place at Soreq Applied Research Accelerator Facility (SARAF) superconducting linear accelerator currently in construction at SNRC. Radiological risks due to the (7)Be produced in the reaction were studied and will be handled through a proper design, including a cold trap and appropriate shielding. A moderator/reflector assembly is planned according to a Monte Carlo simulation, to create a neutron spectrum and intensity maximally effective to the treatment and to reduce prompt gamma radiation dose risks. Copyright © 2011 Elsevier Ltd. All rights reserved.

  12. Liquid metal MHD and heat transfer in a tokamak blanket slotted coolant channel

    International Nuclear Information System (INIS)

    Reed, C.B.; Hua, T.Q.; Black, D.B.; Kirillov, I.R.; Sidorenkov, S.I.; Shapiro, A.M.; Evtushenko, I.A.

    1993-01-01

    A liquid metal MHD (Magnetohydrodynamic)/heat transfer test was conducted at the ALEX (Argonne Liquid Metal Experiment) facility of ANL (Argonne National Laboratory), jointly between ANL and NIIEFA (Efremov Institute). The test section was a rectangular slotted channel geometry (meaning the channel has a high aspect ratio, in this case 10:1, and the long side is parallel to the applied magnetic field). Isothermal and heat transfer data were collected. A heat flux of ∼9 W/cm 2 was applied to the top horizontal surface (the long side) of the test section. Hartmann Numbers to 1050 (2 Tesla), interaction parameters to 9 x 10 3 , Peclet numbers of 10--200, based on the half-width of the small dimension (7mm), and velocities of 1--75 cm/sec. were achieved. The working fluid was NaK (sodium potassium eutectic). All four interior walls were bare, 300-series stainless steel, conducting walls

  13. 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)

  14. Development of aluminide coatings on vanadium-base alloys in liquid lithium

    International Nuclear Information System (INIS)

    Park, J.H.; Dragel, D.

    1993-01-01

    Aluminide coatings were produced on vanadium and vanadium-base alloys by exposure of the materials to liquid lithium that contained 3/5 at.% dissolved aluminum in sealed V and V-20 wt.% Ti capsules at temperatures between 775 and 880 degrees C. After each test, the capsules were opened and the samples were examined by optical microscopy and scanning electron microscopy (SEM), and analyzed by electron-energy-dispersive spectroscopy (EDS) and X-ray diffraction. Hardness of the coating layers and bulk alloys was determined by microidentation techniques. The nature of the coatings, i.e., surface coverage, thickness, and composition, varied with exposure time and temperature, solute concentration in lithium, and alloy composition. Solute elements that yielded adherent coatings on various substrates can provide a means of developing in-situ electrical insulator coatings by reaction of the reactive layers with dissolved nitrogen in liquid lithium

  15. High-power liquid-lithium jet target for neutron production

    Science.gov (United States)

    Halfon, S.; Arenshtam, A.; Kijel, D.; Paul, M.; Berkovits, D.; Eliyahu, I.; Feinberg, G.; Friedman, M.; Hazenshprung, N.; Mardor, I.; Nagler, A.; Shimel, G.; Tessler, M.; Silverman, I.

    2013-12-01

    A compact liquid-lithium target (LiLiT) was built and tested with a high-power electron gun at the Soreq Nuclear Research Center. The lithium target, to be bombarded by the high-intensity proton beam of the Soreq Applied Research Accelerator Facility (SARAF), will constitute an intense source of neutrons produced by the 7Li(p,n)7Be reaction for nuclear astrophysics research and as a pilot setup for accelerator-based Boron Neutron Capture Therapy. The liquid-lithium jet target acts both as neutron-producing target and beam dump by removing the beam thermal power (>5 kW, >1 MW/cm3) with fast transport. The target was designed based on a thermal model, accompanied by a detailed calculation of the 7Li(p,n) neutron yield, energy distribution, and angular distribution. Liquid lithium is circulated through the target loop at ˜200 °C and generates a stable 1.5 mm-thick film flowing at a velocity up to 7 m/s onto a concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power areal densities of >4 kW/cm2 and volume power density of ˜2 MW/cm3 at a lithium flow of ˜4 m/s while maintaining stable temperature and vacuum conditions. The LiLiT setup is presently in online commissioning stage for high-intensity proton beam irradiation (1.91-2.5 MeV, 1-2 mA) at SARAF.

  16. Testing of the prototype FMIT target with liquid lithium

    International Nuclear Information System (INIS)

    Miller, W.C.; Annese, C.E.; Berg, J.D.; Miles, R.R.

    1984-01-01

    Testing of a molten lithium target was performed to evaluate hydraulic stability, determine surface evaporation rates, and map the detailed contour of the high speed, free surface wall jet. The results confirmed predictions by demonstrating acceptable performance of a prototype target

  17. Interactions of solid and liquid lithium with steady state hydrogen and helium plasmas

    International Nuclear Information System (INIS)

    Hirooka, Y.; Nishikawa, M.; Ohgaki, H.; Ohtsuka, Y.

    2005-01-01

    A variety of innovative Plasma-Facing Component (PFC) concepts, employing moving solid or liquid surfaces, have recently been proposed in order to resolve technical issues, associated with the applications of currently used PFCs in future steady state fusion devices. As the first step to evaluate the concept using flowing-liquids for PFCs, steady state hydrogen and helium plasma interactions with solid and standing liquid lithium have been investigated in the present work, using the H α and He-I spectroscopy at the ion bombarding energies up to 150eV and at the lithium temperatures between room temperature and 480 deg C. Data indicate that hydrogen recycling over liquid lithium is clearly reduced, relative to that over solid lithium, whereas helium recycling does not show the same trend. From the kinetic analysis of these recycling time constant data, the activation energies for the overall recycling processes have been evaluated to be 0.02±0.01eV, both for hydrogen and helium plasmas. Also, it has been found that the activation energy is nearly independent of ion bombarding energy. (author)

  18. Dissolved nitrogen in liquid lithium - a problem in fusion reactor chemistry

    International Nuclear Information System (INIS)

    Hubberstey, P.

    1984-01-01

    When dissolved in liquid lithium, nitrogen adopts the role filled by oxygen in liquid sodium systems, reacting readily with stainless steel containment materials to form Li 9 CrN 5 as a surface product; extended reaction leads to pronounced corrosion and embrittlement problems. It also interacts with both carbon and silicon impurities forming Li 2 NCN and Li 5 SiN 3 , respectively; it is inert, however, to oxygen impurity. Although dissolved nitrogen reacts with neither the tritium generated in the breeding process nor the lead added to act as a neutron multiplier, its presence may seriously influence tritium recovery processes since it reacts with and hence may poison the majority of the transition metals (Y,Ti,Zr) presently being considered as tritium getter materials. Its reactivity with these metals forms the basis of the hot trapping technique used to remove dissolved nitrogen from liquid lithium systems; cold trapping is ineffective because of its large solubility even at temperatures just above the melting point of pure lithium (453.6K). Whenever possible, the chemistry of nitrogen dissolved in liquid lithium is rationalised using the thermodynamic concepts and its significance to fusion reactor technology stressed. (author)

  19. Properties and Structure of the LiCl-films on Lithium Anodes in Liquid Cathodes

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg; Hennesø, Erik

    2016-01-01

    Lithium anodes passivated by LiCl layers in different types of liquid cathodes (catholytes) based on LiAlCl4 in SOCl2 or SO2 have been studied by means of impedance spectroscopy. The impedance spectra have been fitted with two equivalent circuits using a nonlinear least squares fit program...

  20. Experiments with Liquid Metal Walls: Status of the Lithium Tokamak Experiment

    OpenAIRE

    Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Kozub, Thomas; Berzak, Laura; Hammett, Gregory; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Jacobson, Craig M.; Lucia, Matthew; Jones, Andrew; Lundberg, Daniel; Timberlake, John; Majeski, Richard

    2010-01-01

    Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas wa...

  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. Performance Tests of a Permeation Sensor for Test Blanket Modules Using Liquid Metal

    International Nuclear Information System (INIS)

    Choi, B. G.; Lee, D. W.; Lee, E. H.; Yoon, J. S.; Kim, S. K.; Shin, K. I.; Jin, H. G.

    2013-01-01

    The tritium extraction from a breeder is one of the key technologies and its methods have been investigated. For developing the tritium extraction methods and evaluating the amount of tritium in the system, a reliable and correct sensor is required to measure the hydrogen concentration in liquid metal breeder. There are several researches for developing the sensors in the ITER participants and especially, EU has developed the permeation sensors trying to selecting materials with low Serviette's constant (solubility) and high hydrogen diffusivity coefficient. However, EU's response time is still too long time about tens of minutes to measure the tritium concentration in the online system. We have been performing the preliminary tests with designed and fabricated sensors to solve the late response of sensor. However, we could not continue the tests because of the membrane's oxidation (pure Fe) and the difficulty of welding nonferrous metals. In present study, a permeation sensor made of vacuum flanges with a porous plate inside is proposed not only to eliminate the difficulty of the fabrication but to optimize the performance of sensor. The permeation sensor to measure the hydrogen isotopes in liquid metal breeder has been proposed and evaluated to overcome the limitation of a long response time for various shapes and materials. We found that the previous sensors have limitation; the oxidation problems (pure Fe) and the difficulty in welding (nonferrous metals). Therefore we proposed a permeation sensor with the vacuum flanges filled with porous disks to eliminate the problems. By using the CF flanges, the problem caused by welding is removed. But the permeable response time of sensors took a long time to reach the pressure equivalent

  3. Study of the corrosion behaviors of 304 austenite stainless steel specimens exposed to static liquid lithium at 600 K

    Energy Technology Data Exchange (ETDEWEB)

    Meng, Xiancai [Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zuo, Guizhong; Ren, Jun; Xu, Wei; Sun, Zhen; Huang, Ming [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Wangyu [Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China); Hu, Jiansheng, E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Deng, Huiqiu, E-mail: hqdeng@hnu.edu.cn [Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China)

    2016-11-15

    Investigation of corrosion behavior of stainless steel served as one kind of structure materials exposed to liquid lithium (Li) is one of the keys to apply liquid Li as potential plasma facing materials (PFM) or blanket coolant in the fusion device. Corrosion experiments of 304 austenite stainless steel (304 SS) were carried out in static liquid Li at 600 K and up to1584 h at high vacuum with pressure less than 4 × 10{sup −4} Pa. After exposure to liquid Li, it was found that the weight of 304 SS slightly decreased with weight loss rate of 5.7 × 10{sup −4} g/m{sup 2}/h and surface hardness increased by about 50 HV. Lots of spinel-like grains and holes were observed on the surface of specimens measured by SEM. By further EDS, XRD and metallographic analyzing, it was confirmed that the main compositions of spinel-like grains were M{sub 23}C{sub 6} carbides, and 304 SS produced a non-uniform corrosion behavior by preferential grain boundary attack, possibly due to the easy formation of M{sub 23}C{sub 6} carbides and/or formation of Li compound at grain boundaries.

  4. CDX-U Operation with a Large Area Liquid Lithium Limiter

    International Nuclear Information System (INIS)

    R. Majeski; M. Boaz; D. Hoffman; B. Jones; R. Kaita; H. Kugel; T. Munsat; J. Spaleta; V. Soukhanovskii; J. Timberlake; L. Zakharov; G. Antar; R. Doerner; S. Luckhardt; R.W. Conn; M. Finkenthal; D. Stutman; R. Maingi; M. Ulrickson

    2002-01-01

    The Current Drive experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has begun experiments with a fully toroidal liquid lithium limiter. CDX-U is a compact [R = 34 cm, a = 22 cm, B(subscript)toroidal = 2 kG, I(subscript)P = 100 kA, T(subscript)e(0) ∼ 100 eV, n(subscript)e(0) ∼ 5 x 10 19 m -3 ] short-pulse (<25 msec) spherical torus with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, can be filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium prior to the experiment. Surface coatings are evident on part of the lithium. Despite the surface coatings, tokamak discharges operated in contact with the lithium-filled tray show evidence of reduced impurities and recycling. The reduction in recycling is largest when the lithium is liquefied by heating to 250 degrees Celsius

  5. Small scale lithium-lead/water-interaction studies

    International Nuclear Information System (INIS)

    Kranert, O.; Kottowski, H.

    1991-01-01

    One current concept in fusion blanket design is to utilize water as the coolant and liquid lithium-lead as the breeding/neutron multiplier material. Considering the complex design of the blanket module, it is likely that a water leakage into the liquid alloy may occur due to a tube rupture provoking an intolerable pressure increase in the blanket module. The pressure increase is caused by the combined chemical and thermohydraulic reaction of lithium-lead with water. Experiments which simulate such a transient event are necessary to obtain information which is important for the blanket module design. The interaction has been investigated by conducting small-scale experiments at various injection pressures, alloy- and coolant temperatures. Besides using eutectic Li 17 Pb 83 , Li 7 Pb 2 , lithium and lead have been used. Among other results, the experiments indicate increasing chemical reaction with increasing lithium concentration. At the same time, the chemical reaction inhibits violent thermohydaulic reactions due to the attenuating effect of the hydrogen produced. The preliminary epxerimental results from Li 17 Pb 83 and Li 7 Pb 2 reveal that the pressure- and temperature transients caused by the chemical and thermohydraulic reactions lie within technically manageable limits. (orig.)

  6. Cycling performance of lithium polymer cells assembled by in situ polymerization of a non-flammable ionic liquid monomer

    International Nuclear Information System (INIS)

    Lee, Yoon-Sung; Kim, Dong-Won

    2013-01-01

    Highlights: • Gel polymer electrolytes were synthesized by in situ polymerization of ionic liquid in the lithium polymer cells. • Flammability of the electrolyte was significantly reduced by polymerizing electrolyte containing a non-flammable ionic liquid monomer. • The cells assembled with polymeric ionic liquid-based electrolytes exhibited reversible cycling behavior with good capacity retention. -- Abstract: Lithium polymer cells composed of a lithium negative electrode and a LiCoO 2 positive electrode were assembled with a gel polymer electrolyte obtained by in situ polymerization of an electrolyte solution containing an ionic liquid monomer with vinyl groups. The polymerization of the electrolyte solution containing the non-flammable ionic liquid monomer resulted in a significant reduction of the flammability of the gel polymer electrolytes. The lithium polymer cell assembled with the stable gel polymer electrolyte delivered a discharge capacity of 134.3 mAh g −1 at ambient temperature and exhibited good capacity retention

  7. Characterization of fueling NSTX H-mode plasmas diverted to a liquid lithium divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kugel, H.W.; Abrams, T. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kallman, J. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Mansfield, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); and others

    2013-07-15

    Deuterium fueling experiments were conducted with the NSTX Liquid Lithium Divertor (LLD). Lithium evaporation recoated the LLD surface to approximate flowing liquid Li to sustain D retention. In the first experiment with the diverted outer strike point on the LLD, the difference between the applied D gas input and the plasma D content reached very high values without disrupting the plasma, as would normally occur in the absence of Li pumping, and there was also little change in plasma D content. In the second experiment, constant fueling was applied, as the LLD temperature was varied to change the surface from solid to liquid. The D retention was relatively constant, and about the same as that for solid Li coatings on graphite, or twice that achieved without Li PFC coatings. Contamination of the LLD surface was also possible due to compound formation and erosion and redeposition from carbon PFCs.

  8. New Polymer and Liquid Electrolytes for Lithium Batteries

    International Nuclear Information System (INIS)

    McBreen, J.; Lee, H. S.; Yang, X. Q.; Sun, X.

    1999-01-01

    All non-aqueous lithium battery electrolytes are Lewis bases that interact with cations. Unlike water, they don't interact with anions. The result is a high degree of ion pairing and the formation of triplets and higher aggregates. This decreases the conductivity and the lithium ion transference and results in polarization losses in batteries. Approaches that have been used to increase ion dissociation in PEO based electrolytes are the use of salts with low lattice energy, the addition of polar plasticizers to the polymer, and the addition of cation completing agents such as crown ethers or cryptands. Complexing of the anions is a more promising approach since it should increase both ion dissociation and the lithium transference. At Brookhaven National Laboratory (BNL) we have synthesized two new families of neutral anion completing agents, each based on Lewis acid centers. One is based on electron deficient nitrogen sites on substituted aza-ethers, wherein the hydrogen on the nitrogen is replaced by electron withdrawing groups such as CF 3 SO 3- . The other is based on electron deficient boron sites on borane or borate compounds with various fluorinated aryl or alkyl groups. Some of the borane based anion receptors can promote the dissolution of LiF in several solvents. Several of these compounds, when added in equivalent amounts, produce 1.2M LiF solutions in DME, an increase in volubility of LiF by six orders of magnitude. Some of these LiF electrolytes have conductivities as high as 6 x 10 -3 Scm -1 . The LiF electrolytes with borane anion acceptors in PC:EC:DEC solvents have excellent electrochemical stability. This has been demonstrated in small Li/LiMn 2 O 4 cells

  9. Shielding consideration for a deuteron activated liquid lithium system

    International Nuclear Information System (INIS)

    Huang, S.T.; Shapiro, A.M.; Lee, J.B.; Miller, W.C.

    1979-09-01

    A parametric study was conducted to evaluate the potential shielding implication due to the 7 Be plateout on the lithium piping in the FMIT facility. Various parameters such as plateout percentage, hot flush efficiency and 7 Be trapping efficiency were varied to assess the overall shielding requirement relationship. The 7 Be plateout was found to place severe limitations on the hands-on maintenance access. Hot flush and 7 Be traps are effective ways of minimizing the 7 Be plateout. To be effective in reducing local shielding requirements, the combined 7 Be trapping and hot flush efficiency shall be greater than 95%

  10. Ferrous alloy metallurgy - liquid lithium corrosion and welding. Progress report, January 1-December 31, 1980

    International Nuclear Information System (INIS)

    Olson, D.L.; Matlock, D.K.

    1980-01-01

    Fatigue crack growth has been used to evaluate the interaction between liquid lithium and an imposed stress. Fatigue crack growth data on type 304L stainless steel at 700C and 2 1/4Cr-1Mo steel between 500 and 700C show that for all imposed test conditions (i.e. frequency, temperature, and nitrogen content in the lithium) the interaction of lithium with the strain at the crack tip results in enhanced crack growth rates. The enhanced growth rates result from the effects of either enhanced grain boundary penetration or a change in crack propagation mechanism due to liquid metal embrittlement. Auger spectroscopy of grain boundary penetrated specimen shows that a lithium-oxygen compound forms at the grain boundary. Moessbauer evaluations of the ferrite layer of corroded type 304 stainless steel are being used to develop a model for weight loss in liquid lithium. The welding research in progress is directed to characterize the influence of variations of the austenitic weld metal composition on the microstructural and mechanical properties of dissimilar metal weldments. Weldments of 2 1/4Cr-1Mo steel to 316 stainless steel have been investigated for fusion microstructure, thermal expansion impact strength and characterization of specific long time in-service failures. Modification of weld metal microstructures by microalloy additions is being investigated as a concept to improve weld metal properties. The behavior of a strip electrode in a gas metal arc is being investigated to determine the feasibility of gas metal arc weld strip overlay cladding

  11. Retention/Diffusivity Studies in Free-Surface Flowing Liquid Lithium

    International Nuclear Information System (INIS)

    R.A. Stubbers; G.H. Miley; M. Nieto; W. Olczak; D.N. Ruzic; A. Hassanein

    2004-01-01

    FLIRE was designed to measure the hydrogen and helium retention and diffusivity in a flowing stream of liquid lithium, and it has accomplished these goals. Retention coefficients for helium in the flowing liquid stream were 0.1-2% for flow speeds of 44 cm/s and implantation energies between 500 and 2000 eV. The energy dependence of retention is linear for the energy range considered, as expected, and the dependence of retention on flow velocity fits the expected square-root of flow speed dependence. Estimates of the helium diffusion coefficient in the flowing lithium stream were ∼ 4 x 10 -7 cm 2 /s, and are independent of implantation energy. This value is much lower than expected, which could be due to several factors, such as mixing, bubble formation or surface film formation. In the case of hydrogen, long term retention and release mechanisms are of greatest importance, since this relates to tritium inventory in flowing lithium PFCs for fusion applications. The amount of hydride formation was measured for flowing lithium exposed to neutral deuterium gas. Thermal desorption spectroscopy (TDS) measurements indicate that the hydride concentration was between 0.1 and 0.2% over a wide range of pressures (6.5 x 10 -5 to 1 Torr). This result implies that the deuterium absorption rate is limited by the surface dissociation rate, since deuterium (hydrogen/tritium) is absorbed in its atomic form, not its molecular form

  12. Retention/Diffusivity Studies in Free-Surface Flowing Liquid Lithium

    Energy Technology Data Exchange (ETDEWEB)

    R.A. Stubbers; G.H. Miley; M. Nieto; W. Olczak; D.N. Ruzic; A. Hassanein

    2004-12-14

    FLIRE was designed to measure the hydrogen and helium retention and diffusivity in a flowing stream of liquid lithium, and it has accomplished these goals. Retention coefficients for helium in the flowing liquid stream were 0.1-2% for flow speeds of 44 cm/s and implantation energies between 500 and 2000 eV. The energy dependence of retention is linear for the energy range considered, as expected, and the dependence of retention on flow velocity fits the expected square-root of flow speed dependence. Estimates of the helium diffusion coefficient in the flowing lithium stream were {approx} 4 x 10{sup -7} cm{sup 2}/s, and are independent of implantation energy. This value is much lower than expected, which could be due to several factors, such as mixing, bubble formation or surface film formation. In the case of hydrogen, long term retention and release mechanisms are of greatest importance, since this relates to tritium inventory in flowing lithium PFCs for fusion applications. The amount of hydride formation was measured for flowing lithium exposed to neutral deuterium gas. Thermal desorption spectroscopy (TDS) measurements indicate that the hydride concentration was between 0.1 and 0.2% over a wide range of pressures (6.5 x 10{sup -5} to 1 Torr). This result implies that the deuterium absorption rate is limited by the surface dissociation rate, since deuterium (hydrogen/tritium) is absorbed in its atomic form, not its molecular form.

  13. 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

  14. Control of the nitrogen concentration in liquid lithium by the hot trap method

    International Nuclear Information System (INIS)

    Sakurai, Toshiharu; Yoneoka, Toshiaki; Tanaka, Satoru; Suzuki, Akihiro; Muroga, Takeo

    2002-01-01

    The nitrogen concentration in liquid lithium was controlled by the hot-trap method. Titanium, vanadium and a V-Ti alloy were used as nitrogen gettering materials. Gettering experiments were conducted at 673, 773 and 823 K for 0.4-2.8 Ms. After immersion, the nitrogen concentration increased in titanium and V-Ti were tested at 823 K. Especially the nitrogen gettering effect by the V-10at.%Ti alloy was found to be large. Nitrogen was considered to exist mainly as solid solution in the V-10at.%Ti alloy. The decrease of the nitrogen concentration in liquid lithium by the V-Ti gettering was also confirmed

  15. Measurement of free-surface of liquid metal lithium jet for IFMIF target

    International Nuclear Information System (INIS)

    Hiroo Kondo; Nobuo Yamaoka; Takuji Kanemura; Seiji Miyamoto; Hiroshi Horiike; Mizuho Ida; Hiroo Nakamura; Izuru Matsushita; Takeo Muroga

    2006-01-01

    This reports an experimental study on flow characteristics of a lithium target flow of International Fusion Materials Irradiation Facility (IFMIF). Surface shapes of the target were tried to measure by pattern projection method that is a three dimensional image measurement method. Irregularity of the surface shape caused by surface wakes was successfully measured by the method. IFMIF liquid lithium target is formed a flat plane jet of 25 mm in depth and 260 mm in width, and flows in a flow velocity range of 10 to 20 m/s. Aim of this study is to develop measurement techniques for monitoring of the target when IFMIF is in operation. The lithium target flow is high speed jet and the temperature high is more than 500 K. Also, light is not transmitted into liquid metal lithium. Therefore, almost of all flow measurement techniques developed for water are not used for lithium flow. In this study, pattern projection method was employed to measure the surface irregularity of the target. In the method, stripe patterns are projected onto the flow surface. The projected patterns are deformed according the surface shape. Three-dimensional surface shape is measured by analyzing the deformed patterns recorded using a CCD camera. The method uses the property that lithium dose not transmit visible lights. The experiments were carried out using a lithium loop at Osaka University. In this facility, lithium plane jet of 10 mm in depth and 70 mm width is obtained in the velocity range of less than 15 m/s using a two contractions nozzle. The pattern projection method was used to measure the amplitude of surface irregularity caused by surface wakes. The surface wakes were generated from small damaged at the nozzle edge caused by erosion, and those were successfully measured by the method. The measurement results showed the amplitude of the surface wakes were approximately equal to a size of damage of a nozzle. The amplitude was decreasing with distance to down stream and with decreasing

  16. Development of liquid lithium divertor for fusion reactor

    International Nuclear Information System (INIS)

    Evtihkin, V. A.; Lyublinskij, I. E.; Vertkov, A.V.; Chumanov, A.V.; Shpolyanskij, V.N.

    2000-01-01

    Development of divertor is one of the most acute problems of the tokamak fusion reactor. The use of such materials as tungsten, beryllium, graphite and CFC's enabled to solve the problem to a certain extent fulfilling the need of the ITER project. The problem still rests unsolved for the DEMO-type reactors. Lithium if used as a material for high heat flux components may provide a successful solution of the problem. A concept of Li divertor based on the use of capillary-pore structures (CPS) is proposed and is being validated by a complex of experimental research and engineering developments. An optional concept of Li divertor for power removal at 400 MW in steady-state (DEMO-S project) is presented. The complex of experimental research is under way to prove the serviceability of the Li CPS in different conditions that would be realized in divertor

  17. Collective dynamics in liquid lithium, sodium, and aluminum

    International Nuclear Information System (INIS)

    Singh, Shaminder; Tankeshwar, K.

    2003-01-01

    Inelastic x-ray scattering data of liquid Li, Na, and Al for dynamical structure factors have been analyzed by proposing a semiempirical model. The model is based on the extension of the hydrodynamic model to the viscoelastic region so that it satisfies the first four nonvanishing sum rules. It has been found that the semiempirical model fits well with the x-ray scattering data for liquid metals investigated here. The physical meaning of the parameters is also discussed

  18. Study on the electrochemical of the metal deposition from ionic liquids for lithium, titanium and dysprosium

    International Nuclear Information System (INIS)

    Berger, Claudia A.

    2017-01-01

    The thesis was aimed to the characterization of electrochemically deposited film of lithium, titanium and dysprosium on Au(111) from different ionic liquids, finally dysprosium on neodymium-iron-boron magnate for industrial applications. The investigation of the deposits were performed using cyclic voltametry, in-situ scanning tunneling microscopy, electrochemical quartz microbalance, XPS and Auger electron spectroscopy. The sample preparation is described in detail. The deposition rate showed a significant temperature dependence.

  19. Lithium bis(fluorosulfonyl)imide-PYR14TFSI ionic liquid electrolyte compatible with graphite

    Czech Academy of Sciences Publication Activity Database

    Nádherná, Martina; Reiter, Jakub; Moškon, J.; Dominko, R.

    2011-01-01

    Roč. 196, č. 18 (2011), s. 7700-7706 ISSN 0378-7753 R&D Projects: GA AV ČR(CZ) KJB200320801; GA AV ČR KJB200320901; GA MŠk(CZ) LC523 Institutional research plan: CEZ:AV0Z40320502 Keywords : Graphite * Ionic liquid * Bis(fluorosulfonyl)imide * Lithium -ion battery * Solid electrolyte interface Subject RIV: CA - Inorganic Chemistry OBOR OECD: Inorganic and nuclear chemistry Impact factor: 4.951, year: 2011

  20. A liquid-nitrogen monitor for lithium-drifted germanium detectors

    International Nuclear Information System (INIS)

    Andeweg, A.H.

    1977-11-01

    An instrument has been developed that makes use of a load cell to monitor the liquid nitrogen in the Dewar flask of a lithium-drifted germaniun detector. The contents are recorded on a chart recorder, and an alarm is sounded when the previously set content has been reached. A signal switches off the high-voltage power supply 30 minutes after the alarm is triggered. The calibration of the load-cell monitor is described in an appendix [af

  1. Extraction of lanthanide elements and bismuth in molten lithium chloride-liquid bismuth-lithium alloy system

    International Nuclear Information System (INIS)

    Harada, Makoto; Adachi, Motonari; Kai, Yuichi; Koike, Kenichi

    1987-01-01

    The equilibrium distributions of neodymium and samarium between molten LiCl and liquid Bi-Li alloy were measured in a wide range of Li-mole fraction in the alloy phase, X Li . These lanthanide elements were extracted through redox reactions. In high X Li range, X Li > 0.03, the distributions of neodymium and bismuth in the salt phase increased markedly. The anomalous increase is attributed to the formation of the compound comprized of Nd, Li, Bi and oxygen in the salt phase. The reaction processes in samarium and neodymium were very fast and the extraction rates are controlled by the diffusion processes of the solutes and metallic lithium. (author)

  2. 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

  3. 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

  4. Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

    Science.gov (United States)

    Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

    2013-10-01

    The present approach to magnetic fusion which relies on high recycling plasma-wall interaction has exhausted itself at the level of TFTR, JET, JT-60 devices with no realistic path to the burning plasma. Instead, magnetic fusion needs a return to its original idea of insulation of the plasma from the wall, which was the dominant approach in the 1970s and upon implementations has a clear path to the DEMO device with PDT ~= 100 MW and Qelectric > 1 . The SFLiLi systems of this talk is the technology tool for implementation of the guiding idea of magnetic fusion. It utilizes the unique properties of flowing LiLi to pump plasma particles and, thus, insulate plasma from the walls. The necessary flow rate, ~= 1 g3/s, is very small, thus, making the use of lithium practical and consistent with safety requirements. The talk describes how chemical activity of LiLi, which is the major technology challenge of using LiLi in tokamaks, is addressed by SFLiLi systems at the level of already performed (HT-7) experiment, and in ongoing implementations for a prototype of SFLiLi for tokamak divertors and the mid-plane limiter for EAST tokamak (to be tested in the next experimental campaign). This work is supported by US DoE contract No. DE-AC02-09-CH11466.

  5. Reaction rates and electrical resistivities of the hydrogen isotopes with, and their solubilities in, liquid lithium

    International Nuclear Information System (INIS)

    Pulham, R.J.; Adams, P.F.; Hubberstey, P.; Parry, G.; Thunder, A.E.

    1976-01-01

    The rate of reaction, k, of hydrogen and of deuterium with liquid lithium have been determined up to pressures of 20kNm -2 and at temperatures between 230 and 270 0 C. The reaction is first order with an apparent activation energy of 52.8 and 55.2 kJmol -1 for hydrogen and deuterium, respectively. The deuterium isotope effect, k/sub H/k/sub D/, decreases from 2.95 at 230 to 2.83 at 270 0 C. Tritium is predicted to react even more slowly than deuterium. The freezing point of lithium is depressed by 0.082 and 0.075 0 C, respectively, by dissolved hydride and deuteride giving eutectics at 0.016 mol percent H and 0.012 mol percent D in the metal-salt phase diagrams. The depression and eutectic concentration are expected to be less for tritium. The increase in the resistivity of liquid lithium caused by dissolved hydrogen isotopes is linear and relatively large, 5 x 10 -8 Ωm (mol percent H or D) -1 . The solubility of lithium hydride and deuteride was determined from the marked change in resistivity on saturation. The liquidus of the metal-salt phase diagram rises steeply from the eutectic point to meet the two-immiscible liquid region. Tritium is expected to be less soluble than deuterium. The partial molar enthalpies of solution are 44.2 and 55.0 kJmol -1 for hydrogen and deuterium, respectively. These values are used to calculate the solvation enthalpies of the isotope anions in the metal

  6. Cation effect on small phosphonium based ionic liquid electrolytes with high concentrations of lithium salt

    Science.gov (United States)

    Chen, Fangfang; Kerr, Robert; Forsyth, Maria

    2018-05-01

    Ionic liquid electrolytes with high alkali salt concentrations have displayed some excellent electrochemical properties, thus opening up the field for further improvements to liquid electrolytes for lithium or sodium batteries. Fundamental computational investigations into these high concentration systems are required in order to gain a better understanding of these systems, yet they remain lacking. Small phosphonium-based ionic liquids with high concentrations of alkali metal ions have recently shown many promising results in experimental studies, thereby prompting us to conduct further theoretical exploration of these materials. Here, we conducted a molecular dynamics simulation on four small phosphonium-based ionic liquids with 50 mol. % LiFSI salt, focusing on the effect of cation structure on local structuring and ion diffusional and rotational dynamics—which are closely related to the electrochemical properties of these materials.

  7. 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

  8. Lithium-lead/water interaction. Large break experiments

    International Nuclear Information System (INIS)

    Savatteri, C.; Gemelli, A.

    1991-01-01

    One current concept in fusion blanket module design is to utilize water as coolant and liquid lithium-lead as breeding/neutron-multiplier material. Considering the possibility of certain off-normal events, it is possible that water leakage into the liquid metal may occur due to a tube rupture. The lithium-lead/water contact can lead to a thermal and chemical reaction which should provoke an intolerable pressure increase in the blanket module. For realistic simulation of such in-blanket events, the Blanket Safety Test (BLAST) facility has been built. It simulates the transient event by injecting subcooled water under high pressure into a stagnant pool of about 500 kg liquid Pb-17Li. Eight fully instrumented large break tests were carried out under different conditions. The aim of the experiments is to study the chemical and thermal process and particularly: The pressurization history of the reaction vessel, the formation and deposition of the reaction products, the identification and propagation of the reaction zones and the temperature transient in the liquid metal. In this paper the results of all tests performed are presented and discussed. (orig.)

  9. Carbon tiles as spectral-shifter for long-life liquid blanket in LHD-type reactor FFHR

    International Nuclear Information System (INIS)

    Sagara, A.; Imagawa, S.; Tanaka, T.; Muroga, T.; Kubota, Y.; Dolan, T.; Hashizume, H.; Kunugi, T.; Fukada, S.; Shimizu, A.; Terai, T.; Mitarai, O.

    2006-01-01

    In terms of engineering feasibility for long-life Flibe blanket in LHD-type reactor FFHR, the Spectral-shifter and Tritium breeder Blanket (STB) concept is evaluated by taking neutron irradiation effects into account under system integration such as Flibe cooling and components replacement. FEM calculations for the neutron wall loading of 1.5 MW/m 2 show that the temperature of the STB armor tile can be kept below 2000 K by optimizing the first metal wall thickness. The heat load experiment on the STB armor mockup confirms feasibility of the temperature control and mechanical joining. Degradation of STB armor tiles due to neutron irradiation requires replacement of them every few years by means of remote handling 'screw coasters' using helical winding, where the replaced tiles are low level wastes. Although the STB concept is feasible within nuclear and thermal properties, more detailed structural optimization is needed including the mechanical and chemical properties

  10. Susceptibility of 2 1/4 Cr-1Mo steel to liquid metal induced embrittlement by lithium-lead solutions

    International Nuclear Information System (INIS)

    Eberhard, B.A.; Edwards, G.R.

    1984-08-01

    An investigation has been conducted on the liquid metal induced embrittlement susceptibility of 2 1/4Cr-1Mo steel exposed to lithium and 1a/o lead-lithium at temperatures between 190 0 C and 525 0 C. This research was part of an ongoing effort to evaluate the compatibility of liquid lithium solutions with potential fusion reactor containment materials. Of particular interest was the microstructure present in a weld heat-affected zone, a microstructure known to be highly susceptible to corrosive attack by liquid lead-lithium solutions. Embrittlement susceptibility was determined by conducting tension tests on 2 1/4Cr-1Mo steel exposed to an inert environment as well as to a lead-lithium liquid and observing the change in tensile behavior. The 2 1/4Cr-1Mo steel was also given a base plate heat treatment to observe its embrittlement susceptibility to 1a/o lead-lithium. The base plate microstructure was severely embrittled at temperatures less than 500 0 C. Tempering the base plate was effective in restoring adequate ductility to the steel

  11. Thermodynamic analysis of chromium solubility data in liquid lithium containing nitrogen: Comparison between experimental data and computer simulation

    International Nuclear Information System (INIS)

    Krasin, Valery P.; Soyustova, Svetlana I.

    2015-01-01

    The mathematical formalism for description of solute interactions in dilute solution of chromium and nitrogen in liquid lithium have been applied for calculating of the temperature dependence of the solubility of chromium in liquid lithium with the various nitrogen contents. It is shown that the derived equations are useful to provide understanding of a relationship between thermodynamic properties and local ordering in the Li–Cr–N melt. Comparison between theory and data reported in the literature for solubility of chromium in nitrogen-contaminated liquid lithium, was allowed to explain the reasons of the deviation of the experimental semi-logarithmic plot of chromium content in liquid lithium as a function of the reciprocal temperature from a straight line. - Highlights: • The activity coefficient of chromium in ternary melt can be obtained by means of integrating the Gibbs–Duhem equation. • In lithium with the high nitrogen content, the dependence of a logarithm of chromium solubility as a function of the reciprocal temperature has essentially nonlinear character. • At temperatures below a certain threshold, the process of dissolution of chromium in lithium will be controlled by the equilibrium concentration of nitrogen required for the formation of ternary nitride Li_9CrN_5at a given temperature.

  12. Intermetallic and electrical insulator coatings on high-temperature alloys in liquid-lithium environments

    International Nuclear Information System (INIS)

    Park, J.H.

    1994-06-01

    In the design of liquid-metal cooling systems for fusion-reactor blanket, applications, the corrosion resistance of structural materials and the magnetohydrodynamic (MHD) force and its subsequent influence on thermal hydraulics and corrosion are major concerns. When the system is cooled by liquid metals, insulator coatings are required on piping surfaces in contact with the coolant. The objective of this study is to develop stable corrosion-resistant electrical insulator coatings at the liquid-metal/structural-material interface, with emphasis on electrically insulating coatings that prevent adverse MHD-generated currents from passing through the structural wall, and Be-V intermetallic coatings for first-wall components that face the plasma. Vanadium and V-base alloys are leading candidate materials for structural applications in a fusion reactor. Various intermetallic films were produced on V-alloys and on Types 304 and 316 stainless steel. The intermetallic layers were developed by exposure of the materials to liquid Li containing 2 at temperatures of 500--1030 degree C. CaO electrical insulator coatings were produced by reaction of the oxygen-rich layer with <5 at. % Ca dissolved in liquid Li at 400--700 degree C. The reaction converted the oxygen-rich layer to an electrically insulating film. This coating method is applicable to reactor components because the liquid metal can be used over and over; only the solute within the liquid metal is consumed. This paper will discuss initial results on the nature of the coatings and their in-situ electrical resistivity characteristics in liquid Li at high temperatures

  13. Laboratory studies of H retention and LiH formation in liquid lithium

    Energy Technology Data Exchange (ETDEWEB)

    Martín-Rojo, A.B. [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain); UC3M Madrid, 126, 28903 Getafe (Spain); Oyarzabal, E. [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain); U.N.E.D. Ciudad Universitaria, S/N, 28040, Madrid Spain (Spain); Tabarés, F.L., E-mail: tabares@ciemat.es [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain)

    2014-12-15

    Highlights: • Absorption and thermal desorption experiments of hydrogen isotopes in liquid lithium have been performed at exposure temperatures up to 400 °C. • The kinetics of the involved processes indicate a two-stage mechanism for hydride production. • TDS peaks at temperatures well below the expected one for thermal decomposition of the hydride were systematically recorded, although only a small fraction of the absorbed gas was released during the TDS cycle. • The absorption of H{sub 2} in a D{sub 2}-loaded sample was investigated at two temperatures, and no obvious influence of the preexisting species in the rate of absorption of H{sub 2} was seen. • Deuterium absorption takes place at a higher rate than that of hydrogen. - Abstract: Laboratory experiments on H/D retention on liquid lithium followed by thermal desorption spectrometry (TDS) have been performed at Ciemat. Two different experimental set ups were used in order to expose liquid Li to hydrogen gas or to hydrogen glow discharge plasmas at temperatures up to 673 K. In the present work the results concerning the gas phase absorption are addressed. Two different kinetics of absorption were identified from the time evolution of the uptake. Alternate exposures to H{sub 2} and D{sub 2} were carried out in order to study the isotope exchange and its possible use for tritium retention control in Fusion Reactor. Although important differences were found in the absorption kinetics of both species, the total retention seems to be governed by the total sum of hydrogenic isotopes, and only small differences were found in the corresponding TDS spectra, on which evidence of some isotope exchange is observed. The results are discussed in relation to the potential use of liquid lithium walls in a Fusion Reactor.

  14. Laboratory studies of H retention and LiH formation in liquid lithium

    International Nuclear Information System (INIS)

    Martín-Rojo, A.B.; Oyarzabal, E.; Tabarés, F.L.

    2014-01-01

    Highlights: • Absorption and thermal desorption experiments of hydrogen isotopes in liquid lithium have been performed at exposure temperatures up to 400 °C. • The kinetics of the involved processes indicate a two-stage mechanism for hydride production. • TDS peaks at temperatures well below the expected one for thermal decomposition of the hydride were systematically recorded, although only a small fraction of the absorbed gas was released during the TDS cycle. • The absorption of H 2 in a D 2 -loaded sample was investigated at two temperatures, and no obvious influence of the preexisting species in the rate of absorption of H 2 was seen. • Deuterium absorption takes place at a higher rate than that of hydrogen. - Abstract: Laboratory experiments on H/D retention on liquid lithium followed by thermal desorption spectrometry (TDS) have been performed at Ciemat. Two different experimental set ups were used in order to expose liquid Li to hydrogen gas or to hydrogen glow discharge plasmas at temperatures up to 673 K. In the present work the results concerning the gas phase absorption are addressed. Two different kinetics of absorption were identified from the time evolution of the uptake. Alternate exposures to H 2 and D 2 were carried out in order to study the isotope exchange and its possible use for tritium retention control in Fusion Reactor. Although important differences were found in the absorption kinetics of both species, the total retention seems to be governed by the total sum of hydrogenic isotopes, and only small differences were found in the corresponding TDS spectra, on which evidence of some isotope exchange is observed. The results are discussed in relation to the potential use of liquid lithium walls in a Fusion Reactor

  15. A Compact Self-Driven Liquid Lithium Loop for Industrial Neutron Generation

    Science.gov (United States)

    Stemmley, Steven; Szott, Matt; Kalathiparambil, Kishor; Ahn, Chisung; Jurczyk, Brian; Ruzic, David

    2017-10-01

    A compact, closed liquid lithium loop has been developed at the University of Illinois to test and utilize the Li-7(d,n) reaction. The liquid metal loop is housed in a stainless steel trench module with embedded heating and cooling. The system was designed to handle large heat and particle fluxes for use in neutron generators as well as fusion devices, solely operating via thermo-electric MHD. The objectives of this project are two-fold, 1) produce a high energy, MeV-level, neutron source and 2) provide a self-healing, low Z, low recycling plasma facing component. The flowing volume will keep a fresh, clean, lithium surface allowing Li-7(d,n) reactions to occur as well as deuterium adsorption in the fluid, increasing the overall neutron output. Expected yields of this system are 107 n/s for 13.5 MeV neutrons and 108 n/s for 2.45 MeV neutrons. Previous work has shown that using a tapered trench design prevents dry out and allows for an increase in velocity of the fluid at the particle strike point. For heat fluxes on the order of 10's MW/m2, COMSOL models have shown that high enough velocities ( 70 cm/s) are attainable to prevent significant lithium evaporation. Future work will be aimed at addressing wettability issues of lithium in the trenches, experimentally determine the velocities required to prevent dry out, and determine the neutron output of the system. The preliminary results and discussion will be presented. DOE SBIR project DE-SC0013861.

  16. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  17. Soft X-ray emission spectroscopy of liquids and lithium battery materials

    International Nuclear Information System (INIS)

    Augustsson, Andreas

    2004-01-01

    Lithium ion insertion into electrode materials is commonly used in rechargeable battery technology. The insertion implies changes in both the crystal structure and the electronic structure of the electrode material. Side-reactions may occur on the surface of the electrode which is exposed to the electrolyte and form a solid electrolyte interface (SEI). The understanding of these processes is of great importance for improving battery performance. The chemical and physical properties of water and alcohols are complicated by the presence of strong hydrogen bonding. Various experimental techniques have been used to study geometrical structures and different models have been proposed to view the details of how these liquids are geometrically organized by hydrogen bonding. However, very little is known about the electronic structure of these liquids, mainly due to the lack of suitable experimental tools. In this thesis examples of studies of lithium battery electrodes and liquid systems using soft x-ray emission spectroscopy will be presented. Monochromatized synchrotron radiation has been used to accomplish selective excitation, in terms of energy and polarization. The electronic structure of graphite electrodes has been studied, before and after lithium intercalation. Changes in the electronic structure upon lithiation due to transfer of electrons into the graphite π-bands have been observed. Transfer of electrons in to the 3d states of transition metal oxides upon lithiation have been studied, through low energy excitations as dd- and charge transfer-excitations. A SEI was detected on cycled graphite electrodes. By the use of selective excitation different carbon sites were probed in the SEI. The local electronic structure of water, methanol and mixtures of the two have been examined using a special liquid cell, to separate the liquid from the vacuum in the experimental chamber. Results from the study of liquid water showed a strong influence on the 3a1 molecular

  18. Liquid lithium loop system to solve challenging technology issues for fusion power plant

    Science.gov (United States)

    Ono, M.; Majeski, R.; Jaworski, M. A.; Hirooka, Y.; Kaita, R.; Gray, T. K.; Maingi, R.; Skinner, C. H.; Christenson, M.; Ruzic, D. N.

    2017-11-01

    Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s-1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s-1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s-1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s-1 of tritium is needed to maintain the fusion fuel cycle

  19. 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)

  20. Potential design modifications for the High Yield Lithium Injection Fusion Energy (HYLIFE) reaction chamber

    International Nuclear Information System (INIS)

    Pitts, J.H.; Hovingh, J.; Meier, W.R.; Monsler, M.J.; Powell, E.G.; Walker, P.E.

    1979-01-01

    Generation of electric power from inertial confinement fusion requires a reaction chamber. One promising type, the High Yield Lithium Injection Fusion Energy (HYLIFE) chamber, includes a falling array of liquid lithium jets. These jets act as: (1) a renewable first wall and blanket to shield metal components from x-ray and neutron exposure, (2) a tritium breeder to replace tritium burned during the fusion process, and (3) an absorber and transfer medium for fusion energy. Over 90% of the energy produced in the reaction chamber is absorbed in the lithium jet fall. Design aspects are included

  1. Improved Cyclability of Liquid Electrolyte Lithium/Sulfur Batteries by Optimizing Electrolyte/Sulfur Ratio

    Directory of Open Access Journals (Sweden)

    Sheng S. Zhang

    2012-12-01

    Full Text Available A liquid electrolyte lithium/sulfur (Li/S cell is a liquid electrochemical system. In discharge, sulfur is first reduced to highly soluble Li2S8, which dissolves into the organic electrolyte and serves as the liquid cathode. In solution, lithium polysulfide (PS undergoes a series of complicated disproportionations, whose chemical equilibriums vary with the PS concentration and affect the cell’s performance. Since the PS concentration relates to a certain electrolyte/sulfur (E/S ratio, there is an optimized E/S ratio for the cyclability of each Li/S cell system. In this work, we study the optimized E/S ratio by measuring the cycling performance of Li/S cells, and propose an empirical method for determination of the optimized E/S ratio. By employing an electrolyte of 0.25 m LiSO3CF3-0.25 m LiNO3 dissolved in a 1:1 (wt:wt mixture of dimethyl ether (DME and 1,3-dioxolane (DOL in an optimized E/S ratio, we show that the Li/S cell with a cathode containing 72% sulfur and 2 mg cm−2 sulfur loading is able to retain a specific capacity of 780 mAh g−1 after 100 cycles at 0.5 mA cm−2 between 1.7 V and 2.8 V.

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

    Energy Technology Data Exchange (ETDEWEB)

    Zmitko, M., E-mail: milan.zmitko@f4e.europa.eu [Fusion for Energy (F4E), 08019 Barcelona (Spain); Poitevin, Y. [Fusion for Energy (F4E), 08019 Barcelona (Spain); Boccaccini, L., E-mail: lorenzo.boccaccini@inr.fzk.de [Institut Fuer Neutronenphysik und Reaktortechnik, FZK, D-76021 Karlsruhe (Germany); Salavy, J.-F., E-mail: jfsalavy@cea.fr [CEA/Saclay, DEN/DM2S, F-91191 Gif-sur-Yvette (France); Knitter, R., E-mail: regina.knitter@imf.fzk.de [Institut Fuer Materialforschung III, FZK, D-76021 Karlsruhe (Germany); Moeslang, A., E-mail: anton.moeslang@imf.fzk.de [Institut Fuer Materialforschung I, FZK, D-76021 Karlsruhe (Germany); Magielsen, A.J., E-mail: magielsen@nrg.eu [NRG Petten, 1755 ZG Petten (Netherlands); Hegeman, J.B.J. [NRG Petten, 1755 ZG Petten (Netherlands); Laesser, R. [Fusion for Energy (F4E), 08019 Barcelona (Spain)

    2011-10-01

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

  3. A study on liquid lithium flow in rectangular duck perpendicular to a intense magnetic field

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Chen Ke; Liu Yang; Zhang Qinshun

    2001-01-01

    A research on high-speed liquid-metal lithium flow through a non-expanding rectangular duck under uniform intense magnetic field is presented. A equations set with Poisson equation and Helmholtz equation, which control the electrical field and flow field respectively, has been deduced by analysis and PHsolver, a program to solve the equations set, has also been finished. The current density distribution and flow field in the non-expanding rectangular channel with intense magnetic field have been obtained from PHsolver by applying the wall-function in the boundary wall. The velocity profile in the duck appears M-shaped

  4. Investigation of tritium inventory and permeation behaviour in the liquid breeder blanket concept of Demo as a function of design and material parameters

    International Nuclear Information System (INIS)

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

    1991-01-01

    A numerical code has been used to estimate the time dependence of tritium inventory and of tritium transport into the coolant, into the first wall boxes and through the liquid breeder in the Pb-17Li blanket concept of DEMO. Several issues in both design and material parameters have been considered and the effect on inventory and permeation of coatings with low surface recombination coefficient and/or low diffusivity at various surfaces of the structural material has been studied. TiC has been chosen as reference material for these calculations and a general database on coating efficiency as a function of its properties has also been produced on the basis of TiC data

  5. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, B. William [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chiu, Ing L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-26

    Future fusion power plants based on the deuterium-tritium (DT) fuel cycle will be required to breed the T fuel via neutron reactions with lithium, which will be incorporated in a breeding blanket that surrounds the fusion source. Recent work by LLNL proposed the used of liquid Li as the breeder in an inertial fusion energy (IFE) power plant. Subsequently, an LDRD was initiated to develop alternatives ternary alloy liquid metal breeders that have reduced chemical reactivity with water and air compared to pure Li. Part of the work plan was to experimentally investigate the phase diagrams of ternary alloys. Of particular interest was measurement of the melt temperature, which must be low enough to be compatible with the temperature limits of the steel used in the construction of the chamber and heat transfer system.

  6. Interactions between drops of a molten aluminum-lithium alloy and liquid water

    International Nuclear Information System (INIS)

    Nelson, L.S.

    1994-01-01

    In certain hypothesized nuclear reactor accident scenarios, 1- to 10-g drops of molten aluminum-lithium alloys might contact liquid water. Because vigorous steam explosions have occurred when large amounts of molten aluminum-lithium alloys were released into water or other coolants, it becomes important to know whether there will be explosions if smaller amounts of these molten alloys similarly come into contact with water. Therefore, the authors released drops of molten Al-3.1 wt pct Li alloy into deionized water at room temperature. The experiments were performed at local atmospheric pressure (0.085 MPa) without pressure transient triggers applied to the water. The absence of these triggers allowed them to (a) investigate whether spontaneous initiation of steam explosions would occur with these drops and (b) study the alloy-water chemical reactions. The drop sizes and melt temperatures were chosen to simulate melt globules that might form during the hypothesized melting of the aluminum-lithium alloy components

  7. Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

    International Nuclear Information System (INIS)

    Kugel, H.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.P.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.K.; Menard, J.; Stotler, D.; Zakharov, L.E.; Maingi, Rajesh; Nygren, R.E.; Soukhanovskii, V.; Wakeland, P.

    2009-01-01

    Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15 25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW 1), to enable ne scan capability (2) in the H-mode, to test the ability to operate at significantly lower density (e.g., ne/nGW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  8. Spectroscopic diagnostics for liquid lithium divertor studies on National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Soukhanovskii, V. A.; Roquemore, A. L.; Bell, R. E.; Kaita, R.; Kugel, H. W.

    2010-01-01

    The use of lithium-coated plasma facing components for plasma density control is studied in the National Spherical Torus Experiment (NSTX). A recently installed liquid lithium divertor (LLD) module has a porous molybdenum surface, separated by a stainless steel liner from a heated copper substrate. Lithium is deposited on the LLD from two evaporators. Two new spectroscopic diagnostics are installed to study the plasma surface interactions on the LLD: (1) A 20-element absolute extreme ultraviolet (AXUV) diode array with a 6 nm bandpass filter centered at 121.6 nm (the Lyman-α transition) for spatially resolved divertor recycling rate measurements in the highly reflective LLD environment, and (2) an ultraviolet-visible-near infrared R=0.67 m imaging Czerny-Turner spectrometer for spatially resolved divertor D I, Li I-II, C I-IV, Mo I, D 2 , LiD, CD emission and ion temperature on and around the LLD module. The use of photometrically calibrated measurements together with atomic physics factors enables studies of recycling and impurity particle fluxes as functions of LLD temperature, ion flux, and divertor geometry.

  9. Reduced cost design of liquid lithium target for international fusion material irradiation facility (IFMIF)

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

    2001-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) is being jointly planned to provide an accelerator-based D-Li neutron source to produce intense high energy neutrons (2 MW/m 2 ) up to 200 dpa and a sufficient irradiation volume (500 cm 3 ) for testing the candidate materials and components up to about a full lifetime of their anticipated use in ITER and DEMO. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid lithium flow with a speed of 20 m/s. Following Conceptual Design Activity (1995-1998), a design study with focus on cost reduction without changing its original mission has been done in 1999. The following major changes to the CAD target design have been considered in the study and included in the new design: i) number of the Li target has been changed from 2 to 1, ii) spare of impurity traps of the Li loop was removed although the spare will be stored in a laboratory for quick exchange, iii) building volume was reduced via design changes in lithium loop length. This paper describes the reduced cost design of the lithium target system and recent status of Key Element Technology activities. (author)

  10. Numerical analysis of high-speed liquid lithium free-surface flow

    International Nuclear Information System (INIS)

    Gordeev, Sergej; Heinzel, Volker; Stieglitz, Robert

    2012-01-01

    Highlights: ► The free surface behavior of a high speed lithium jet is investigated by means of a CFD LES analysis. ► The study is aiming to validate adequate LES technique. ► The Osaka University experiments with liquid lithium jet have been simulated. ► Four cases with jet flow velocities of 4, 9, 13 and 15 m/s are analyzed. ► Calculation results show a good qualitative and a quantitative agreement with the experimental data. - Abstract: The free-surface stability of the target of the International Fusion Material Irradiation Facility (IFMIF) is one of the crucial issues, since the spatio-temporal behavior of the free-surface determines the neutron flux to be generated. This article investigates the relation between the evolution of a wall boundary layer in a convergent nozzle and the free surface shape of a high speed lithium jet by means of a CFD LES analysis using the Osaka University experiments. The study is aiming to validate adequate LES technique to analyze the individual flow phenomena observed. Four cases with jet flow velocities of 4, 9, 13 and 15 m/s are analyzed. First analyses of calculation results show that the simulation exhibits a good qualitative and a quantitative agreement with the experimental data, which allows in the future a more realistic prediction of the IFMIF target behavior.

  11. Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, W.; Bell, M.; Berzak,L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren,R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov L. E.

    2008-09-26

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW~1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with longpulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  12. Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

    International Nuclear Information System (INIS)

    Kugel, W.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren, R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov, L. E.

    2008-01-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW∼1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization

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

    International Nuclear Information System (INIS)

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

    1995-08-01

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

  14. Ionic diffusion and salt dissociation conditions of lithium liquid crystal electrolytes.

    Science.gov (United States)

    Saito, Yuria; Hirai, Kenichi; Murata, Shuuhei; Kishii, Yutaka; Kii, Keisuke; Yoshio, Masafumi; Kato, Takashi

    2005-06-16

    Salt dissociation conditions and dynamic properties of ionic species in liquid crystal electrolytes of lithium were investigated by a combination of NMR spectra and diffusion coefficient estimations using the pulsed gradient spin-echo NMR techniques. Activation energies of diffusion (Ea) of ionic species changed with the phase transition of the electrolyte. That is, Ea of the nematic phase was lower than that of the isotropic phase. This indicates that the aligned liquid crystal molecules prepared efficient conduction pathways for migration of ionic species. The dissociation degree of the salt was lower compared with those of the conventional electrolyte solutions and polymer gel electrolytes. This is attributed to the low concentration of polar sites, which attract the dissolved salt and promote salt dissociation, on the liquid crystal molecules. Furthermore, motional restriction of the molecules due to high viscosity and molecular oriented configuration in the nematic phase caused inefficient attraction of the sites for the salt. With a decreased dissolved salt concentration of the liquid crystal electrolyte, salt dissociation proceeded, and two diffusion components attributed to the ion and ion pair were detected independently. This means that the exchange rate between the ion and the ion pair is fairly slow once the salt is dissociated in the liquid crystal electrolytes due to the low motility of the medium molecules that initiate salt dissociation.

  15. Development of a high energy pulsed plasma simulator for the study of liquid lithium trenches

    International Nuclear Information System (INIS)

    Jung, S.; Christenson, M.; Curreli, D.; Bryniarski, C.; Andruczyk, D.; Ruzic, D.N.

    2014-01-01

    Highlights: • A pulse device for a liquid lithium trench study is developed. • It consists of a coaxial plasma gun, a theta pinch, and guiding magnets. • A large energy enhancement is observed with the use of the plasma gun. • A further increase in energy and velocity is observed with the theta pinch. - Abstract: To simulate detrimental events in a tokamak and provide a test-stand for a liquid-lithium infused trench (LiMIT) device [1], a pulsed plasma source utilizing a theta pinch in conjunction with a coaxial plasma accelerator has been developed. The plasma is characterized using a triple Langmuir probe, optical methods, and a calorimeter. Clear advantages have been observed with the application of a coaxial plasma accelerator as a pre-ionization source. The experimental results of the plasma gun in conjunction with the existing theta pinch show a significant improvement from the previous energy deposition by a factor of 14 or higher, resulting in a maximum energy and heat flux of 0.065 ± 0.002 MJ/m 2 and 0.43 ± 0.01 GW/m 2 . A few ways to further increase the plasma heat flux for LiMIT experiments are discussed

  16. Scientific opportunities at SARAF with a liquid lithium jet target neutron source

    Science.gov (United States)

    Silverman, Ido; Arenshtam, Alex; Berkovits, Dan; Eliyahu, Ilan; Gavish, Inbal; Grin, Asher; Halfon, Shlomi; Hass, Michael; Hirsh, T. Y.; Kaizer, Boaz; Kijel, Daniel; Kreisel, Arik; Mardor, Israel; Mishnayot, Yonatan; Palchan, Tala; Perry, Amichay; Paul, Michael; Ron, Guy; Shimel, Guy; Shor, Asher; Tamim, Noam; Tessler, Moshe; Vaintraub, Sergey; Weissman, Leo

    2018-05-01

    SARAF (Soreq Applied Research Accelerator Facility) is based on a 5 mA, 40 MeV, proton/deuteron accelerator. Phase-I, operational since 2010, provides proton and deuteron beams up to 4 and 5 MeV, respectively, for basic and applied research activities. The high power Liquid-Lithium jet Target (LiLiT), with 1.912 MeV proton beam, provides high flux quasi-Maxwellian neutrons at kT 30 keV (about 2 × 1010 n/s/cm2/mA on the irradiated sample, about 1 cm from the target), enabling studies of s-process reactions relevant to nucleo-synthesis of the heavy elements in giant AGB stars. With higher energy proton beams and with deuterons, LiLiT can provide higher fluxes of high energy neutrons up to 20 MeV. The experimental program with SARAF phase-I will be enhanced shortly with a new target room complex which is under construction. Finally, SARAF phase-II, planned to start operation at 2023, will enable full capabilities with proton/ deuteron beams at 5 mA and 40 MeV. Liquid lithium targets will then be used to produce neutron sources with intensities of 1015 n/s, which after thermalization will provide thermal neutron (25 meV) fluxes of about 1012 n/s/cm2 at the entrance to neutron beam lines to diffraction and radiography stations.

  17. Development of a high energy pulsed plasma simulator for the study of liquid lithium trenches

    Energy Technology Data Exchange (ETDEWEB)

    Jung, S., E-mail: jung73@illinois.edu [Department of Nuclear, Plasma, and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana 61801 (United States); Christenson, M.; Curreli, D. [Department of Nuclear, Plasma, and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana 61801 (United States); Bryniarski, C. [Department of Electrical and Computer Engineering, University of Illinois at Urbana-Champaign, Urbana 61801 (United States); Andruczyk, D.; Ruzic, D.N. [Department of Nuclear, Plasma, and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana 61801 (United States)

    2014-12-15

    Highlights: • A pulse device for a liquid lithium trench study is developed. • It consists of a coaxial plasma gun, a theta pinch, and guiding magnets. • A large energy enhancement is observed with the use of the plasma gun. • A further increase in energy and velocity is observed with the theta pinch. - Abstract: To simulate detrimental events in a tokamak and provide a test-stand for a liquid-lithium infused trench (LiMIT) device [1], a pulsed plasma source utilizing a theta pinch in conjunction with a coaxial plasma accelerator has been developed. The plasma is characterized using a triple Langmuir probe, optical methods, and a calorimeter. Clear advantages have been observed with the application of a coaxial plasma accelerator as a pre-ionization source. The experimental results of the plasma gun in conjunction with the existing theta pinch show a significant improvement from the previous energy deposition by a factor of 14 or higher, resulting in a maximum energy and heat flux of 0.065 ± 0.002 MJ/m{sup 2} and 0.43 ± 0.01 GW/m{sup 2}. A few ways to further increase the plasma heat flux for LiMIT experiments are discussed.

  18. Rock-salt structure lithium deuteride formation in liquid lithium with high-concentrations of deuterium: a first-principles molecular dynamics study

    Science.gov (United States)

    Chen, Mohan; Abrams, T.; Jaworski, M. A.; Carter, Emily A.

    2016-01-01

    Because of lithium’s possible use as a first wall material in a fusion reactor, a fundamental understanding of the interactions between liquid lithium (Li) and deuterium (D) is important. We predict structural and dynamical properties of liquid Li samples with high concentrations of D, as derived from first-principles molecular dynamics simulations. Liquid Li samples with four concentrations of inserted D atoms (LiDβ , β =0.25 , 0.50, 0.75, and 1.00) are studied at temperatures ranging from 470 to 1143 K. Densities, diffusivities, pair distribution functions, bond angle distribution functions, geometries, and charge transfer between Li and D atoms are calculated and analyzed. The analysis suggests liquid-solid phase transitions can occur at some concentrations and temperatures, forming rock-salt LiD within liquid Li. We also observe formation of some D2 molecules at high D concentrations.

  19. The use of lithium as a marker for the retention of liquids in the oral cavity after rinsing.

    Science.gov (United States)

    Hanning, Sara M; Kieser, Jules A; Ferguson, Martin M; Reid, Malcolm; Medlicott, Natalie J

    2014-01-01

    The aim of this study was to validate the use of lithium as a marker to indicate the retention of simple liquids in the oral cavity and use this to determine how much liquid is retained in the oral cavity following 30 s of rinsing. This is a validation study in which saliva was spiked with known concentrations of lithium. Twenty healthy participants then rinsed their mouths with either water or a 1 % w/v carboxymethylcellulose (CMC) solution for 30 s before expectorating into a collection cup. Total volume and concentration of lithium in the expectorant were then measured, and the percentage of liquid retained was calculated. The mean amount of liquid retained was 10.4 ± 4.7 % following rinsing with water and 15.3 ± 4.1 % following rinsing with 1 % w/v CMC solution. This difference was significant (p < 0.01). Lithium was useful as a marker for the retention of liquids in the oral cavity, and a value for the amount of water and 1 % w/v CMC solution remaining in the oral cavity following a 30-s rinse was established. The present study quantifies the retention of simple fluids in the oral cavity, validating a technique that may be applied to more complex fluids such as mouth rinses. Further, the application of this method to specific population groups such as those with severe xerostomia may assist in developing effective saliva substitutes.

  20. 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

  1. Measurement of hydrogen solubility and desorption rate in V-4Cr-4Ti and liquid lithium-calcium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Park, J.H.; Erck, R.; Park, E.T. [Argonne National Lab., IL (United States)] [and others

    1997-04-01

    Hydrogen solubility in V-4Cr-4Ti and liquid lithium-calcium was measured at a hydrogen pressure of 9.09 x 10{sup {minus}4} torr at temperatures between 250 and 700{degrees}C. Hydrogen solubility in V-4Cr-4Ti and liquid lithium decreased with temperature. The measured desorption rate of hydrogen in V-4Cr-4Ti is a thermally activated process; the activation energy is 0.067 eV. Oxygen-charged V-4Cr-4Ti specimens were also investigated to determine the effect of oxygen impurity on hydrogen solubility and desorption in the alloy. Oxygen in V-4Cr-4Ti increases hydrogen solubility and desorption kinetics. To determine the effect of a calcium oxide insulator coating on V-4Cr-4Ti, hydrogen solubility in lithium-calcium alloys that contained 0-8.0 percent calcium was also measured. The distribution ratio R of hydrogen between liquid lithium or lithium-calcium and V-4Cr-4Ti increased as temperature decreased (R {approx} 10 and 100 at 700 and 250{degrees}C, respectively). However at <267{degrees}C, solubility data could not be obtained by this method because of the slow kinetics of hydrogen permeation through the vanadium alloy.

  2. Measurement of hydrogen solubility and desorption rate in V-4Cr-4Ti and liquid lithium-calcium alloys

    International Nuclear Information System (INIS)

    Park, J.H.; Erck, R.; Park, E.T.

    1997-01-01

    Hydrogen solubility in V-4Cr-4Ti and liquid lithium-calcium was measured at a hydrogen pressure of 9.09 x 10 -4 torr at temperatures between 250 and 700 degrees C. Hydrogen solubility in V-4Cr-4Ti and liquid lithium decreased with temperature. The measured desorption rate of hydrogen in V-4Cr-4Ti is a thermally activated process; the activation energy is 0.067 eV. Oxygen-charged V-4Cr-4Ti specimens were also investigated to determine the effect of oxygen impurity on hydrogen solubility and desorption in the alloy. Oxygen in V-4Cr-4Ti increases hydrogen solubility and desorption kinetics. To determine the effect of a calcium oxide insulator coating on V-4Cr-4Ti, hydrogen solubility in lithium-calcium alloys that contained 0-8.0 percent calcium was also measured. The distribution ratio R of hydrogen between liquid lithium or lithium-calcium and V-4Cr-4Ti increased as temperature decreased (R ∼ 10 and 100 at 700 and 250 degrees C, respectively). However at <267 degrees C, solubility data could not be obtained by this method because of the slow kinetics of hydrogen permeation through the vanadium alloy

  3. Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Brooks, A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Lopes-Cardozo, N. [TU/Eindhoven, Eindhoven (Netherlands); Menard, J.; Ono, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Rindt, P. [TU/Eindhoven, Eindhoven (Netherlands); Tresemer, K. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2016-11-15

    Highlights: • An upgrade path for the NSTX-U tokamak is proposed that maintains scientific productivity while enabling exploration of novel, liquid metal PFC. • Pre-filled liquid metal divertor targets are proposed as an intermediate step that mitigates technical and scientific risks associated with liquid metal PFC. • Analysis of leading edge features show a strong link between engineering design considerations and expected performance as a PFC. • A method for optimizing porous liquid metal targets restrained by capillary forces is provided indicating pore-sizes well within current technical capabilities. - Abstract: Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. Two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

  4. Neutronic performance of two European breeder-inside-tube (BIT) blankets for DEMO: the helium-cooled ceramic LiAlO2 with Be multiplier and the water-cooled liquid Li17Pb

    International Nuclear Information System (INIS)

    Petrizzi, L.; Rado, V.

    1995-01-01

    In support of ENEA activity in the European Community Test Programme, neutron analysis has been performed on the two latest blanket designs: helium-cooled ceramic breeder-inside-tube (BIT) (with LiAlO 2 and Be multiplier) and water-cooled liquid Li 17 Pb in cylindrical modules (CM). The powerful MCNP Monte Carlo code was used (version 4.2). A detailed and accurate description of the geometrical model has been performed by inserting the main reactor details and avoiding breeder material dilution inside the modules. The tritium breeding ratio (TBR) performance is low for the solid breeder BIT blanket (with 10 ports 1.011) due mainly to low blanket coverage near the exhaust duct, and this solution should be revised. The CM Li 17 Pb blanket reaches a sufficient TBR (1.059, with ports) to rely on tritium self-sufficiency. Shielding properties, with respect to the toroidal field coils, have been estimated in a simplified model by means of the ANISN code, supplied with a nuclear data library consistent with that used by MCNP. The analysis suggests that a careful shield thickness/composition design should be used to ensure the shielding capability of the whole blanket plus shield system. (orig.)

  5. Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

    Energy Technology Data Exchange (ETDEWEB)

    Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Jaworski, M.A.; Kaita, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Hirooka, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Gray, T.K. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2017-04-15

    Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety concerns (Federici et al., 2001) . It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance (Ono et al., 2013, 2014) . The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium (LL) divertor (RLLD) concept (Ono et al., 2013) and its variant, the active liquid lithium divertor concept (ARLLD) (Ono et al., 2014) , taking advantage of the enhanced non-coronal Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/s of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤450 °C than the first wall ∼600–700 °C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ∼1 l/s (l/s) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust/impurities are removed by relatively simple filter and cold/hot trap systems. Using a

  6. 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

  7. The solid-liquid extraction separation of lithium isotopes by porous composite materials doped with ionic liquids and 2,2'-binaphthyldiyl-17-crown-5

    International Nuclear Information System (INIS)

    Xiao-Li Sun; Ling Gu; Dan Qiu; Dong-Hong Ren; Zaijun Li; Zhi-Guo Gu; Jiangnan University, Wuxi

    2015-01-01

    A green and efficient solid-liquid extraction method of lithium isotopes separation by porous composite materials doped with imidazolium ionic liquids and 2,2'-binaphthyldiyl-17-crown-5 has been reported in this paper. The composite materials of mesoporous silica and impregnated resin were synthesized by sol-gel and direct impregnation process, respectively. Various extraction parameters such as the concentration of lithium salt, anion of lithium salt, initial pH, time and temperature were investigated. Under optimized conditions, the maximum single-stage separation factor of 6 Li/ 7 Li was 1.048 ± 0.002, the maximum extraction efficiency was 15.86 %. The sorbents can be regenerated easily with HCl solution and reused repeatedly. (author)

  8. 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.)

  9. 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

  10. Green and efficient extraction strategy to lithium isotope separation with double ionic liquids as the medium and ionic associated agent

    International Nuclear Information System (INIS)

    Xu Jingjing; Li Zaijun; Gu Zhiguo; Wang Guangli; Liu Junkang

    2013-01-01

    The paper reported a green and efficient extraction strategy to lithium isotope separation. A 4-methyl-10-hydroxybenzoquinoline (ROH), hydrophobic ionic liquid-1,3-di(isooctyl)imidazolium hexafluorophosphate ([D(i-C 8 )IM][PF 6 ]), and hydrophilic ionic liquid-1-butyl-3-methylimidazolium chloride (ILCl) were used as the chelating agent, extraction medium and ionic associated agent. Lithium ion (Li + ) first reacted with ROH in strong alkali solution to produce a lithium complex anion. It then associated with IL + to form the Li(RO) 2 IL complex, which was rapidly extracted into the organic phase. Factors for effect on the lithium isotope separation were examined. To obtain high extraction efficiency, a saturated ROH in the [D(i-C 8 )IM][PF 6 ] (0.3 mol l -1 ), mixed aqueous solution containing 0.3 mol l -1 lithium chloride, 1.6 mol l -1 sodium hydroxide and 0.8 mol l -1 ILCl and 3:1 were selected as the organic phase, aqueous phase and phase ratio (o/a). Under optimized conditions, the single-stage extraction efficiency was found to be 52 %. The saturated lithium concentration in the organic phase was up to 0.15 mol l -1 . The free energy change (ΔG), enthalpy change (ΔH) and entropy change (ΔS) of the extraction process were -0.097 J mol -1 , -14.70 J mol K -1 and -48.17 J mol -1 K -1 , indicating a exothermic process. The partition coefficients of lithium will enhance with decrease of the temperature. Thus, a 25 deg C of operating temperature was employed for total lithium isotope separation process. Lithium in Li(RO) 2 IL was stripped by the sodium chloride of 5 mol l -1 with a phase ratio (o/a) of 4. The lithium isotope exchange reaction in the interface between organic phase and aqueous phase reached the equilibrium within 1 min. The single-stage isotope separation factor of 7 Li- 6 Li was up to 1.023 ± 0.002, indicating that 7 Li was concentrated in organic phase and 6 Li was concentrated in aqueous phase. All chemical reagents used can be well recycled

  11. Effect of lithium tetrafluoroborate on the solubility of carbon dioxide in the ionic liquid 1-butyl-3-methylimidazolium tetrafluoroborate

    NARCIS (Netherlands)

    Durano Arno, S.; Lucas, S.; Shariati - Sarabi, A.; Peters, C.J.

    2012-01-01

    In this work, the phase behavior of the ternary system of carbon dioxide +1-butyl-3-methylimidazolium tetrafluoroborate + lithium tetrafluoroborate has been investigated. Mixtures of known concentrations of the salt, ionic liquid and carbon dioxide were prepared and their bubble point pressures were

  12. Energy deposition in liquid metals for D-T, D-D and T-T fusion sources

    International Nuclear Information System (INIS)

    Ragheb, M.M.H.; Zahakaylo, D.

    1983-01-01

    The nuclear performance of candidate liquid metals: lithium, lead, sodium, potassiu, Na(22%) K(78%), Na(56%) K(44%), is estimated with respect to their neutron and gamma-ray heat deposition rates. Three different neutron sources are considered: DT, DD and TT fusion neutrons. This is intended for the cooling of inertial confinement cavities using fusion pellets with internal tritrium breeding that will possibly eliminate the need to breed tritium in a lithium blanket. Compared to lithium with respect to neutron and gamma energy generation, blanket multiplication and pumping power, it appears that the considered metals can be used only if the environmental and safety advantages from the reduction of the tritium inventory and the avoidance of lithium, outweight the lithium advantages in higher energy production and lower pumping requirement by one to two orders of magnitude. (orig.) [de

  13. The production of lithium oxide microspheres from the disintegration of a liquid jet

    International Nuclear Information System (INIS)

    Al-Ubaidi, M.R.; Anno, J.N.

    1989-01-01

    Microspheres of lithium hydroxide (LiOH) were produced from in-flight solidification of droplets formed by the disintegration of an acoustically driven, mechanically vibrated cylindrical liquid jet of molten LiOH. The molten material at 470 to 480 degrees C was fed through a 25-gauge (0.0267-cm bore diameter) nozzle, interiorly electroplated with silver, under ∼27.6-kPa (4-psig) pressure, and at a mechanical vibration frequency of 10 Hz. The resulting jet issued into a 5.5-cm-diam vertical glass drop tube entraining a 94.5 cm 3 /s (12 ft 3 /h) argon gas stream at 75 degrees C. The 100-cm-long drop tube was sufficient to allow the droplets of molten LiOH resulting from jet disintegration to solidify in-flight without catastrophic thermal shock, being then collected a solid microspheres. These LiOH microspheres were then vacuum processed to lithium oxide (Li 2 O). Preliminary experiments resulted in microspheres with diameters varying from 120 to 185 μm, but with evidence of impurity contamination occurring during the initial stages of the process

  14. Gelled Electrolyte Containing Phosphonium Ionic Liquids for Lithium-Ion Batteries

    Directory of Open Access Journals (Sweden)

    Mélody Leclère

    2018-06-01

    Full Text Available In this work, new gelled electrolytes were prepared based on a mixture containing phosphonium ionic liquid (IL composed of trihexyl(tetradecylphosphonium cation combined with bis(trifluoromethanesulfonimide [TFSI] counter anions and lithium salt, confined in a host network made from an epoxy prepolymer and amine hardener. We have demonstrated that the addition of electrolyte plays a key role on the kinetics of polymerization but also on the final properties of epoxy networks, especially thermal, thermo-mechanical, transport, and electrochemical properties. Thus, polymer electrolytes with excellent thermal stability (>300 °C combined with good thermo-mechanical properties have been prepared. In addition, an ionic conductivity of 0.13 Ms·cm−1 at 100 °C was reached. Its electrochemical stability was 3.95 V vs. Li0/Li+ and the assembled cell consisting in Li|LiFePO4 exhibited stable cycle properties even after 30 cycles. These results highlight a promising gelled electrolyte for future lithium ion batteries.

  15. Selective solid-liquid extraction of lithium halide salts using a ditopic macrobicyclic receptor.

    Science.gov (United States)

    Mahoney, Joseph M; Beatty, Alicia M; Smith, Bradley D

    2004-11-29

    A ditopic salt receptor that is known to bind and extract solid NaCl, KCl, NaBr, and KBr into organic solution as their contact ion pairs is now shown by NMR and X-ray crystallography to bind and extract solid LiCl and LiBr as water-separated ion pairs. The receptor can transport these salts from an aqueous phase through a liquid organic membrane with a cation selectivity of K+ > Na+ > Li+. However, the selectivity order is strongly reversed when the receptor extracts solid alkali metal chlorides and bromides into organic solution. For a three-component mixture of solid LiCl, NaCl, and KCl, the ratio of salts extracted and complexed to the receptor in CDCl3 was 94:4:2, respectively. The same strong lithium selectivity was also observed in the case of a three-component mixture of solid LiBr, NaBr, and KBr where the ratio of extracted salts was 92:5:3. This observation is attributed to the unusually high solubility of lithium salts in organic solvents. The study suggests that ditopic receptors with an ability to extract solid salts as associated ion pairs may have application in separation processes.

  16. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

  17. Direct LiT Electrolysis in a Metallic Fusion Blanket

    International Nuclear Information System (INIS)

    Olson, Luke

    2016-01-01

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

  18. 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.

  19. 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

  20. Tritium breeding experiments with lithium titanate in thermal-type mockups

    International Nuclear Information System (INIS)

    Klix, Axel; Takahashi, Akito; Ochiai, Kentaro; Nishitani, Takeo

    2004-01-01

    Lithium titanate, an advanced tritium breeding material, is currently investigated in integral mock-up experiments at FNS. A method was developed which allows to measure low tritium concentrations directly in this material. The local tritium production rate was obtained by small lithium titanate pellet detectors inserted into the breeding layers which are dissolved after irradiation of the assemblies, and the accumulated tritium was counted by liquid scintillation techniques. The measurement method was applied in mock0-up experiments with candidate materials for the future DEMO reactor breeding blanket. Experimental assemblies consisted of sheets of low activation ferritic steel F82H, lithium titanate, and berylium. Tritium production rate profiles were obtained and compared with results from calculations with the Monte Carlo neutron transport code MCNP-4C. In case of the mock-ups with 95% enriched lithium titanate, the C/E ratios were within the error estimate while larger discrepancies were observed in case of 40% enriched lithium titanate. (author)

  1. Liquid lithium target as a high intensity, high energy neutron source

    Science.gov (United States)

    Parkin, Don M.; Dudey, Norman D.

    1976-01-01

    This invention provides a target jet for charged particles. In one embodiment the charged particles are high energy deuterons that bombard the target jet to produce high intensity, high energy neutrons. To this end, deuterons in a vacuum container bombard an endlessly circulating, free-falling, sheet-shaped, copiously flowing, liquid lithium jet that gushes by gravity from a rectangular cross-section vent on the inside of the container means to form a moving web in contact with the inside wall of the vacuum container. The neutrons are produced via break-up of the beam in the target by stripping, spallation and compound nuclear reactions in which the projectiles (deuterons) interact with the target (Li) to produce excited nuclei, which then "boil off" or evaporate a neutron.

  2. Liquid lithium target as a high intensity, high energy neutron source

    International Nuclear Information System (INIS)

    Parkin, D.M.; Dudey, N.D.

    1976-01-01

    The invention described provides a target jet for charged particles. In one embodiment the charged particles are high energy deuterons that bombard the target jet to produce high intensity, high energy neutrons. To this end, deuterons in a vacuum container bombard an endlessly circulating, free-falling, sheet-shaped, copiously flowing, liquid lithium jet that gushes by gravity from a rectangular cross-section vent on the inside of the container means to form a moving web in contact with the inside wall of the vacuum container. The neutrons are produced via break-up of the beam in the target by stripping, spallation and compound nuclear reactions in which the projectiles (deuterons) interact with the target (Li) to produce excited nuclei, which then ''boil off'' or evaporate a neutron

  3. Tensile property of low activation vanadium alloy after liquid lithium exposure

    International Nuclear Information System (INIS)

    Nagasaka, Takuya; Muroga, Takeo; Li, Meimei; Hoelzer, David T.; Zinkle, Steven J.; Grossbeck, Martin L.; Matsui, Hideki

    2005-10-01

    A candidate low activation vanadium (V) alloy, V-4Cr-4Ti (NIFS-HEAT-2), was exposed to liquid lithium (Li) at 973 and 1073 K for up to 1963 hr. Contamination by carbon (C) and nitrogen (N) from the Li on the order of thousands of wppm were observed. Oxygen (O) levels were reduced to the several 10 wppm level by Li exposure at 1073 K, but not at 973 K. The Li exposure caused strength degradation as measured by tensile tests at 973 and 1073 K. On the other hand, good ductility was demonstrated after the Li exposure even with the significant contamination of C and N. From microstructural observations, C and N are likely to be scavenged by Ti-C-N type precipitates. Reduction of O was attributed to disappearance of Ti-C-O type precipitates. (author)

  4. Structure and electrical resistivity of alkali-alkali and lithium-based liquid binary alloys

    International Nuclear Information System (INIS)

    Mishra, A.K.; Mukherjee, K.K.

    1990-01-01

    Harmonic model potential, developed and used for simple metals is applied here to evaluate hardsphere diameters, which ensure minimum interionic pair potential for alkali-alkali (Na-K, Na-Rb, Na-Cs, K-Rb, K-Cs) and lithium-based (Li-Na, Li-Mg, Li-In, Li-Tl) liquid binary alloys as a function of composition for use in the determination of their partial structure factors. These structure factors are then used to calculate electrical resistivities of alloys considered. The computed values of electrical resistivity as a function of composition agree both, in magnitude and gradient reasonably well with experimental values in all cases except in Cs systems, where the disagreement is appreciable. (author)

  5. Study on surface wave characteristics of free surface flow of liquid metal lithium for IFMIF

    International Nuclear Information System (INIS)

    Hoashi, Eiji; Sugiura, Hirokazu; Yoshihashi-Suzuki, Sachiko; Yamaoka, Nobuo; Horiike, Hiroshi; Kanemura, Takuji; Kondo, Hiroo

    2011-01-01

    The international fusion materials irradiation facility (IFMIF) presents an intense neutron source to develop fusion reactor materials. The free surface flow of a liquid metal Lithium (Li) is planned as a target irradiated by two deuteron beams to generate intense neutrons and it is thus important to obtain knowledge of the surface wave characteristic for the safety and the efficiency of system in the IFMIF. We have been studying on surface wave characteristics experimentally using the liquid metal Li circulation facility at Osaka University and numerically using computational fluid dynamics (CFD) code, FLUENT. This paper reports the results of the surface fluctuation, the wave height and the surface velocity in the free surface flow of the liquid metal Li examined experimentally and numerically. In the experiment, an electro-contact probe apparatus was used to obtain the surface fluctuation and the wave height, and a high speed video was used to measure the surface velocity. We resulted in knowledge of the surface wave growth mechanism. On the other hand, a CFD simulation was also conducted to obtain information on the relation of the free surface with the inner flow. In the simulation, the model included from a two-staged contraction nozzle to a flow channel with a free surface flow region and simulation results were compared with the experimental data. (author)

  6. Self-consistent description of local density dynamics in simple liquids. The case of molten lithium.

    Science.gov (United States)

    Mokshin, A V; Galimzyanov, B N

    2018-02-28

    The dynamic structure factor is the quantity, which can be measured by means of Brillouin light-scattering as well as by means of inelastic scattering of neutrons and x-rays. The spectral (or frequency) moments of the dynamic structure factor define directly the sum rules of the scattering law. The theoretical scheme formulated in this study allows one to describe the dynamics of local density fluctuations in simple liquids and to obtain the expression of the dynamic structure factor in terms of the spectral moments. The theory satisfies all the sum rules, and the obtained expression for the dynamic structure factor yields correct extrapolations into the hydrodynamic limit as well as into the free-particle dynamics limit. We discuss correspondence of this theory with the generalized hydrodynamics and with the viscoelastic models, which are commonly used to analyze the data of inelastic neutron and x-ray scattering in liquids. In particular, we reveal that the postulated condition of the viscoelastic model for the memory function can be directly obtained within the presented theory. The dynamic structure factor of liquid lithium is computed on the basis of the presented theory, and various features of the scattering spectra are evaluated. It is found that the theoretical results are in agreement with inelastic x-ray scattering data.

  7. Self-consistent description of local density dynamics in simple liquids. The case of molten lithium

    Science.gov (United States)

    Mokshin, A. V.; Galimzyanov, B. N.

    2018-02-01

    The dynamic structure factor is the quantity, which can be measured by means of Brillouin light-scattering as well as by means of inelastic scattering of neutrons and x-rays. The spectral (or frequency) moments of the dynamic structure factor define directly the sum rules of the scattering law. The theoretical scheme formulated in this study allows one to describe the dynamics of local density fluctuations in simple liquids and to obtain the expression of the dynamic structure factor in terms of the spectral moments. The theory satisfies all the sum rules, and the obtained expression for the dynamic structure factor yields correct extrapolations into the hydrodynamic limit as well as into the free-particle dynamics limit. We discuss correspondence of this theory with the generalized hydrodynamics and with the viscoelastic models, which are commonly used to analyze the data of inelastic neutron and x-ray scattering in liquids. In particular, we reveal that the postulated condition of the viscoelastic model for the memory function can be directly obtained within the presented theory. The dynamic structure factor of liquid lithium is computed on the basis of the presented theory, and various features of the scattering spectra are evaluated. It is found that the theoretical results are in agreement with inelastic x-ray scattering data.

  8. Development of windowless liquid lithium targets for fragmentation and fission of 400-kW uranium beams

    CERN Document Server

    Nolen, J A; Hassanein, A; Novick, V J; Plotkin, P; Specht, J R

    2003-01-01

    The driver linac of the proposed rare isotope accelerator facility is designed to deliver 2x10 sup 1 sup 3 uranium ions per second at 400 MeV/u on target for radionuclide production via the fission and fragmentation mechanisms. The ion optics of the large acceptance, high-resolution fragment separators that follow the production target require primary beam spot widths of 1 mm. To cope with the resulting high power densities, windowless liquid lithium targets are being developed. The present designs build on existing experience with liquid lithium and liquid sodium systems that have been used for fusion and fission applications. However, no completely windowless systems have been developed or tested to date. For the beam power indicated above (400 kW), the flow requirements are up to about 20 m/s and 10 l/s linear and volume flow rates, respectively. The required target thickness is 1-1.5 g/cm sup 2 (2-3 cm lithium thickness). At this time a prototype windowless system with a lithium thickness of 1-2 cm is und...

  9. Vapor condensation on the surface of a liquid blanket jet in an inertial-confinement fusion reactor

    International Nuclear Information System (INIS)

    Takahashi, Minoru; Inoue, Akira; Fujinuma, Hajime; Tsukui, Jun.

    1991-01-01

    As the fundamental study on lithium jet cooling of an inertial-confinement fusion reactor, the experiment was performed to investigate for the steady condensation of saturated steam on a vertical downward water jet. The experimental parameters were the nozzle diameter of 3 and 5 mm, the jet length of 60∼316 mm, the outlet velocity of 2∼12 m/s, the outlet temperature of 30∼70degC, and the pressure of 0.03∼0.44 MPa, which corresponds to the Reynolds number of 1.35 x 10 4 ∼2.71 x 10 5 and the Prandtl number of 1.0∼5.2. As the Reynolds number or the jet length is increased, the Stanton number decreases and then increases again. As the steam pressure is increased, it increases monotonously. These characteristics of condensation heat transfer have been classical into four regions based on the criteria for jet break-up and surface disturbance, or entrainment. The empirical correlations for the Stanton number have been obtained for these regions, and the validity was confirmed by comparing them with the previous correlations. (author)

  10. Effect of lithium salts addition on the ionic liquid based extraction of essential oil from Farfarae Flos.

    Science.gov (United States)

    Li, Zhen-Yu; Zhang, Sha-Sha; Jie-Xing; Qin, Xue-Mei

    2015-01-01

    In this study, an ionic liquids (ILs) based extraction approach has been successfully applied to the extraction of essential oil from Farfarae Flos, and the effect of lithium chloride was also investigated. The results indicated that the oil yields can be increased by the ILs, and the extraction time can be reduced significantly (from 4h to 2h), compared with the conventional water distillation. The addition of lithium chloride showed different effect according to the structures of ILs, and the oil yields may be related with the structure of cation, while the chemical compositions of essential oil may be related with the anion. The reduction of extraction time and remarkable higher efficiency (5.41-62.17% improved) by combination of lithium salt and proper ILs supports the suitability of the proposed approach. Copyright © 2014 Elsevier B.V. All rights reserved.

  11. 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

  12. Liquid lithium surface control and its effect on plasma performance in the HT-7 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, G.Z.; Ren, J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, J.S., E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, Z.; Yang, Q.X.; Li, J.G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zakharov, L.E. [Princeton University Plasma Physics Laboratory Princeton, NJ 08543 (United States); Ruzic, David N. [University of Illinois, Urbana, IL 61801 (United States)

    2014-12-15

    Highlights: • Strong interaction between plasma and Li would cause strong Li emission and lead to disruptive plasmas, and probable reasons were analyzed. • Serious Li would be emitted from the free statics surface mainly due to J × B force leading to plasma instable and disruptions. • CPS surface would partially suppress the emission and be beneficial for plasma operation. • Li emission from flowing LLLs on free surfaces on SS trenches and on SS plate were compared. - Abstract: Experiments with liquid lithium limiters (LLLs) have been successfully performed in HT-7 since 2009 and the effects of different limiter surface structures on the ejection of Li droplets have been studied and compared. The experiments have demonstrated that strong interaction between the plasma and the liquid surface can cause intense Li efflux in the form of ejected Li droplets – which can, in turn, lead to plasma disruptions. The details of the LLL plasma-facing surface were observed to be extremely important in determining performance. Five different LLLs were evaluated in this work: two types of static free-surface limiters and three types of flowing liquid Li (FLLL) structures. It has been demonstrated that a FLLL with a slowly flowing thin liquid Li film on vertical flow plate which was pre-treated with evaporated Li was much less susceptible to Li droplet ejection than any of the other structures tested in this work. It was further observed that the plasmas run against this type of limiter were reproducibly well-behaved. These results provide technical references for the design of FLLLs in future tokamaks so as to avoid strong Li ejection and to decrease disruptive plasmas.

  13. Conceptual Design of Main Cooling System for a Fusion Power Reactor with Water Cooled Lithium-Lead Blanket. TW1-TRP-PPCS1, Deliverable 8

    International Nuclear Information System (INIS)

    Natalizio, Antonio; Collen, Jan

    2002-06-01

    The HTS (Heat Transfer System) conceptual design developed for the PPCS (Power-Plant Conceptual Study) plant model is compliant with the single failure criterion - i.e., the failure of a single active component (e.g., pump) will not cause the reactor to shutdown. The system effective availability (capacity factor), however, is only marginally better than that of the SEAFP design, as the number of loops could not be decreased further, due to coolant inventory limitations. The PPCS Plant Model A has about 70 % more fusion power than the SEAFP model. Therefore, keeping the same number of loops as in the SEAFP model would have implied a 70 % larger inventory. To improve plant availability and safety, however, the number of blanket and first wall loops have been reduced from eight to six, implying a further increase in loop inventory of about 25 %. For these and other reasons, the coolant inventory, at risk from a loss-of-coolant accident, has increased significantly, relative to the SEAFP design (∼130 vs. 50 m 3 ). The proposed heat transport system conceptual design meets, or exceeds, all project specifications

  14. Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

    Science.gov (United States)

    McLean, A. G.; Gan, K. F.; Ahn, J.-W.; Gray, T. K.; Maingi, R.; Abrams, T.; Jaworski, M. A.; Kaita, R.; Kugel, H. W.; Nygren, R. E.; Skinner, C. H.; Soukhanovskii, V. A.

    2013-07-01

    Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of Tsurface near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q⊥,peak = 5 MW/m2 inter-ELM and up to 10 MW/m2 during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

  15. Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

    Energy Technology Data Exchange (ETDEWEB)

    McLean, A.G., E-mail: mclean@fusion.gat.com [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Gan, K.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Ahn, J.-W.; Gray, T.K.; Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Abrams, T.; Jaworski, M.A.; Kaita, R.; Kugel, H.W. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

    2013-07-15

    Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of T{sub surface} near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q{sub ⊥,peak} = 5 MW/m{sup 2} inter-ELM and up to 10 MW/m{sup 2} during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

  16. Protected Lithium-Metal Anodes in Batteries: From Liquid to Solid.

    Science.gov (United States)

    Yang, Chunpeng; Fu, Kun; Zhang, Ying; Hitz, Emily; Hu, Liangbing

    2017-09-01

    High-energy lithium-metal batteries are among the most promising candidates for next-generation energy storage systems. With a high specific capacity and a low reduction potential, the Li-metal anode has attracted extensive interest for decades. Dendritic Li formation, uncontrolled interfacial reactions, and huge volume effect are major hurdles to the commercial application of Li-metal anodes. Recent studies have shown that the performance and safety of Li-metal anodes can be significantly improved via organic electrolyte modification, Li-metal interface protection, Li-electrode framework design, separator coating, and so on. Superior to the liquid electrolytes, solid-state electrolytes are considered able to inhibit problematic Li dendrites and build safe solid Li-metal batteries. Inspired by the bright prospects of solid Li-metal batteries, increasing efforts have been devoted to overcoming the obstacles of solid Li-metal batteries, such as low ionic conductivity of the electrolyte and Li-electrolyte interfacial problems. Here, the approaches to protect Li-metal anodes from liquid batteries to solid-state batteries are outlined and analyzed in detail. Perspectives regarding the strategies for developing Li-metal anodes are discussed to facilitate the practical application of Li-metal batteries. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  17. A 20 kw beam-on-target test of a high-power liquid lithium target for RIA

    International Nuclear Information System (INIS)

    Reed, Claude B.; Nolen, Jerry A.; Specht, James R.; Novick, Vincent J.; Plotkin, Perry

    2004-01-01

    The high-power heavy-ion beams produced by the Rare Isotope Accelerator (RIA) driver linac have large energy deposition density in solids and in many cases no solid materials would survive the full beam power. Liquid lithium technology has been proposed to solve this problem in RIA. Specifically, a windowless target for the production of radioactive ions via fragmentation, consisting of a jet of about 3 cm thickness of flowing liquid lithium, exposed to the beamline vacuum [1,2] is being developed. To demonstrate that power densities equivalent to a 200-kW RIA uranium beam, deposited in the first 4 mm of a flowing lithium jet, can be handled by the windowless target design, a high power 1 MeV Dynamitron was leased and a test stand prepared to demonstrate the target's capability of absorbing and carrying away a 20kW heat load without disrupting either the 5 mm x 10 mm flowing lithium jet target or the beam line vacuum

  18. New Ether-functionalized Morpholinium- and Piperidinium-based Ionic Liquids as Electrolyte Components in Lithium and Lithium-Ion Batteries.

    Science.gov (United States)

    Navarra, Maria Assunta; Fujimura, Kanae; Sgambetterra, Mirko; Tsurumaki, Akiko; Panero, Stefania; Nakamura, Nobuhumi; Ohno, Hiroyuki; Scrosati, Bruno

    2017-06-09

    Here, two ionic liquids, N-ethoxyethyl-N-methylmorpholinium bis(trifluoromethanesulfonyl)imide (M 1,2O2 TFSI) and N-ethoxyethyl-N-methylpiperidinium bis(trifluoromethanesulfonyl)imide (P 1,2O2 TFSI) were synthesized and compared. Fundamental relevant properties, such as thermal and electrochemical stability, density, and ionic conductivity were analyzed to evaluate the effects caused by the presence of the ether bond in the side chain and/or in the organic cation ring. Upon lithium salt addition, two electrolytes suitable for lithium batteries applications were found. Higher conducting properties of the piperidinium-based electrolyte resulted in enhanced cycling performances when tested with LiFePO 4 (LFP) cathode in lithium cells. When mixing the P 1,2O2 TFSI/LiTFSI electrolyte with a tailored alkyl carbonate mixture, the cycling performance of both Li and Li-ion cells greatly improved, with prolonged cyclability delivering very stable capacity values, as high as the theoretical one in the case of Li/LFP cell configurations. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  19. 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

  20. Electrospun polymer membrane activated with room temperature ionic liquid: Novel polymer electrolytes for lithium batteries

    Science.gov (United States)

    Cheruvally, Gouri; Kim, Jae-Kwang; Choi, Jae-Won; Ahn, Jou-Hyeon; Shin, Yong-Jo; Manuel, James; Raghavan, Prasanth; Kim, Ki-Won; Ahn, Hyo-Jun; Choi, Doo Seong; Song, Choong Eui

    A new class of polymer electrolytes (PEs) based on an electrospun polymer membrane incorporating a room-temperature ionic liquid (RTIL) has been prepared and evaluated for suitability in lithium cells. The electrospun poly(vinylidene fluoride- co-hexafluoropropylene) P(VdF-HFP) membrane is activated with a 0.5 M solution of LiTFSI in 1-butyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide (BMITFSI) or a 0.5 M solution of LiBF 4 in 1-butyl-3-methylimidazolium tetrafluoroborate (BMIBF 4). The resulting PEs have an ionic conductivity of 2.3 × 10 -3 S cm -1 at 25 °C and anodic stability at >4.5 V versus Li +/Li, making them suitable for practical applications in lithium cells. A Li/LiFePO 4 cell with a PE based on BMITFSI delivers high discharge capacities when evaluated at 25 °C at the 0.1 C rate (149 mAh g -1) and the 0.5 C rate (132 mAh g -1). A very stable cycle performance is also exhibited at these low current densities. The properties decrease at the higher, 1 C rate, when operated at 25 °C. Nevertheless, improved properties are obtained at a moderately elevated temperature of operation, i.e. 40 °C. This is attributed to enhanced conductivity of the electrolyte and faster reaction kinetics at higher temperatures. At 40 °C, a reversible capacity of 140 mAh g -1 is obtained at the 1 C rate.

  1. 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

  2. Vapor-liquid equilibria of the water + 1,3-propanediol and water + 1,3-propanediol + lithium bromide systems

    Energy Technology Data Exchange (ETDEWEB)

    Mun, S Y; Lee, H

    1999-12-01

    Vapor-liquid equilibrium data of the water + 1,3-propanediol and water + 1,3-propanediol + lithium bromide systems were measured at 60, 160, 300, and 760 mmHg at temperatures ranging from 315 to 488 K. The apparatus used in this work is a modified still especially designed for the measurement of low-pressure VLE, in which both liquid and vapor are continuously recirculated. For the analysis of salt-containing solutions, a method incorporating refractometry and gravimetry was used. From the experimental measurements, the effect of lithium bromide on the VLE behavior of water + 1,3-propanediol was investigated. The experimental data of the salt-free system were successfully correlated using the Wilson, NRTL, and UNIQUAC models. In addition, the extended UNIQUAC model of Sander et al. was applied to the VLE calculation of salt-containing mixtures.

  3. Size effects on the transport coefficient of liquid lithium, sodium and potassium using a soft sphere potential

    International Nuclear Information System (INIS)

    Adebayo, G.A.; Anusionwu, B.C.

    2004-08-01

    The dependence of the self diffusion coefficient of atoms in liquid Lithium, Sodium and Potassium, interacting through a soft sphere potential, on the number of atoms have been investigated using Molecular Dynamics Simulation at various temperatures. Our calculations predict non-linear relationship between the diffusion coefficient and the number of particles at high densities and medium or low temperatures. The radial distribution function obtained agrees well with experiment. (author)

  4. Development of liquid-lithium film jet-flow for the target of (7)Li(p,n)(7)Be reactions for BNCT.

    Science.gov (United States)

    Kobayashi, Tooru; Miura, Kuniaki; Hayashizaki, Noriyosu; Aritomi, Masanori

    2014-06-01

    A feasibility study on liquid lithium target in the form of a flowing film was performed to evaluate its potential use as a neutron generation target of (7)Li(p,n)(7)Be reaction in BNCT. The target is a windowless-type flowing film on a concave wall. Its configuration was adapted for a proton beam which is 30mm in diameter and with energy and current of up to 3MeV and 20mA, respectively. The flowing film of liquid lithium was 0.6mm in thickness, 50mm in width and 50mm in length. The shapes of the nozzle and concave back wall, which create a stable flowing film jet, were decided based on water experiments. A lithium hydrodynamic experiment was performed to observe the stability of liquid lithium flow behavior. The flowing film of liquid lithium was found to be feasible at temperatures below the liquid lithium boiling saturation of 342°C at the surface pressure of 1×10(-3)Pa. Using a proto-type liquid lithium-circulating loop for BNCT, the stability of the film flow was confirmed for velocities up to 30m/s at 220°C and 250°C in vacuum at a pressure lower than 10(-3) Pa. It is expected that for practical use, a flowing liquid lithium target of a windowless type can solve the problem of radiation damage and target cooling. Copyright © 2013 Elsevier Ltd. All rights reserved.

  5. CaO insulator coatings on a vanadium-base alloy in liquid 2 at.% calcium-lithium

    International Nuclear Information System (INIS)

    Park, J.H.; Kassner, T.F.

    1996-01-01

    The electrical resistance of CaO coatings produced on V-4%Cr-4%Ti and V-15%Cr-5%Ti by exposure of the alloy (round bottom samples 6-in. long by 0.25-in. dia.) to liquid lithium that contained 2 at.% dissolved calcium was measured as a function of time at temperatures between 300-464 degrees C. The solute element, calcium in liquid lithium, reacted with the alloy substrate at these temperatures for 17 h to produce a calcium coating ∼7-8 μm thick. The calcium-coated vanadium alloy was oxidized to form a CaO coating. Resistance of the coating layer on V-15Cr-5Ti, measured in-situ in liquid lithium that contained 2 at.% calcium, was 1.0 x 10 10 Ω-cm 2 at 300 degrees C and 400 h, and 0.9 x 10 10 Ω-cm 2 at 464 degrees C and 300 h. Thermal cycling between 300 and 464 degrees C changed the resistance of the coating layer, which followed insulator behavior. Examination of the specimen after cooling to room temperature revealed no cracks in the CaO coating. The coatings were evaluated by optical microscopy, scanning electron microscopy (SEM), electron dispersive spectroscopy (EDS), and X-ray analysis. Adhesion between CaO and vanadium alloys was enhanced as exposure time increased

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

    International Nuclear Information System (INIS)

    Natesan, K.

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

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

  8. 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

  9. 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.)

  10. Liquid-metal MHD flow in a duct whose cross section changes from a rectangle to a trapezoid, with applications in fusion blanket designs

    International Nuclear Information System (INIS)

    Walker, J.S.

    1986-04-01

    This paper treats the liquid-metal MHD flow in a semi-infinite rectangular duct and a semi-infinite trapezoidal duct, which are connected by a finite-length transition duct. There is a strong, transverse, uniform magnetic field. The walls parallel to the magnetic field (sides) remain parallel, while the walls intersecting the magnetic field are twisted in the transition duct to provide the change in cross sectional shape. The left side has a constant height, while the height of the right side increases or decreases in the transition duct. This geometry gives a skewed velocity profile with a high velocity near the left side, provided the right side is relatively thick. All walls are thin and electrically conducting, but the sides are considerably thicker than the other walls. The application is to fusion-reactor blankets in which a high velocity near the first wall (separating the plasma chamber from the coolant) improves the thermal performance. Junctions of different ducts with walls parallel to the magnetic field are treated for the first time. In expansions, contractions and other geometric transition ducts, as well as in straight ducts with axially varying magnetic fields, the fluid flow and electric currents are concentrated in boundary layers adjacent to the sides and in the side. At a junction with a straight duct with a uniform magnetic field, the flow and current must transfer from the boundary layers adn sides to the core regions. These transfers at junctions play a key role in any three-dimensional flow

  11. Liquid-metal MHD flow in a duct whose cross section changes from a rectangle to a trapezoid, with applications in fusion blanket designs

    Energy Technology Data Exchange (ETDEWEB)

    Walker, J.S.

    1986-04-01

    This paper treats the liquid-metal MHD flow in a semi-infinite rectangular duct and a semi-infinite trapezoidal duct, which are connected by a finite-length transition duct. There is a strong, transverse, uniform magnetic field. The walls parallel to the magnetic field (sides) remain parallel, while the walls intersecting the magnetic field are twisted in the transition duct to provide the change in cross sectional shape. The left side has a constant height, while the height of the right side increases or decreases in the transition duct. This geometry gives a skewed velocity profile with a high velocity near the left side, provided the right side is relatively thick. All walls are thin and electrically conducting, but the sides are considerably thicker than the other walls. The application is to fusion-reactor blankets in which a high velocity near the first wall (separating the plasma chamber from the coolant) improves the thermal performance. Junctions of different ducts with walls parallel to the magnetic field are treated for the first time. In expansions, contractions and other geometric transition ducts, as well as in straight ducts with axially varying magnetic fields, the fluid flow and electric currents are concentrated in boundary layers adjacent to the sides and in the side. At a junction with a straight duct with a uniform magnetic field, the flow and current must transfer from the boundary layers adn sides to the core regions. These transfers at junctions play a key role in any three-dimensional flow.

  12. Experiments on 18-8 stainless steels exposed to liquid lithium. I. 1,100-hour corrosion tests in lithium of 400, 500 and 6000C in natural circulation type testing apparatus

    International Nuclear Information System (INIS)

    Nihei, I.; Sumiya, I.; Fukaya, Y.; Yamazaki, Y.

    The Japan Atomic Energy Research Institute has planned and started to carry out a series of experiments concerning fusion reactor materials. This report gives the results of the first experiments. The first test materials selected were 18-8 stainless steels, and the experiments were designed to test their behavior when exposed to liquid lithium. Natural circulation type corrosion testing devices (pots) were used as the testing apparatus, and the tests were conducted with lithium temperatures up to 600 0 C

  13. Preliminary Evaluation of the Adequacy of Lithium Resources of the World and China for D-T Fusion Reactors

    Science.gov (United States)

    Wang, Yongliang; Ni, Muyi; Jiang, Jieqiong; Wu, Yican; FDS-Team

    2012-07-01

    This paper studied the adequacy of the World and China lithium resources, considering the most promising uses in the future, involving nuclear fusion and electric-vehicles. The lithium recycle model for D-T fusion power plant and electric-vehicles, and the logistic growth prediction model of the primary energy for the World and China were constructed. Based on these models, preliminary evaluation of lithium resources adequacy of the World and China for D-T fusion reactors was presented under certain assumptions. Results show that: a. The world terrestrial reserves of lithium seems too limited to support a significant D-T power program, but the lithium reserves of China are relatively abundant, compared with the world case. b. The lithium resources contained in the oceans can be called the “permanent" energy. c. The change in 6Li enrichment has no obvious effect on the availability period of the lithium resources using FDS-II (Liquid Pb-17Li breeder blanket) type of reactors, but it has a stronger effect when PPCS-B (Solid Li4 SiO4 ceramics breeder blanket) is used.

  14. Solid-state supercapacitors with ionic liquid based gel polymer electrolyte: Effect of lithium salt addition

    Science.gov (United States)

    Pandey, G. P.; Hashmi, S. A.

    2013-12-01

    Performance characteristics of the solid-state supercapacitors fabricated with ionic liquid (IL) incorporated gel polymer electrolyte and acid treated multiwalled carbon nanotube (MWCNT) electrodes have been studied. The effect of Li-salt (LiPF6) addition in the IL (1-ethyl-3-methylimidazolium tris(pentafluoroethyl) trifluorophosphate, EMImFAP) based gel electrolyte on the performance of supercapacitors has been specifically investigated. The LiPF6/IL/poly(vinylidine fluoride-co-hexafluoropropylene) (PVdF-HFP) gel electrolyte film possesses excellent electrochemical window of 4 V (from -2.0 to 2.0 V), high ionic conductivity ∼2.6 × 10-3 S cm-1 at 20 °C and high enough thermal stability. The comparative performance of supercapacitors employing electrolytes with and without lithium salt has been evaluated by impedance spectroscopy and cyclic voltammetric studies. The acid-treated MWCNT electrodes show specific capacitance of ∼127 F g-1 with IL/LiPF6 containing gel polymer electrolyte as compared to that with the gel polymer electrolyte without Li-salt, showing the value of ∼76 F g-1. The long cycling stability of the solid state supercapacitor based on the Li-salt containing gel polymer electrolyte confirms the electrochemical stability of the electrolyte.

  15. Review of JAEA activities on the IFMIF liquid lithium target in FY2006

    International Nuclear Information System (INIS)

    Ida, Mizuho; Nakamura, Hiroo; Miyashita, Makoto; Sugimoto, Masayoshi; Chida, Teruo; Furuya, Kazuyuki; Yoshida, Eiichi; Hirakawa, Yasuhi; Miyake, Osamu; Hirabayashi, Masaru; Ara, Kuniaki

    2008-03-01

    Engineering Validation Design and Engineering Design Activity (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) is under going. IFMIF is an accelerator-based Deuterium-Lithium (D-Li) neutron source to produce intense high energy neutrons and a sufficient irradiation volume for testing candidate materials for fusion reactors. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid Li flow with a speed of 20 m/s. In target system, nuclear heating due to neutron causes thermal stress especially on a back-wall of the target assembly. In addition, radioactive species such as beryllium-7, tritium and activated corrosion products are generated. In this report, thermal stress analyses of the back-wall, mechanical tests on weld specimen made of the back-wall material, estimations of beryllium-7 behavior and worker dose at the IFMIF Li loop and consideration on major EVEDA tasks are summarized. (author)

  16. Accessibility evaluation of the IFMIF liquid lithium loop considering activated erosion/corrosion materials deposition

    International Nuclear Information System (INIS)

    Nakamura, H.; Takemura, M.; Yamauchi, M.; Fischer, U.; Ida, M.; Mori, S.; Nishitani, T.; Simakov, S.; Sugimoto, M.

    2005-01-01

    This paper presents an evaluation of accessibility of the Li loop piping considering activated corrosion product. International Fusion Materials Irradiation Facility (IFMIF) is a deuteron-lithium (Li) stripping reaction neutron source for fusion materials testing. Target assembly and back wall are designed as fully remote maintenance component. Accessibility around the Li loop piping will depend on activation level of the deposition materials due to the back wall erosion/corrosion process under liquid Li flow. Activation level of the corrosion products coming from the AISI 316LN back wall is calculated by the ACT-4 of the THIDA-2 code system. The total activities after 1 day, 1 week, 1 month and 1 year cooling are 3.1 x 10 14 , 2.8 x 10 14 , 2.3 x 10 14 and 7.5 x 10 13 Bq/kg, respectively. Radiation dose rate around the Li loop pipe is calculated by QAD-CGGP2R code. Activated area of the back wall is 100 cm 2 . Corrosion rate is assumed 1 μm/year. When 10% of the corrosion material is supposed to be deposited on the inner surface of the pipe, the dose rate is calculated to be less than a permissible level of 10 μSv/h for hands-on maintenance, therefore, the maintenance work is assessed to be possible

  17. Acquisition of Co metal from spent lithium-ion battery using emulsion liquid membrane technology and emulsion stability test

    Science.gov (United States)

    Yuliusman; Wulandari, P. T.; Amiliana, R. A.; Huda, M.; Kusumadewi, F. A.

    2018-03-01

    Lithium-ion batteries are the most common type to be used as energy source in mobile phone. The amount of lithium-ion battery wastes is approximated by 200 – 500 ton/year. In one lithium-ion battery, there are 5 – 20% of cobalt metal, depend on the manufacturer. One of the way to recover a valuable metal from waste is leaching process then continued with extraction, which is the aim of this study. Spent lithium-ion batteries will be characterized with EDX and AAS, the result will show the amount of cobalt metal with form of LiCoO2 in the cathode. Hydrochloric acid concentration used is 4 M, temperature 80°C, and reaction time 1 hour. This study will discuss the emulsion stability test on emulsion liquid membrane. The purpose of emulsion stability test in this study was to determine optimum concentration of surfactant and extractant to produce a stable emulsion. Surfactant and extractant used were SPAN 80 and Cyanex 272 respectively with both concentrations varied. Membrane and feed phase ratios used in this experiment was 1 : 2. The optimum results of this study were SPAN 80 concentrations of 10% w/v and Cyanex 272 0.7 M.

  18. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  19. 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)

  20. Phase IIA and IIB experiments of JAERI/U.S.DOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1989-12-01

    Phase IIA and IIB experiments on fusion blanket neutronics has been performed on a basis of JAERI/USDOE collaborative program. In the Phase II experimental series, a D-T neutron source and a test blanket were contained by a lithium-carbonate enclosure to adjust the incident neutron spectrum to the test blanket so as to simulate that of a fusion reactor. First two series of the Phase II, IIA and IIB, focused especially on influences of beryllium configurations for neutron multiplying zone to neutronic parameters. Measured parameters were tritium production rate using Li-glass and NE213 scintillators, and Li-metal foil and Lithium-oxide block with liquid scintillation technique; neutron spectrum using NE213 scintillator and proton recoil proportional counter; reaction rate using foil activation technique. These parameters were compared among six different beryllium configurations of the experimental system. Consistency between different techniques for each measured parameter was also tested among different experimental systems and confirmed to be within experimental errors. This report describes, in detail, experimental conditions, assemblies, equipments and neutron source in Part I. The part II compiles all information required for a calculational analysis of this experiment, e.g., dimensions of the target room, target assembly, experimental assembly, their material densities and numerical data of experimental results. This compilation provides benchmark data to test calculation models and computing code systems used for a nuclear design of a fusion reactor. (author)

  1. 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

  2. First experiments with a liquid-lithium based high-intensity 25-keV neutron source

    International Nuclear Information System (INIS)

    Paul, M.

    2014-01-01

    A high-intensity neutron source based on a Liquid-Lithium Target (LiLiT) and the 7 Li(p,n) reaction was developed at SARAF (Soreq Applied Research Accelerator Facility, Israel) and is used for nuclear astrophysics experiments. The setup was commissioned with a 1.3 mA proton beam at 1.91 MeV, producing a neutron yield of ~ 2 ×10 10 n/s, more than one order of magnitude larger than conventional 7 Li(p,n)-based neutron sources and peaked at ~25 keV. The LiLiT device consists of a high-velocity (> 4 m/s) vertical jet of liquid lithium (~200 °C) whose free surface is bombarded by the proton beam. The lithium jet acts both as the neutron-producing target and as a power beam dump. The target dissipates a peak power areal density of 2.5 kW/cm 2 and peak volume density of 0.5 MW/cm 3 with no change of temperature or vacuum regime in the vacuum chamber. Preliminary results of Maxwellian-averaged cross section measurements for stable isotopes of Zr and Ce, performed by activation in the neutron flux of LiLiT, and nuclear-astrophysics experiments in planning will be described. (author)

  3. Development and implementation of flowing liquid lithium limiter control system for EAST

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, XiaoLin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230031 (China); Chen, Yue [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, JianSheng, E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Li, JianGang; Zuo, GuiZhong; Ren, Jun; Zhou, Yue; Li, ChangZheng; Sun, Zheng; Xu, Wei; Meng, XianCai; Huang, Ming; Zheng, XingWei; Yao, Xingjia [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • Development of a FLiLi remote control system for EAST. • Intelligent instruments are used to realize FLiLi remote control. • Good operating results of the control system were obtained in the EAST campaign. - Abstract: A control system of a flowing liquid lithium (FLiLi) limiter for the Experimental Advanced Superconducting Tokamak (EAST) was developed and implemented. The control system is not only able to control the direct current (DC) electromagnetic pump and heating power but can also set scanning parameters, receive the shot number, acquire the temperature, etc. The system consists of multifunctional LAN eXtensions for Instrumentation (LXI) instrument, temperature-acquisition module, programmable DC power supply, and programmable logic controller (PLC). The multi-range DC power supply is programmed to meet the operational requirements of the DC electromagnetic pump. The LXI instrument and temperature-acquisition module are used to obtain temperature data. The PLC is adopted to control the temperature of the FLiLi limiter. A safety interlock and protection function was developed for the FLiLi limiter control system. The software was designed by using LabVIEW to achieve data interaction between multiple protocols. The FLiLi limiter control system can acquire experimental data at a speed of 100 S/s and store it for later analysis. The control system was successfully applied to a FLiLi limiter to study the interaction between plasma and a fixed wall in the EAST campaign. This paper presents the framework, the implementation details, and results of the control system.

  4. Evaluation of thermal conductivity for liquid lead lithium alloys at various Li concentrations based on measurement and evaluation of density, thermal diffusivity and specific heat of alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Masatoshi, E-mail: kondo.masatoshi@nr.titech.ac.jp [Tokyo Institute of Technology, 2-12-1, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Nakajima, Yuu; Tsuji, Mitsuyo [Tokai University, 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa 259-1292 (Japan); Nozawa, Takashi [Japan Atomic Energy Agency, Rokkasyo-mura, Kamikita-gun, Aomori 039-3212 (Japan)

    2016-11-01

    Graphical abstract: Thermal diffusivities and thermal conductivities of liquid Pb–Li alloys (Pb–5Li, Pb–11Li and Pb–17Li). - Highlights: • The densities and specific heats of liquid Pb–Li alloys are evaluated based on the previous studies, and mathematically expressed in the equations with the functions of temperature and Li concentration. • The thermal diffusivities of liquid Pb–Li alloys (i.e., Pb–5Li, Pb–11Li and Pb–17Li) are obtained by laser flash method, and mathematically expressed in the equations with the functions of temperature and Li concentration. • The thermal conductivities of liquid Pb–Li alloys were evaluated and mathematically expressed in the equations with the functions of temperature and Li concentration. - Abstract: The thermophysical properties of lead lithium alloy (Pb–Li) are essential for the design of liquid Pb–Li blanket system. The purpose of the present study is to make clear the density, the thermal diffusivity and the heat conductivity of the alloys as functions of temperature and Li concentration. The densities of the solid alloys were measured by means of the Archimedean method. The densities of the alloys at 300 K as a function of Li concentration (0 at% < χ{sub Li} < 28 at%) were obtained in the equation as ρ{sub (300} {sub K)} [g/cm{sup 3}] = −6.02 × 10{sup −2} × χ{sub Li} + 11.3. The density of the liquid alloys was formulated as functions of temperature and Li concentration (0 at% < χ{sub Li} < 30 at%), and expressed in the equation as ρ [g/cm{sup 3}] = (9.00 × 10{sup −6} × T − 7.01 × 10{sup −2}) × χ{sub Li} + 11.4 − 1.19 × 10{sup −3}T. The thermal diffusivity of Pb, Pb–5Li, Pb–11Li and Pb–17Li were measured by means of laser flash method. The thermal diffusivity of Pb–17Li was obtained in the equation as α{sub Pb–17Li} [cm{sup 2}/s] = 3.46 × 10{sup −4}T + 1.05 × 10{sup −1} for the temperature range between 573 K and 773 K. The thermal conductivity of

  5. Current Status on the Korean Test Blanket Module Development for testing in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Jung, Ki Sok

    2010-01-01

    Korea has proposed and designed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER). Ferrite Martensitic (FM) steel is used as the structural material and helium (He) is used as a coolant to cool the first wall (FW) and breeding zone. Liquid lithium (Li) is circulated for a tritium breeding, not for a cooling purpose. Main purpose for developing the TBM is to develop the design technology for DEMO and fusion reactor and it should be proved through the experiment in the ITER with TBM. Therefore, we have developed the design scheme and related codes including the safety analysis for obtain the license to be tested in the ITER. In order to develop and install at the ITER, several technologies were developed in parallel; fabrication, breeder, He cooling, tritium extraction and so on. Figure 1 shows the overall TBM development scheme. In Korea, official strategy for developing the TBM is to participate to other parties' concept such as US and EU ones, in which PbLi (lead lithium eutectic), He, and FM steel were used for liquid breeder, coolant, and structural material, respectively

  6. Liquid-phase plasma synthesis of silicon quantum dots embedded in carbon matrix for lithium battery anodes

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Ying [Institute of Functional Nano and Soft Materials (FUNSOM) and Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou (China); College of Chemistry and Chemical Engineering, Bohai University, Jinzhou 121000 (China); Yu, Hang; Li, Haitao; Ming, Hai; Pan, Keming; Huang, Hui [Institute of Functional Nano and Soft Materials (FUNSOM) and Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou (China); Liu, Yang, E-mail: yangl@suda.edu.cn [Institute of Functional Nano and Soft Materials (FUNSOM) and Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou (China); Kang, Zhenhui, E-mail: zhkang@suda.edu.cn [Institute of Functional Nano and Soft Materials (FUNSOM) and Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou (China)

    2013-10-15

    Graphical abstract: - Highlights: • Silicon quantum dots embedded in carbon matrix (SiQDs/C) were fabricated. • SiQDs/C exhibits excellent battery performance as anode materials with high specific capacity. • The good performance was attributed to the marriage of small sized SiQDs and carbon. - Abstract: Silicon quantum dots embedded in carbon matrix (SiQDs/C) nanocomposites were prepared by a novel liquid-phase plasma assisted synthetic process. The SiQDs/C nanocomposites were demonstrated to show high specific capacity, good cycling life and high coulmbic efficiency as anode materials for lithium-ion battery.

  7. Liquid-phase plasma synthesis of silicon quantum dots embedded in carbon matrix for lithium battery anodes

    International Nuclear Information System (INIS)

    Wei, Ying; Yu, Hang; Li, Haitao; Ming, Hai; Pan, Keming; Huang, Hui; Liu, Yang; Kang, Zhenhui

    2013-01-01

    Graphical abstract: - Highlights: • Silicon quantum dots embedded in carbon matrix (SiQDs/C) were fabricated. • SiQDs/C exhibits excellent battery performance as anode materials with high specific capacity. • The good performance was attributed to the marriage of small sized SiQDs and carbon. - Abstract: Silicon quantum dots embedded in carbon matrix (SiQDs/C) nanocomposites were prepared by a novel liquid-phase plasma assisted synthetic process. The SiQDs/C nanocomposites were demonstrated to show high specific capacity, good cycling life and high coulmbic efficiency as anode materials for lithium-ion battery

  8. A 2D Finite Element Modelling of Tritium Permeation Through Cooling Plates for The HCLL DEMO Blanket Module

    International Nuclear Information System (INIS)

    Gabriel, F.; Escuriol, Y.; Dabbene, F.; Salavy, J.F.; Giancarli, L.; Gastaldi, O.

    2006-01-01

    As the Tritium self sufficiency is one of the major challenges for fusion reactor, breeding blankets represent one of the major technological breakthroughs required from passing from ITER to the next step reactor, usually called DEMO. One of the two blanket concepts developed in the EU is the Helium Cooled Lithium Lead (HCLL) blanket which uses the eutectic Pb-15.7Li metal liquid as both breeder and neutron multiplier. The structures, made of EUROFER, a low activation ferritic martensitic steel, are cooled by pressurized helium at 8 MPa and inlet/outlet temperature 300/500 o C. In this concept, the LiPb is fed from the top of the blanket and distributed in parallel vertical channels among pairs of cells (one cell for the radial movement towards the plasma, the other for the return). The liquid metal fills the in-box volume and is slowly re-circulated (few mm per seconds) to remove the produced tritium. In this paper, a local finite element modelling of the tritium permeation rate through the HCLL breeder unit cooling plates is presented. The tritium concentration in the helium circuit and remaining in the lithium lead circuit are evaluated by solving partial differential equations governing the tritium concentration balance, the thermal field and the lithium lead velocity field for a simplified 2D geometrical representation of the breeder unit. This allows estimating the sensitivity effect of coupling these different equations in order to deduce a relevant but simplified modelling for tritium permeation. This is to compare with tritium inventories studies, were the tritium permeation rate is estimated using simplified analytical modelling which generally leads to over estimate the tritium permeation rate to the coolant and so has strong influence on the coolant purification plant design. The finite element modelling performed shows that the Tritium permeation is considerable lower than the one obtained in previous estimations where nominal values of the governing

  9. 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

  10. 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

  11. 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.

  12. The effect of lead concentration on the corrosion susceptibility of 2 1/4Cr-1Mo steel in a lead-lithium liquid

    International Nuclear Information System (INIS)

    Wilkinson, B.D.; Edwards, G.R.; Hoffman, N.J.

    1982-01-01

    The intergranular penetration of 21/4Cr-1Mo steel by lead-lithium liquids containing 0, 17.6, and 53 w/o lead has been investigated at temperatures from 300 0 C to 600 0 C for times up to 1000 hours. Limited tests using a 99.3 w/o lead-lithium liquid were also conducted. Tempering was found to remove the susceptibility of as-quenched 21/4Cr-1Mo steel to penetration at 500 0 C by lead-lithium liquids containing up to 53 w/o lead. Penetration by the 99.3 w/o lead-lithium liquid in 1000 hours at 500 0 C was found to be negligible even when the steel was in the as-quenched condition. An Arrhenius analysis yielded the same low initial activation energy (approx. equal to25 kJ/mole) for liquids containing 0, 17.6, and 53 w/o lead. The initial penetration rate for lead-free lithium was significantly greater than that for the lead-bearing liquids, a factor thought to be related to the effect of lead on the wettability of the liquid. The same secondary activation energy (approx. equal to120 kJ/mole) was also found for the three liquids. Furthermore, the secondary penetration rate was found to be insensitive to lead content. Anomalous behavior at 500 0 C, observed in this study as well as in previous studies, is discussed, and a hypothetical explanation for the behavior is presented. (orig.)

  13. A green strategy for lithium isotopes separation by using mesoporous silica materials doped with ionic liquids and benzo-15-crown-5

    International Nuclear Information System (INIS)

    Wen Zhou; Xiao-Li Sun; Lin Gu; Fei-Fei Bao; Xin-Xin Xu; Chun-Yan Pang; Zaijun Li; Zhi-Guo Gu; Jiangnan University, Wuxi

    2014-01-01

    Three new mesoporous silica materials IL15SGs (HF15SG, TF15SG and DF15SG) doped with benzo-15-crown-5 and imidazolium based ionic liquids (C 8 mim + PF 6 - , C 8 mim + BF 4 - or C 8 mim + NTf 2 - ) have been prepared by a simple approach to separating lithium isotopes. The formed mesoporous structures of silica gels have been confirmed by transmission electron microscopy image and N 2 gas adsorption-desorption isotherm. Imidazolium ionic liquids acted as templates to prepare mesoporous materials, additives to stabilize extractant within silica gel, and synergetic agents to separate the lithium isotopes. Factors such as lithium salt concentration, initial pH, counter anion of lithium salt, extraction time, and temperature on the lithium isotopes separation were examined. Under optimized conditions, the extraction efficiency of HF15SG, TF15SG and DF15SG were found to be 11.43, 10.59 and 13.07 %, respectively. The heavier isotope 7 Li was concentrated in the solution phase while the lighter isotope 6 Li was enriched in the gel phase. The solid-liquid extraction maximum single-stage isotopes separation factor of 6 Li- 7 Li in the solid-liquid extraction was up to 1.046 ± 0.002. X-ray crystal structure analysis indicated that the lithium salt was extracted into the solid phase with crown ether forming [(Li 0.5 ) 2 (B 15 ) 2 (H 2 O)] + complexes. IL15SGs were also easily regenerated by stripping with 20 mmol L -1 HCl and reused in the consecutive removal of lithium ion in five cycles. (author)

  14. Design development and manufacturing sequence of the European water-cooled Pb-17Li test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Futterer, M.A.; Bielak, B.; Deffain, J.P.; Giancarli, L.; Li Puma, A.; Salavy, J.F.; Szczepanski, J. [CEA Saclay, Gif-sur-Yvette (France). FDRN/DMT/SERMA; Dellis, C. [CEA Grenoble, DTA-CEREM/SGM, Grenoble (France); Nardi, C. [ENEA Frascati, ERG-FUS-TECN-MEC, Frascati (Italy); Schleisiek, K. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit

    1998-09-01

    In 1996, the European Community started the development of a water-cooled Pb17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the basic performance phase prior to D-T operation. The test module is designed to be a representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined. (orig.) 11 refs.

  15. Measurements of time-dependent liquid-metal magnetohydrodynamic flows in a flat rectangular duct

    International Nuclear Information System (INIS)

    Buehler, L.; Horanyi, S.

    2009-01-01

    In the helium-cooled lead lithium (HCLL) blanket, which has been chosen as a reference concept for a liquid-metal breeding blanket to be tested in ITER, the heat is removed by helium cooled plates aligned with the strong toroidal magnetic field that confines the fusion plasma. The liquid breeder lead lithium circulates through gaps of rectangular cross-section between the cooling plates to transport the generated tritium towards external extraction facilities. Under the action of the strong magnetic field, liquid metal flows in conducting rectangular ducts exhibit jet-like velocity profiles in the thin boundary layers near the side walls, which are parallel to the magnetic field like the cooling plates in HCLL blankets. The velocity in these side layers may exceed several times the mean velocity in the duct and it is known that these layers become unstable for sufficiently high Reynolds numbers. The present paper summarizes experimental results for such unstable time-dependent flows in strong magnetic fields, which have been obtained in the MEKKA liquid metal laboratory of the Forschungszentrum Karlsruhe. In particular, spatial and temporal scales of perturbation patterns are identified. The results suggest that the flow between cooling plates in a HCLL blanket is laminar and stable. The observed time-dependent flow behavior appears at larger velocities so that the present results are more relevant for applications in dual coolant concepts where high-velocity jets have been predicted along side walls.

  16. Upgrading the data acquisition and control systems of the European Breeding Blanket Test Facility

    International Nuclear Information System (INIS)

    Mannori, Simone; Sermenghi, Valerio; Utili, Marco; Malavasi, Andrea; Gianotti, Daniel

    2013-01-01

    Highlights: • Data Acquisition and Control Systems (DACS) upgrading of experimental plant for full size thermo hydraulic testing of nuclear subsystems. • DACS development using integrated hardware/software platform with graphical programming (LabVIEW). • Development of simplified models for real-time simulation. • Rapid prototyping with real time simulation of the complete plant. • Using the code developed for the real time simulator for the real plant DACS. -- Abstract: The EBBTF (European Breeding Blanket Test Facility) experimental plant is a key component for the development of the breeding blankets (TBMs test blanket modules, HCLL helium cooled lithium lead and HCPB helium cooled pebble bed types) used by ITER. EBBTF is an experimental plant which provides the double breeding/cooling loops (liquid metal and gas) required for HCLL testing. EBBTF is composed of four subsystems (TBM, IELLLO integrated European lead lithium loop, HE-FUS3 helium fusion loop, version 3 and helium compressor build by ATEKO) with dedicated control systems realized with hardware/software combinations covering 15 years (1995–2010) time span. At the end of 2010 we began to upgrade the HE-FUS3 data acquisition control systems (DACS) replacing the obsolete PLC Siemens S5 with National Instruments Compact FieldPoint and LabVIEW. The control room has been completely reorganized using high resolution monitors and workstations linked with standard Ethernet interfaces. The data acquisition, control, safety and SCADA software has been completely developed in ENEA using LabVIEW. In this paper we are going to discuss the technical difficulties and the solutions that we have used to accomplish the upgrade

  17. Thermally-responsive, nonflammable phosphonium ionic liquid electrolytes for lithium metal batteries: operating at 100 degrees celsius.

    Science.gov (United States)

    Lin, X; Kavian, R; Lu, Y; Hu, Q; Shao-Horn, Y; Grinstaff, M W

    2015-11-13

    Rechargeable batteries such as Li ion/Li metal batteries are widely used in the electronics market but the chemical instability of the electrolyte limits their use in more demanding environmental conditions such as in automotive, oil exploration, or mining applications. In this study, a series of alkyl phosphonium ionic liquid electrolyte are described with high thermal stability and solubility for LiTFSI. A lithium metal battery (LMB) containing a tailored phosphonium ionic liquid/LiTFSI electrolyte operates at 100 °C with good specific capacities and cycling stability. Substantial capacity is maintained during 70 cycles or 30 days. Instant on-off battery operation is realized via the significant temperature dependence of the electrolyte material, demonstrating the robustness and potential for use at high temperature.

  18. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. 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.)

  20. Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

    International Nuclear Information System (INIS)

    Swami, H.L.; Danani, C.

    2015-01-01

    In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

  1. Preliminary assessment of interactions between the FMIT deuteron beam and liquid-lithium target

    International Nuclear Information System (INIS)

    Hassberger, J.A.

    1983-03-01

    Scoping calculations were performed to assess the limit of response of the FMIT lithium target to the deuteron-beam interactions. Results indicate that most response modes have acceptably minor impacts on the lithium-target behavior. Individual modes of response were studied separately to assess sensitivity of the target to various phenomena and to identify those needing detailed evaluation. A few responses are of sufficient magnitude to warrant further investigation. Potential for several different responses combining additively is identified as the major area requiring further consideration

  2. 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

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

    International Nuclear Information System (INIS)

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

    1986-06-01

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

  4. 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.).

  5. 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.)

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  8. 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

  9. Lithium-modulated conduction band edge shifts and charge-transfer dynamics in dye-sensitized solar cells based on a dicyanamide ionic liquid.

    Science.gov (United States)

    Bai, Yu; Zhang, Jing; Wang, Yinghui; Zhang, Min; Wang, Peng

    2011-04-19

    Lithium ions are known for their potent function in modulating the energy alignment at the oxide semiconductor/dye/electrolyte interface in dye-sensitized solar cells (DSCs), offering the opportunity to control the associated multichannel charge-transfer dynamics. Herein, by optimizing the lithium iodide content in 1-ethyl-3-methylimidazolium dicyanamide-based ionic liquid electrolytes, we present a solvent-free DSC displaying an impressive 8.4% efficiency at 100 mW cm(-2) AM1.5G conditions. We further scrutinize the origins of evident impacts of lithium ions upon current density-voltage characteristics as well as photocurrent action spectra of DSCs based thereon. It is found that, along with a gradual increase of the lithium content in ionic liquid electrolytes, a consecutive diminishment of the open-circuit photovoltage arises, primarily owing to a noticeable downward movement of the titania conduction band edge. The conduction band edge displacement away from vacuum also assists the formation of a more favorable energy offset at the titania/dye interface, and thereby leads to a faster electron injection rate and a higher exciton dissociation yield as implied by transient emission measurements. We also notice that the adverse influence of the titania conduction band edge downward shift arising from lithium addition upon photovoltage is partly compensated by a concomitant suppression of the triiodide involving interfacial charge recombination. © 2011 American Chemical Society

  10. Applications of the aqueous self-cooled blanket (ASCB) concept to the Next European Torus (NET)

    International Nuclear Information System (INIS)

    Embrechts, M.J.; Bogaerts, W.; Cardella, A.; Chazalon, M.; Danner, W.; Dinner, P.; Libin, B.

    1987-01-01

    The Aqueous Self-Cooled Blanket Concept (ASCB) leads to a low-technology blanket design that relies on just structural material and coolant with small amounts of lithium compound dissolved in the coolant to provide for tritium production. The application of the ASCB concept in NET is being considered as a driver blanket that would operate at low temperature and low pressure and provide a reliable environment for machine operation during the technology phase. Shielding and tritium production are the primary objectives for such a low-technology blanket. Net tritium breeding is not a design requirement per se for a driver blanket for NET. A DEMO relevant ASCB based blanket test module with (local) tritium self-sufficiency and energy recovery as primary objectives might also be tested in NET if future developments confirm their viability

  11. Comparative studies of H absorption/desorption kinetics and evaporation of liquid lithium in different porous systems and free surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Oyarzabal, E., E-mail: eider.oyarzabal@externos.ciemat.es [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain); Calle de Guzmán el Bueno, 133, 28003 Madrid (Spain); Martín-Rojo, A.B. [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain); Calle de Guzmán el Bueno, 133, 28003 Madrid (Spain); Tabarés, F.L. [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain)

    2017-04-15

    In the present work, a study of the two most relevant properties of liquid lithium with respect to its suitability as a Plasma Facing Component (PFC) element in a Reactor, namely, its evaporation rate and the uptake/release of hydrogen, eventually leading to the formation of a stable hydride was carried out for Li in different porous systems and Li as a free surface. These properties were characterized in a temperature range of 200–500 °C. The H{sub 2} absorption kinetics at low pressure (<1torr) were measured for the different studied porous systems and then outgassed. Particle balance and chemical analysis were used to assess the retention properties of lithium for each case. Thermal Desorption Spectroscopy (TDS) analysis was used for the assessment of possible hydride formation. Evaporation rates were determined by using a Quartz Microbalance (QMB). A significant reduction of the evaporation rate was observed when Li was trapped in a microstructure of sintered stainless steel with a characteristic porous size of 5–10 μm. On the other hand, a negligible rate of H{sub 2} uptake was found at temperatures above 500 °C in all cases.

  12. The design status of the liquid lithium target facility of IFMIF at the end of the engineering design activities

    Energy Technology Data Exchange (ETDEWEB)

    Nitti, F.S., E-mail: francesco.nitti@enea.it [IFMIF/EVEDA Project Team, Rokkasho Japan (Japan); Ibarra, A. [CIEMAT, Madrid (Spain); Ida, M. [IHI Corporation, Tokyo (Japan); Favuzza, P. [ENEA Research Center Firenze (Italy); Furukawa, T. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Groeschel, F. [KIT Research Center, Karlsruhe (Germany); Heidinger, R. [F4E Research Center, Garching (Germany); Kanemura, T. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Knaster, J. [IFMIF/EVEDA Project Team, Rokkasho Japan (Japan); Kondo, H. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Micchiche, G. [ENEA Research Center, Brasimone (Italy); Sugimoto, M. [JAEA Research Center, Rokkasho Japan (Japan); Wakai, E. [JAEA Research Center, Tokai-mura, Ibaraki (Japan)

    2015-11-15

    Highlights: • Results of validation and design activity for the Li loop facility of IFMIF. • Demonstration of Li target stability, with surface disturbance <1 mm. • Demonstration of start-up and shut down procedures of Li loop. • Complete design of the heat removal system and C and O purification system. • Conceptual design of N and H isotopes purification systems. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is an experimental facility conceived for qualifying and characterizing structural materials for nuclear fusion applications. The Engineering Validation and Engineering Design Activity (EVEDA) is a fundamental step towards the final design. It presented two mandates: the Engineering Validation Activities (EVA), still on-going, and the Engineering Design Activities (EDA) accomplished on schedule in June 2013. Five main facilities are identified in IFMIF, among which the Lithium Target Facility constituted a technological challenge overcome thanks to the success of the main validation challenges impacting the design. The design of the liquid Lithium Target Facility at the end of the EDA phase is here detailed.

  13. The design status of the liquid lithium target facility of IFMIF at the end of the engineering design activities

    International Nuclear Information System (INIS)

    Nitti, F.S.; Ibarra, A.; Ida, M.; Favuzza, P.; Furukawa, T.; Groeschel, F.; Heidinger, R.; Kanemura, T.; Knaster, J.; Kondo, H.; Micchiche, G.; Sugimoto, M.; Wakai, E.

    2015-01-01

    Highlights: • Results of validation and design activity for the Li loop facility of IFMIF. • Demonstration of Li target stability, with surface disturbance <1 mm. • Demonstration of start-up and shut down procedures of Li loop. • Complete design of the heat removal system and C and O purification system. • Conceptual design of N and H isotopes purification systems. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is an experimental facility conceived for qualifying and characterizing structural materials for nuclear fusion applications. The Engineering Validation and Engineering Design Activity (EVEDA) is a fundamental step towards the final design. It presented two mandates: the Engineering Validation Activities (EVA), still on-going, and the Engineering Design Activities (EDA) accomplished on schedule in June 2013. Five main facilities are identified in IFMIF, among which the Lithium Target Facility constituted a technological challenge overcome thanks to the success of the main validation challenges impacting the design. The design of the liquid Lithium Target Facility at the end of the EDA phase is here detailed.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Porfiri, T.

    1996-06-01

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

  17. Note: Proton irradiation at kilowatt-power and neutron production from a free-surface liquid-lithium target

    Energy Technology Data Exchange (ETDEWEB)

    Halfon, S.; Feinberg, G. [Soreq NRC, Yavne 81800 (Israel); Racah Institute of Physics, Hebrew University, Jerusalem 91904 (Israel); Arenshtam, A.; Kijel, D.; Weissman, L.; Aviv, O.; Berkovits, D.; Dudovitch, O.; Eisen, Y.; Eliyahu, I.; Haquin, G.; Hazenshprung, N.; Kreisel, A.; Mardor, I.; Shimel, G.; Shor, A.; Silverman, I.; Yungrais, Z. [Soreq NRC, Yavne 81800 (Israel); Paul, M., E-mail: paul@vms.huji.ac.il; Tessler, M. [Racah Institute of Physics, Hebrew University, Jerusalem 91904 (Israel)

    2014-05-15

    The free-surface Liquid-Lithium Target, recently developed at Soreq Applied Research Accelerator Facility (SARAF), was successfully used with a 1.9 MeV, 1.2 mA (2.3 kW) continuous-wave proton beam. Neutrons (∼2 × 10{sup 10} n/s having a peak energy of ∼27 keV) from the {sup 7}Li(p,n){sup 7}Be reaction were detected with a fission-chamber detector and by gold activation targets positioned in the forward direction. The setup is being used for nuclear astrophysics experiments to study neutron-induced reactions at stellar energies and to demonstrate the feasibility of accelerator-based boron neutron capture therapy.

  18. Molecular dynamics study of thermodynamic stability and dynamics of [Li(glyme)]+ complex in lithium-glyme solvate ionic liquids

    Science.gov (United States)

    Shinoda, Wataru; Hatanaka, Yuta; Hirakawa, Masashi; Okazaki, Susumu; Tsuzuki, Seiji; Ueno, Kazuhide; Watanabe, Masayoshi

    2018-05-01

    Equimolar mixtures of glymes and organic lithium salts are known to produce solvate ionic liquids, in which the stability of the [Li(glyme)]+ complex plays an important role in determining the ionic dynamics. Since these mixtures have attractive physicochemical properties for application as electrolytes, it is important to understand the dependence of the stability of the [Li(glyme)]+ complex on the ion dynamics. A series of microsecond molecular dynamics simulations has been conducted to investigate the dynamic properties of these solvate ionic liquids. Successful solvate ionic liquids with high stability of the [Li(glyme)]+ complex have been shown to have enhanced ion dynamics. Li-glyme pair exchange rarely occurs: its characteristic time is longer than that of ion diffusion by one or two orders of magnitude. Li-glyme pair exchange most likely occurs through cluster formation involving multiple [Li(glyme)]+ pairs. In this process, multiple exchanges likely take place in a concerted manner without the production of energetically unfavorable free glyme or free Li+ ions.

  19. Effect of liquid oil additive on lithium-ion battery ceramic composite separator prepared with an aqueous coating solution

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Woo [Division of Advanced Materials Engineering, Kongju National University, 1223–24, Cheonan-daero, Cheonan, Chungnam, 31080 (Korea, Republic of); Ryou, Myung-Hyun [Department of Chemical & Biological Engineering, Hanbat National University, 125, Dongseodaero, Yuseong-gu, Daejeon, 34158 (Korea, Republic of); Lee, Yong Min, E-mail: yongmin.lee@hanbat.ac.kr [Department of Chemical & Biological Engineering, Hanbat National University, 125, Dongseodaero, Yuseong-gu, Daejeon, 34158 (Korea, Republic of); Cho, Kuk Young, E-mail: kycho@hanyang.ac.kr [Department of Materials Science and Chemical Engineering, Hanyang University, 55, Hanyangdaehak-ro, Sangrok-gu, Ansan, Gyeonggi-do, 15588 (Korea, Republic of)

    2016-08-05

    Ceramic composite separators (CCSs) play a critical role in ensuring safety for lithium-ion batteries (LIBs), especially for mid- and large-sized devices. However, production of CCSs using organic solvents has some cost and environmental concerns. An aqueous process for fabricating CCSs is attractive because of its cost-effectiveness and environmental-friendliness because organic solvents are not used. The success of an aqueous coating system for LIBs is dependent upon minimizing moisture content, as moisture has a negatively impact on LIB performance. In this study, CCSs were fabricated using an aqueous coating solution containing Al{sub 2}O{sub 3} and an acrylic binder. Compared with polyethylene (PE) separators, CCSs coated with an aqueous coating solution showed improved thermal stability, electrolyte uptake, puncture strength, ionic conductivity, and rate capability. In addition, our new approach of introducing a small amount of an oily liquid to the aqueous coating solution reduced the water adsorption by 11.7% compared with coatings that do not contain the oily liquid additive. - Highlights: • Ceramic composite separator is fabricated using aqueous coating process. • Coated separator showed enhanced mechanical and thermal stability. • Liquid oil additive in coating solution reduce moisture reabsorption of separator. • Oil additive in aqueous coating solution does not deteriorate LIB performance.

  20. Pyrrolidinium-based ionic liquid electrolyte with organic additive and LiTFSI for high-safety lithium-ion batteries

    International Nuclear Information System (INIS)

    Yang, Binbin; Li, Cuihua; Zhou, Junhui; Liu, Jianhong; Zhang, Qianling

    2014-01-01

    Highlights: • New ionic liquid electrolytes composed by PYR 13 TFSI and EC/DMC-5%VC. • Mixed electrolyte for use in high-safety lithium-ion batteries. • LiTFSI concentration in IL electrolyte greatly affects the rate capability of the cell. • The optimal mixed electrolyte is ideal for applications at high temperature. - Abstract: In this paper, we report on the physicochemical properties of mixed electrolytes based on an ionic liquid N-propyl-N-methylpyrrolidiniumbis (trifluoromethanesulfonyl) imide (PYR 13 TFSI), organic additives, and lithium bis (trifluoromethanesulfonyl) imide (LiTFSI) for high safety lithium-ion batteries. The proposed optimal content of ionic liquid in the mixed electrolyte is 65 vol%, which results in non- flammability, high thermal stability, a wide electrochemical window of 4.8 V, low viscosity, low bulk resistance and the lowest interface resistance to lithium anode. The effects of the concentration of LiTFSI in the above electrolyte are critical to the rate performance of the LiFePO 4 -based battery. We have found the suitable LiTFSI concentration (0.3 M) for good capacity retention and rate capability

  1. High power accelerator-based boron neutron capture with a liquid lithium target and new applications to treatment of infectious diseases

    Energy Technology Data Exchange (ETDEWEB)

    Halfon, S. [Soreq NRC, Yavne 81800 (Israel); Racah Institute of Physics, Hebrew University, Jerusalem 91904 (Israel)], E-mail: halfon@phys.huji.ac.il; Paul, M. [Racah Institute of Physics, Hebrew University, Jerusalem 91904 (Israel); Steinberg, D. [Biofilm Laboratory, Institute of Dental Sciences, Faculty of Dentistry, Hebrew University-Hadassah (Israel); Nagler, A.; Arenshtam, A.; Kijel, D. [Soreq NRC, Yavne 81800 (Israel); Polacheck, I. [Clinical Microbiology and Infectious Diseases, Hadassah-Hebrew University Medical Center (Israel); Srebnik, M. [Department of Medicinal Chemistry and Natural Products, School of Pharmacy, Hebrew University, Jerusalem 91120 (Israel)

    2009-07-15

    A new conceptual design for an accelerator-based boron neutron capture therapy (ABNCT) facility based on the high-current low-energy proton beam driven by the linear accelerator at SARAF (Soreq Applied Research Accelerator Facility) incident on a windowless forced-flow liquid-lithium target, is described. The liquid-lithium target, currently in construction at Soreq NRC, will produce a neutron field suitable for the BNCT treatment of deep-seated tumor tissues, through the reaction {sup 7}Li(p,n){sup 7}Be. The liquid-lithium target is designed to overcome the major problem of solid lithium targets, namely to sustain and dissipate the power deposited by the high-intensity proton beam. Together with diseases conventionally targeted by BNCT, we propose to study the application of our setup to a novel approach in treatment of diseases associated with bacterial infections and biofilms, e.g. inflammations on implants and prosthetic devices, cystic fibrosis, infectious kidney stones. Feasibility experiments evaluating the boron neutron capture effectiveness on bacteria annihilation are taking place at the Soreq nuclear reactor.

  2. Progress in blanket designs using SiCf/SiC composites

    International Nuclear Information System (INIS)

    Giancarli, L.; Golfier, H.; Nishio, S.; Raffray, R.; Wong, C.; Yamada, R.

    2002-01-01

    This paper summarizes the most recent design activities concerning the use of SiC f /SiC composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the TAURO blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium-lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and ARIES-AT blankets are essentially formed by a SiC f /SiC box acting as a container for the lithium-lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. The DREAM blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li 2 O (or other lithium ceramics) as breeder material and of SiC as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R and D on SiC f /SiC

  3. Approach to lithium burn-up effect in lithium ceramics

    International Nuclear Information System (INIS)

    Rasneur, B.

    1994-01-01

    The lithium burn-up in Li 2 ZrO 3 is simulated by removing lithium under Li 2 O form and trapping it in high specific surface area powder while heating during 15 days or 1 month at moderate temperature so that lithium mobility be large enough without causing any sintering neither of the specimens nor of the powder. In a first treatment at 775 deg C during 1 month. 30% of the lithium content could be removed inducing a lithium concentration gradient in the specimen and the formation of a lithium-free monoclinic ZrO 2 skin. Improvements led to similar results at 650 deg C and 600 deg C, the latter temperatures are closer to the operating temperature of the ceramic breeder blanket of a fusion reactor. (author) 4 refs.; 4 figs.; 1 tab

  4. Lithium polymer cell assembled by in situ chemical cross-linking of ionic liquid electrolyte with phosphazene-based cross-linking agent

    International Nuclear Information System (INIS)

    Choi, Ji-Ae; Kang, Yongku; Kim, Dong-Won

    2013-01-01

    Highlights: ► Ionic liquid-based cross-linked gel polymer electrolytes were synthesized and their electrochemical properties were investigated. ► Lithium polymer cells with in situ cross-linked gel polymer electrolytes exhibited reversible cycling behavior with good capacity retention. ► The use of ionic liquid-based cross-linked gel polymer electrolytes significantly improved the thermal stability of the cells. -- Abstract: Ionic liquid-based cross-linked gel polymer electrolytes were prepared with a phosphazene-based cross-linking agent, and their electrochemical properties were investigated. Lithium polymer cells composed of lithium anode and LiCoO 2 cathode were assembled with ionic liquid-based cross-linked gel polymer electrolyte and their cycling performance was evaluated. The interfacial adhesion between the electrodes and the electrolyte by in situ chemical cross-linking resulted in stable capacity retention of the cell. A reduction in the ionic mobility in both the electrolyte and the electrode adversely affected discharge capacity and high rate performance of the cell. DSC studies demonstrated that the use of ionic liquid-based cross-linked gel polymer electrolytes provided a significant improvement in the thermal stability of the cell

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

    International Nuclear Information System (INIS)

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

    1987-11-01

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

  6. 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.

  7. New approaches to the design of polymer and liquid electrolytes for lithium batteries

    Science.gov (United States)

    McBreen, J.; Lee, H. S.; Yang, X. Q.; Sun, X.

    All non-aqueous lithium battery electrolytes are Lewis bases that interact with cations. Unlike water, they do not interact with anions. The result is a high degree of ion pairing and the formation of triplets and higher aggregates. This decreases the conductivity and the lithium ion transference, and results in polarization losses in batteries. Approaches that have been used to increase ion dissociation in poly(ethylene oxide) (PEO)-based electrolytes are the use of salts with low lattice energy, the addition of polar plasticizers to the polymer, and the addition of cation complexing agents such as crown ethers or cryptands. Complexing of the anions is a more promising approach, since it should increase both ion dissociation and the lithium transference. At Brookhaven National Laboratory (BNL) we have synthesized two new families of neutral anion complexing agents, each based on Lewis acid centers. One is based on electron deficient nitrogen sites on substituted aza-ethers, wherein the hydrogen on the nitrogen is replaced by electron withdrawing groups such as CF 3SO 3-. The other is based on electron deficient boron sites on borane or borate compounds with various fluorinated aryl or alkyl groups. Some of the borane-based anion receptors can promote the dissolution of LiF in several solvents. Several of these compounds, when added in equivalent amounts, produce 1.2 M LiF solutions in DME, an increase in solubility of LiF by six orders of magnitude. Some of these LiF electrolytes have conductivities as high as 6×10 -3 S cm -1. The LiF electrolytes with borane anion acceptors in PC:EC:DEC solvents have excellent electrochemical stability. This has been demonstrated in small Li/LiMn 2O 4 cells.

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

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1977-09-01

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

  9. A QuaternaryPoly(ethylene carbonate)-Lithium Bis(trifluoromethanesulfonyl)imide-Ionic Liquid-Silica Fiber Composite Polymer Electrolyte for Lithium Batteries

    International Nuclear Information System (INIS)

    Kimura, Kento; Matsumoto, Hidetoshi; Hassoun, Jusef; Panero, Stefania; Scrosati, Bruno; Tominaga, Yoichi

    2015-01-01

    Highlights: • A quaternary PEC-LiTFSI-Pyr 14 TFSI-Silica fiber electrolyte was prepared by a solvent casting method. • Both electrochemical and mechanical properties were improved by the presence of the Silica fiber. • The electrolyte showed a t Li+ value of 0.36 with an anodic stability extended up to 4.5 V vs. Li/Li + . • A prototype Li/LiFePO 4 polymer cell delivered a discharge capacity of about 100 mAh g −1 (75 °C, C/15). - Abstract: Poly(ethylene carbonate) (PEC) is known as an alternating copolymer derived from carbon dioxide (CO 2 ) and an epoxide as monomers. Here, we describe a new quaternary PEC-based composite electrolyte containing lithium bis(trifluoromethanesulfonyl) imide (LiTFSI) salt, N-n-butyl-N-methylpyrrolidinium bis(trifluoromethanesulfonyl) imide (Pyr 14 TFSI) ionic liquid, and an electrospun silica (SiO 2 ) fiber (SiF) with a submicron diameter in view of its possible applications in solid-state Li polymer batteries. A free-standing electrolyte membrane is prepared by a solvent casting method. The Pyr 14 TFSI ionic liquid enhances the ionic conductivity of the electrolyte as a result of its plasticizing effect. The electrochemical properties, such as ionic conductivity and Li transference number (t Li+ ), as well as mechanical strength of the electrolyte, are further improved by the SiF. We show that the quaternary electrolyte has a conductivity of the order of 10 −7 S cm −1 at ambient temperature and a high t Li+ value of 0.36 with an excellent flexibility. A prototype Li polymer cell using LiFePO 4 as a cathode material is assembled and tested. We demonstrate that this battery delivers a reversible charge-discharge capacity close to 100 mAh g −1 at 75 °C and C/15 rate. We believe that this work may pave the road to utilize CO 2 as a carbon source for highly-demanded, functional battery materials in future

  10. Hydraulics and heat transfer in the IFMIF liquid lithium target: CFD calculations

    OpenAIRE

    Peña, A.; Esteban, G.A.; Sancho, J.; Kolesnik, V.; Abánades Velasco, Alberto

    2009-01-01

    CFD (Computational fluid dynamics) calculation turns out to be a good approximation to the real behavior of the lithium (Li) flow of the target of the international fusion materials irradiation facility (IFMIF). A three-dimensional (3D) modelling of the IFMIF design Li target assembly, made with the CFD commercial code ANSYS-FLUENT has been carried out. The simulation by a structural mesh is focused on the thermal-hydraulic analysis inside the Li jet flow. For, this purpose, the two deuteron ...

  11. D-shaped configurations in FTU for testing liquid lithium limiter: Preliminary studies and experiments

    Directory of Open Access Journals (Sweden)

    G. Ramogida

    2017-08-01

    A possible alternative connection of the poloidal field coils in FTU is here proposed, with the aim of achieving a true X-point configuration with a magnetic single null well inside the plasma chamber and strike points on the lithium limiter. A preliminary assessment of this design allowed estimating the required power supply upgrade and showed its compatibility with the existing mechanical structure and cooling system, at least for plasmas with current up to 300 kA and flat-top duration up to 4s.

  12. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  15. Rapid and simple pretreatment of human body fluids using electromembrane extraction across supported liquid membrane for capillary electrophoretic determination of lithium.

    Science.gov (United States)

    Strieglerová, Lenka; Kubáň, Pavel; Boček, Petr

    2011-05-01

    Electromembrane extraction was used for simultaneous sample cleanup and preconcentration of lithium from untreated human body fluids. The sample of a body fluid was diluted 100 times with 0.5 mM Tris solution and lithium was extracted by electromigration through a supported liquid membrane composed of 1-octanol into 100 mM acetic acid acceptor solution. Matrix compounds, such as proteins, red blood cells, and other high-molecular-weight compounds were efficiently retained on the supported liquid membrane. The liquid membrane was anchored in pores of a short segment of a polypropylene hollow fiber, which represented a low cost, single use, disposable extraction unit and was discarded after each use. Acceptor solutions were analyzed using capillary electrophoresis with capacitively coupled contactless conductivity detection (CE-C(4) D) and baseline separation of lithium was achieved in a background electrolyte solution consisting of 18 mM L-histidine and 40 mM acetic acid at pH 4.6. Repeatability of the electromembrane extraction-CE-C(4) D method was evaluated for the determination of lithium in standard solutions and real samples and was better than 0.6 and 8.2% for migration times and peak areas, respectively. The concentration limit of detection of 9 nM was achieved. The developed method was applied to the determination of lithium in urine, blood serum, blood plasma, and whole blood at both endogenous and therapeutic concentration levels. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  16. Liquid oxygen LOX compatibility evaluations of aluminum lithium (Al-Li) alloys: Investigation of the Alcoa 2090 and MMC weldalite 049 alloys

    Science.gov (United States)

    Diwan, Ravinder M.

    1989-01-01

    The behavior of liquid oxygen (LOX) compatibility of aluminum lithium (Al-Li) alloys is investigated. Alloy systems of Alcoa 2090, vintages 1 to 3, and of Martin Marietta Corporation (MMC) Weldalite 049 were evaluated for their behavior related to the LOX compatibility employing liquid oxygen impact test conditions under ambient pressures and up to 1000 psi. The developments of these aluminum lithium alloys are of critical and significant interest because of their lower densities and higher specific strengths and improved mechanical properties at cryogenic temperatures. Of the different LOX impact tests carried out at the Marshall Space Flight Center (MSFC), it is seen that in certain test conditions at higher pressures, not all Al-Li alloys are LOX compatible. In case of any reactivity, it appears that lithium makes the material more sensitive at grain boundaries due to microstructural inhomogeneities and associated precipitate free zones (PFZ). The objectives were to identify and rationalize the microstructural mechanisms that could be relaxed to LOX compatibility behavior of the alloy system in consideration. The LOX compatibility behavior of Al-Li 2090 and Weldalite 049 is analyzed in detail using microstructural characterization techniques with light optical metallography, scanning electron microscopy (SEM), electron microprobe analysis, and surface studies using secondary ion mass spectrometry (SIMS), electron spectroscopy in chemical analysis (ESCA) and Auger electron spectroscopy (AES). Differences in the behavior of these aluminum lithium alloys are assessed and related to their chemistry, heat treatment conditions, and microstructural effects.

  17. 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

  18. Method of producing spherical lithium aluminate particles

    International Nuclear Information System (INIS)

    Yang, L.; Medico, R.R.; Baugh, W.A.

    1983-01-01

    Spherical particles of lithium aluminate are formed by initially producing aluminium hydroxide spheroids, and immersing the spheroids in a lithium ion-containing solution to infuse lithium ions into the spheroids. The lithium-infused spheroids are rinsed to remove excess lithium ion from the surface, and the rinsed spheroids are soaked for a period of time in a liquid medium, dried and sintered to form lithium aluminate spherical particles. (author)

  19. Novel Pyrrolinium-based Ionic Liquids for Lithium Ion Batteries: Effect of the Cation on Physicochemical and Electrochemical Properties

    International Nuclear Information System (INIS)

    Kim, Hyung-Tae; Kwon, Oh Min; Mun, Junyoung; Oh, Seung M.; Yim, Taeeun; Kim, Young Gyu

    2017-01-01

    Lithium ion batteries (LIBs) are one of the most promising energy conversion/storage systems, but the low thermal stability of the current electrolytes in LIBs should be improved to expand their potential applications. To enhance the safety properties of LIBs, novel pyrrolinium-based ionic liquids (ILs) were proposed as an alternative electrolyte to the current carbonate electrolyte, which have some task-specific functional groups, i.e., a planar C=N double bond, a C-O ether linkage, and no unstable C-H bond, designed to improve their electrochemical performances as well as the physicochemical properties. As a result, the pyrrolinium-based ILs exhibited much improved physicochemical and electrochemical properties compared to those of the known ILs. Among the prepared ILs, N-allyl-2-methoxypyrrolinium bis(fluorosulfonyl)imide (A(OMe)Pyrl-FSI, 4) showed the high ionic conductivity (10.2 mS cm −1 ), the very good cycling performance (99.3% of retention ratio after 50 cycles) with a LiFePO 4 electrode, and the much improved lithium ion transference number (0.19). IL 4 also had the remarkable rate capability at 5 C-rate with the retention ratio of 81.2% (124.8 mA h g −1 ), compared to the initial discharge capacity of 153.7 mA h g −1 at 0.1 C-rate. In addition, both their high thermal stability and non-flammability were also confirmed.

  20. Method and apparatus to produce and maintain a thick, flowing, liquid lithium first wall for toroidal magnetic confinement DT fusion reactors

    Science.gov (United States)

    Woolley, Robert D.

    2002-01-01

    A system for forming a thick flowing liquid metal, in this case lithium, layer on the inside wall of a toroid containing the plasma of a deuterium-tritium fusion reactor. The presence of the liquid metal layer or first wall serves to prevent neutron damage to the walls of the toroid. A poloidal current in the liquid metal layer is oriented so that it flows in the same direction as the current in a series of external magnets used to confine the plasma. This current alignment results in the liquid metal being forced against the wall of the toroid. After the liquid metal exits the toroid it is pumped to a heat extraction and power conversion device prior to being reentering the toroid.

  1. Innovative application of ionic liquid to separate Al and cathode materials from spent high-power lithium-ion batteries

    Energy Technology Data Exchange (ETDEWEB)

    Zeng, Xianlai; Li, Jinhui, E-mail: jinhui@tsinghua.edu.cn

    2014-04-01

    Highlights: • Manual dismantling is superior in spent high-power LiBs recycling. • Heated ionic liquid can effectively separate Al and cathode materials. • Fourier’s law was adopted to determine the heat transfer mechanism. • The process of spent LiBs recycling with heated ionic liquid dismantling was proposed. - Abstract: Because of the increasing number of electric vehicles, there is an urgent need for effective recycling technologies to recapture the significant amount of valuable metals contained in spent lithium-ion batteries (LiBs). Previous studies have indicated, however, that Al and cathode materials were quite difficult to separate due to the strong binding force supplied by the polyvinylidene fluoride (PVDF), which was employed to bind cathode materials and Al foil. This research devoted to seek a new method of melting the PVDF binder with heated ionic liquid (IL) to separate Al foil and cathode materials from the spent high-power LiBs. Theoretical analysis based on Fourier’s law was adopted to determine the heat transfer mechanism of cathode material and to examine the relationship between heating temperature and retention time. All the experimental and theoretic results show that peel-off rate of cathode materials from Al foil could reach 99% when major process parameters were controlled at 180 °C heating temperature, 300 rpm agitator rotation, and 25 min retention time. The results further imply that the application of IL for recycling Al foil and cathode materials from spent high-power LiBs is highly efficient, regardless of the application source of the LiBs or the types of cathode material. This study endeavors to make a contribution to an environmentally sound and economically viable solution to the challenge of spent LiB recycling.

  2. Evaluation of US demo helium-cooled blanket options

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  3. Numerical study of the flow conditioner for the IFMIF liquid lithium target

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S., E-mail: sergej.gordeev@kit.edu [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute for Thechnology, Campus North, Hermann v. Helmholtz Platz 1, D76344, Eggenstein-Leopoldshafen (Germany); Gröschel, F. [KIT Fusion Program, Karlsruhe Institute for Thechnology, Campus North, Hermann v. Helmholtz Platz 1, D76344, Eggenstein-Leopoldshafen (Germany); Heinzel, V.; Hering, W.; Stieglitz, R. [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute for Thechnology, Campus North, Hermann v. Helmholtz Platz 1, D76344, Eggenstein-Leopoldshafen (Germany)

    2014-10-15

    Highlights: • A detailed numerical analysis of the flow conditioner efficiency has been performed. • The calculations show that the present design of the flow conditioner cannot suppress swirl motions emerging from the bend. • The transient simulation reveals flow instabilities between the separation zone and the accelerated outer region. • Calculation shows that pitched guide vanes upstream the elbow reduces a generation of backflow areas downstream. - Abstract: IFMIF (International Fusion Materials Irradiation Facility) is an accelerator-based deuteron–lithium (D–Li) neutron source to simulate the neutron irradiation field in a fusion reactor. The target assembly of the IFMIF consists of the flow conditioners and the nozzle, which has to form a stable lithium jet. This work focuses on a numerical study of the flow conditioner efficiency, in which two different types of flow conditioners are compared by means of a detailed numerical analysis with respect to specific hydraulic effects in the pipe elbow and the inflow conditioners. The adequateness of three different turbulence models to simulate a flow through a 90° bend of circular cross section has been examined. The calculations show that a honeycomb-screen combination is not capable to suppress effectively large scale swirl motions emerging from the bend. An increasing number of screens improves the flow uniformity downstream, but increases the pressure drop. In order to detect any transient effects in the separation area a flow straightener configuration consisting of a honeycomb with a subsequent screen has been analyzed by means of a detached eddy simulation (DES). A frequency analysis of the normalized static pressure amplitude conducted by means of a detached eddy simulation (DES) reveals instabilities in the shear layer between the separation zone and the accelerated outer region, which additionally increase the inhomogeneity of the axial velocity distribution. A set of six circumferentially

  4. Liquid wall boiler and moderator (BAM) for heavy ion-pellet fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.R.; Lazareth, O.; Fillo, J.

    1977-11-01

    Thick liquid wall blankets appear to be of great promise for heavy ion pellet fusion reactors. They avoid the severe problems of intense radiation and blast damage that would be encountered with solid blanket structures. The liquid wall material can be chosen so that its vapor pressure at the working temperature of the power cycle is well below the value at which it might interfere with the propagation of the heavy ion beam. The liquid wall can be arranged so that it does not contact any surrounding solid structure when the pellet explosion occurs, including the ends. The ends can be magnetically closed just before the pellet explosion, or a time phased flow can be used, which will leave a clear central zone into which the pellet is injected. Parametric analysis comparing three candidate liquid wall materials were carried out. The three materials were lithium, flibe, and lead (with a low concentration of disolved lithium). Lead appeared to be the best choice for the liquid wall, although any of the three should allow a practical reactor system. The parametric analyses examined the effects of pellet yield (0 to 10 GJ), pellet mass (3 g to 3 kg), liquid wall thickness (10 cm to 80 cm), vapor condensation time (0 to 10 milliseconds), degree of neutron moderation in the pellet (none to 100%), liquid wall chamber size (radius of 1.5 meters to 4 meters), Pb/Li 6 ratio (100 to 5,000), and thickness of graphite moderating zone behind the liquid wall

  5. Pyrrolidinium FSI and TFSI-Based Polymerized Ionic Liquids as Electrolytes for High-Temperature Lithium-Ion Batteries

    Directory of Open Access Journals (Sweden)

    Manfred Kerner

    2018-02-01

    Full Text Available Promising electrochemical and dynamical properties, as well as high thermal stability, have been the driving forces behind application of ionic liquids (ILs and polymerized ionic liquids (PILs as electrolytes for high-temperature lithium-ion batteries (HT-LIBs. Here, several ternary lithium-salt/IL/PIL electrolytes (PILel have been investigated for synergies of having both FSI and TFSI anions present, primarily in terms of physico-chemical properties, for unique application in HT-LIBs operating at 80 °C. All of the electrolytes tested have low Tg and are thermally stable ≥100 °C, and with TFSI as the exclusive anion the electrolytes (set A have higher thermal stabilities ≥125 °C. Ionic conductivities are in the range of 1 mS/cm at 100 °C and slightly higher for set A PILel, which, however, have lower oxidation stabilities than set B PILel with both FSI and TFSI anions present: 3.4–3.7 V vs. 4.2 V. The evolution of the interfacial resistance increases for all PILel during the first 40 h, but are much lower for set B PILel and generally decrease with increasing Li-salt content. The higher interfacial resistances only influence the cycling performance at high C-rates (1 C, where set B PILel with high Li-salt content performs better, while the discharge capacities at the 0.1 C rate are comparable. Long-term cycling at 0.5 C, however, shows stable discharge capacities for 100 cycles, with the exception of the set B PILel with high Li-salt content. Altogether, the presence of both FSI and TFSI anions in the PILel results in lower ionic conductivities and decreased thermal stabilities, but also higher oxidation stabilities and reduced interfacial resistances and, in total, result in an improved rate capability, but compromised long-term capacity retention. Overall, these electrolytes open for novel designs of HT-LIBs.

  6. Demonstration of a high-intensity neutron source based on a liquid-lithium target for Accelerator based Boron Neutron Capture Therapy.

    Science.gov (United States)

    Halfon, S; Arenshtam, A; Kijel, D; Paul, M; Weissman, L; Berkovits, D; Eliyahu, I; Feinberg, G; Kreisel, A; Mardor, I; Shimel, G; Shor, A; Silverman, I; Tessler, M

    2015-12-01

    A free surface liquid-lithium jet target is operating routinely at Soreq Applied Research Accelerator Facility (SARAF), bombarded with a ~1.91 MeV, ~1.2 mA continuous-wave narrow proton beam. The experiments demonstrate the liquid lithium target (LiLiT) capability to constitute an intense source of epithermal neutrons, for Accelerator based Boron Neutron Capture Therapy (BNCT). The target dissipates extremely high ion beam power densities (>3 kW/cm(2), >0.5 MW/cm(3)) for long periods of time, while maintaining stable conditions and localized residual activity. LiLiT generates ~3×10(10) n/s, which is more than one order of magnitude larger than conventional (7)Li(p,n)-based near threshold neutron sources. A shield and moderator assembly for BNCT, with LiLiT irradiated with protons at 1.91 MeV, was designed based on Monte Carlo (MCNP) simulations of BNCT-doses produced in a phantom. According to these simulations it was found that a ~15 mA near threshold proton current will apply the therapeutic doses in ~1h treatment duration. According to our present results, such high current beams can be dissipated in a liquid-lithium target, hence the target design is readily applicable for accelerator-based BNCT. Copyright © 2015 Elsevier Ltd. All rights reserved.

  7. Facile preparation of polymer electrolytes based on the polymerized ionic liquid poly((4-vinylbenzyl)trimethylammonium bis(trifluoromethanesulfonylimide)) for lithium secondary batteries

    International Nuclear Information System (INIS)

    Li, Mingtao; Wang, Lu; Yang, Bolun; Du, Tingting; Zhang, Ying

    2014-01-01

    Graphical abstract: (A) The main components of PIL electrolytes, (B) A PIL electrolyte sample. - Highlights: • A new polymer electrolyte incorporating a DEME-TFSI liquid is prepared. • The ionic conductivity of the electrolytes reaches 7.58 × 10 −4 S cm −1 at 60 °C. • Batteries discharge 130 mAh g −1 at 0.1 C rates with good capacity retention. - Abstract: The polymer electrolytes based on a novel poly((4-vinylbenzyl)trimethylammonium bis(trifluoromethanesulfonylimide)) polymeric ionic liquid (PIL) as polymer host and containing DEME-TFSI ionic liquid, LiTFSI salt and nano silica are prepared. The polymer electrolyte is chemically stable even at a higher temperature of 60 °C in contact with lithium anode. Particularly, the electrolyte exhibits high lithium ion conductivity, wide electrochemical stability window and good lithium stripping/plating performance. When the IL content reaches 60% (the weight ratio of DEME-TFSI/PIL), the PIL electrolyte presents a higher ionic conductivity, and it is 7.58 × 10 −4 S cm −1 at 60 °C. Preliminary battery tests show that Li/LiFePO 4 cells with the PIL electrolytes are capable to deliver above 130 mAh g −1 at 60 °C with very good capacity retention

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

    International Nuclear Information System (INIS)

    Logan, B.G.

    1992-01-01

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

  9. 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.

  10. First wall and blanket design for the STARFIRE commercial tokamak power reactor

    International Nuclear Information System (INIS)

    Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

    1979-01-01

    The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

  11. Peningkatan mutu blanket karet alam melalui proses predrying dan penyemprotan asap cair

    Directory of Open Access Journals (Sweden)

    Afrizal Vachlepi

    2017-06-01

    Full Text Available Most of Indonesian rubber products SIR 20 are made from the material of raw rubber obtained from smallholders. However, the quality of this material is not good enough. Thus, quality improvement has to be carried out by manufacturers. The liquid smoke used during the blanket hanging process can improve the quality of the rubber products SIR 20. This research aimed to determine and study the effects of liquid smoke spraying and blanket hanging duration on the drying factor, the dry rubber content, technical quality, vulcanization characteristics, and physical properties of vulcanized natural rubber. Treatments consisted of various hanging duration (6, 8, and 10 days, and without hanging and spraying (with and without spraying of liquid smoke. The results showed that the spraying of liquid smoke on natural rubber blankets could improve the technical quality of the natural rubber, especially the values of Po and PRI. The spraying of liquid smoke could reduce the blanket hanging duration to 6-8 days. The blankets sprayed with liquid smoke had the optimum cure time of around 15 minutes and 19 seconds and the scorch time of around 3 minutes and 22 seconds. These values indicated that the vulcanization characteristics of blankets which were sprayed with liquid smoke were generally better than those of blankets which were not sprayed with liquid smoke

  12. 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.

  13. The significance of elemental sulfur dissolution in liquid electrolyte lithium sulfur batteries

    NARCIS (Netherlands)

    Harks, Peter Paul R.M.L.; Robledo, Carla B.; Verhallen, Tomas W.; Notten, Peter H.L.; Mulder, Fokko M.

    2017-01-01

    It is shown that the dissolution of elemental sulfur into, and its diffusion through, the electrolyte allows cycling of lithium–sulfur batteries in which the sulfur is initially far removed and electrically insulated from the current collector. These findings help to understand why liquid

  14. Liquid lithium corrosion research. Progress report, April 1, 1975--December 31, 1975

    International Nuclear Information System (INIS)

    Olson, D.L.; Bradley, W.L.

    1975-01-01

    Kinetics for grain boundary penetration and weight loss were determined for 304L stainless steel in Ti-gettered liquid Li. Empirical weight loss rate coefficient expressions were determined as function of both N content and temperature which predict the weight loss behavior of stainless steel for Li with high N content. Capillarity was eliminated as a possible driving force for grain boundary penetration of Fe by Li with the evidence of a grain boundary groove. Coupled ionic diffusion in a corrosion product surface film is suggested as the controlling mechanism for grain boundary grooving of Fe in N-contaminated liquid Li. Stress-enhanced grain boundary penetration showed that the penetration of stressed Fe is a function of t sup 1/3 rather than t sup 1/2, suggesting that primary creep of the Fe is continually breaking the protective corrosion product in the grain boundaries. A new purification procedure and systems were designed and constructed to achieve very low N contents in the liquid Li (less than 300 ppM N). A gas purification train, based on a molten Al bubbler for the removal of O and N, was designed and built, and vacuum melting of as-received Li and gettering are being initiated. Experiments designed to determine the role of Cr in stainless steel resistance are in progress. Also experiments have been designed which will investigate how liquid metal corrosion is a function of the contact potential between the metals

  15. Thermoelectric Efficiency Improvement in Vacuum Tubes of Decomposing Liquid Lithium-Ammonia Solutions

    International Nuclear Information System (INIS)

    Lee, Jungyoon; Kim, Miae; Shim, Kyuchol; Kim, Jibeom; Jeon, Joonhyeon

    2013-01-01

    Lithium-ammonia (Li-NH 3 ) solutions are possible to be successfully made under the vacuum condition but there still remains a problem of undergoing stable and reliable decomposition in vacuum for high-efficiency thermoelectric power generation. This paper describes a new method for improving the thermoelectric conversion efficiency of Li-NH 3 solutions in vacuum. The proposed method uses a ‘U’-shaped Pyrex vacuum tube for the preparation and decomposition of pure fluid Li-NH 3 solutions. The tube is shaped so that a gas passageway (‘U’) connecting both legs of the ‘U’ helps to balance pressure inside both ends of the tube (due to NH 3 gasification) during decomposition on the hot side. Thermoelectric experimental results show that solution reaction in the ‘U’-shaped tube proceeds more stably and efficiently than in the ‘U’-shaped tube, and consequently, thermoelectric conversion efficiency is improved. It is also proved that the proposed method can provide a reversible reaction, which can rotate between synthesis and decomposition in the tube, for deriving the long-time, high-efficiency thermoelectric power

  16. Innovative application of ionic liquid to separate Al and cathode materials from spent high-power lithium-ion batteries.

    Science.gov (United States)

    Zeng, Xianlai; Li, Jinhui

    2014-04-30

    Because of the increasing number of electric vehicles, there is an urgent need for effective recycling technologies to recapture the significant amount of valuable metals contained in spent lithium-ion batteries (LiBs). Previous studies have indicated, however, that Al and cathode materials were quite difficult to separate due to the strong binding force supplied by the polyvinylidene fluoride (PVDF), which was employed to bind cathode materials and Al foil. This research devoted to seek a new method of melting the PVDF binder with heated ionic liquid (IL) to separate Al foil and cathode materials from the spent high-power LiBs. Theoretical analysis based on Fourier's law was adopted to determine the heat transfer mechanism of cathode material and to examine the relationship between heating temperature and retention time. All the experimental and theoretic results show that peel-off rate of cathode materials from Al foil could reach 99% when major process parameters were controlled at 180°C heating temperature, 300 rpm agitator rotation, and 25 min retention time. The results further imply that the application of IL for recycling Al foil and cathode materials from spent high-power LiBs is highly efficient, regardless of the application source of the LiBs or the types of cathode material. This study endeavors to make a contribution to an environmentally sound and economically viable solution to the challenge of spent LiB recycling. Copyright © 2014 Elsevier B.V. All rights reserved.

  17. 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

  18. Tribological Behavior of Si3N4/Ti3SiC2 Contacts Lubricated by Lithium-Based Ionic Liquids

    Directory of Open Access Journals (Sweden)

    Haizhong Wang

    2014-01-01

    Full Text Available The tribological performance of Si3N4 ball sliding against Ti3SiC2 disc lubricated by lithium-based ionic liquids (ILs was investigated using an Optimol SRV-IV oscillating reciprocating friction and wear tester at room temperature (RT and elevated temperature (100°C. Glycerol and the conventional imidazolium-based IL 1-hexyl-3-methylimidazolium bis(trifluoromethylsulfonylimide (L-F106 were used as references under the same experimental conditions. The results show that the lithium-based ILs had higher thermal stabilities than glycerol and lower costs associated with IL preparation than L-F106. The tribotest results show that the lithium-based ILs were effective in reducing the friction and wear of Si3N4/Ti3SiC2 contacts. [Li(urea]TFSI even produced better tribological properties than glycerol and L-F106 both at RT and 100°C. The SEM/EDS and XPS results reveal that the excellent tribological endurance of Si3N4/Ti3SiC2 contacts lubricated by lithium-based ILs was mainly attributed to the formation of surface protective films composed of various tribochemical products.

  19. An approach of ionic liquids/lithium salts based microwave irradiation pretreatment followed by ultrasound-microwave synergistic extraction for two coumarins preparation from Cortex fraxini.

    Science.gov (United States)

    Liu, Zaizhi; Gu, Huiyan; Yang, Lei

    2015-10-23

    Ionic liquids/lithium salts solvent system was successfully introduced into the separation technique for the preparation of two coumarins (aesculin and aesculetin) from Cortex fraxini. Ionic liquids/lithium salts based microwave irradiation pretreatment followed by ultrasound-microwave synergy extraction (ILSMP-UMSE) procedure was developed and optimized for the sufficient extraction of these two analytes. Several variables which can potentially influence the extraction yields, including pretreatment time and temperature, [C4mim]Br concentration, LiAc content, ultrasound-microwave synergy extraction (UMSE) time, liquid-solid ratio, and UMSE power were optimized by Plackett-Burman design. Among seven variables, UMSE time, liquid-solid ratio, and UMSE power were the statistically significant variables and these three factors were further optimized by Box-Behnken design to predict optimal extraction conditions and find out operability ranges with maximum extraction yields. Under optimum operating conditions, ILSMP-UMSE showed higher extraction yields of two target compounds than those obtained by reference extraction solvents. Method validation studies also evidenced that ILSMP-UMSE is credible for the preparation of two coumarins from Cortex fraxini. This study is indicative of the proposed procedure that has huge application prospects for the preparation of natural products from plant materials. Copyright © 2015 Elsevier B.V. All rights reserved.

  20. Vanadium—lithium in-pile loop for comprehensive tests of vanadium alloys and multipurpose coatings

    Science.gov (United States)

    Lyublinski, I. E.; Evtikhin, V. A.; Ivanov, V. B.; Kazakov, V. A.; Korjavin, V. M.; Markovchev, V. K.; Melder, R. R.; Revyakin, Y. L.; Shpolyanskiy, V. N.

    1996-10-01

    The reliable information on design and material properties of self-cooled Li sbnd Li blanket and liquid metal divertor under neutron radiation conditions can be obtained using the concept of combined technological and material in-pile tests in a vanadium—lithium loop. The method of in-pile loop tests includes studies of vanadium—base alloys resistance, weld resistance under mechanical stress, multipurpose coating formation processes and coatings' resistance under the following conditions: high temperature (600-700°C), lithium velocities up to 10 m/s, lithium with controlled concentration of impurities and technological additions, a neutron load of 0.4-0.5 MW/m 2 and level of irradiation doses up to 5 dpa. The design of such an in-pile loop is considered. The experimental data on corrosion and compatibility with lithium, mechanical properties and welding technology of the vanadium alloys, methods of coatings formation and its radiation tests in lithium environment in the BOR-60 reactor (fast neutron fluence up to 10 26 m -2, irradiation temperature range of 500-523°C) are presented and analyzed as a basis for such loop development.

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

  5. Structural stability and self-healing capability of Er2O3 in situ coating on V-4Cr-4Ti in liquid lithium

    International Nuclear Information System (INIS)

    Yao, Zhenyu; Suzuki, Akihiro; Muroga, Takeo; Nagasaka, Takuya

    2006-01-01

    The in situ Er 2 O 3 insulating coating is under development for the self-cooled Li/V-alloy type fusion blanket. In this study, the structural stability and self-healing capability of the coating are investigated. Since the cracking in the coating was not observed after exposure when Li was removed with a weak lotion (liquid NH 3 ), the cracking observed in the previous studies is not a practical issue in a real blanket. The re-exposure of the coating in pure Li showed that the coating once formed in Li (Er) is thought to be stable in pure Li. Thus, coating has the possibility to be serviced in a Li environment without an Er supply. By prior exposure to Li (Er) at 873 K, the exhaustion of the oxygen storage in V-alloy substrate during exposure at 973 K could be delayed effectively. The self-healing capability of the coating was demonstrated by the examination with the re-exposing cracked coating in Li (Er)

  6. 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)

  7. On the Development of Hydrogen Isotope Extraction Technologies for a Full LiMIT-Style PFC Liquid Lithium Loop

    Science.gov (United States)

    Christenson, Michael; Szott, Matthew; Stemmley, Steven; Mettler, Jeremy; Wendeborn, John; Moynihan, Cody; Ahn, Chisung; Andruczyk, Daniel; Ruzic, David

    2017-10-01

    Lithium has proven over numerous studies to improve core confinement, allowing access to operational regimes previously unattainable when using solid, high-Z divertor and limiter modules in magnetic confinement devices. Lithium readily absorbs fuel species, and while this is advantageous, it is also detrimental with regards to tritium inventory and safety concerns. As such, extraction technologies for the recovery of hydrogenic isotopes captured by lithium require development and testing in the context of a larger lithium loop recycling system. Proposed reclamation technologies at the University of Illinois at Urbana-Champaign (UIUC) will take advantage of the thermophysical properties of the lithium-hydrogen-lithium hydride system as the driving force for recovery. Previous work done at UIUC indicates that hydrogen release from pure lithium hydride reaches a maximum of 7 x 1018 s-1 at 665 °C. While this recovery rate is appreciable, reactor-scale scenarios will require isotope recycling to happen on an even faster timescale. The ratio of isotope dissolution to hydride precipitate formation must therefore be determined, along with the energy needed to recoup trapped hydrogen isotopes. Extraction technologies for use with a LiMIT-style loop system will be discussed and results will be presented. DOE/ALPS DE-FG02-99ER54515.

  8. 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.)

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

    International Nuclear Information System (INIS)

    Woodruff, G.L.

    1987-01-01

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

  11. Compatibility of dip-coated Er2O3 coating by MOD method with liquid Li

    International Nuclear Information System (INIS)

    Zhang Dongxun; Kondo, Masatoshi; Tanaka, Teruya; Muroga, Takeo; Valentyn, Tsisar

    2011-01-01

    An electrical insulating ceramic coating on the self-cooled lithium blanket is a promising technology for suppressing MHD pressure drop in the blanket system. Er 2 O 3 is thought to be one of the potential candidate materials for ceramic coatings because of their high electrical resistivity and high compatibility with liquid lithium. In this study, Er 2 O 3 coating was fabricated on the ferritic steels by dip-coating method with MOD (metal organic decomposition) liquid precursor followed by baking in different atmosphere. The coated specimens were immersed at 500 o C in the static liquid lithium to test the compatibility. It was shown that the compatibility of the coating was degraded when Fe 2 O 3 or Fe 3 O 4 was formed as the main composition of the substrate oxidation layer during the baking. On the other hand, thin Cr 2 O 3 layer in the substrate oxidation layer did not influence the stability of Er 2 O 3 coating. Atmosphere controlling for suppressing the substrate oxidation, especially Fe 2 O 3 or Fe 3 O 4 , during the baking is shown to be essential for the compatibility of MOD Er 2 O 3 coating on ferritic steels.

  12. A solid-breeder blanket and power conversion system for the Mirror Advanced Reactor Study (MARS)

    International Nuclear Information System (INIS)

    Bullis, R.; Clarkson, I.

    1983-01-01

    A solid-breeder blanket has been designed for a commercial fusion power reactor based on the tandem mirror concept (MARS). The design utilizes lithium oxide, cooled by helium which powers a conventional steam electric generating cycle. Maintenance and fabricability considerations led to a modular configuration 6 meters long which incorporates two magnets, shield, blanket and first wall. The modules are arranged to form the 150 meter long reactor central cell. Ferritic steel is used for the module primary structure. The lithium oxide is contained in thin-walled vanadium alloy tubes. A tritium breeding ratio of 1.25 and energy multiplication of 1.1 is predicted. The blanket design appears feasible with only a modest advance in current technology

  13. Study of interactions between liquid lead-lithium alloy and austenitic and martensitic steels

    International Nuclear Information System (INIS)

    Simon, N.

    1992-06-01

    In the framework of Fusion Technology, the behaviour of structural materials in presence of liquid alloy Pb17Li is investigated. First, the diffusion coefficients of Fe and Cr have been determined at 500 deg C. Then mass transfer experiments in Pb17Li have been conducted in an anisothermal container with pure metals (Fe, Cr, Ni), Fe-Cr steels and austenitic steels. These experiments showed a very high loss of Nickel, which is an accordance with its high solubility, and Cr showed mass-losses one order of magnitude higher than for pure iron, as the diffusion coefficient of Cr is three orders of magnitude higher than for pure Fe. The corrosion rate of binary Fe-Cr and pure Fe are identical. In austenitic steels, the gamma lattice allows a higher mass-transfer of Cr than the alpha lattice, the presence of Cr slows downs the dissolution of Ni, and the porosity of corrosion layers results of losses of Cr and Ni. Finally, a review of our results and those of other laboratories allowed an identification of the corrosion limiting step. In the case of 1.4914 martensitic steel it is the diffusion of Fe in Pb17Li, while in the case of 316L austenitic steel it is the diffusion of Cr in Pb17Li

  14. A Liquid Inorganic Electrolyte Showing an Unusually High Lithium Ion Transference Number: A Concentrated Solution of LiAlCl4 in Sulfur Dioxide

    Directory of Open Access Journals (Sweden)

    Martin Winter

    2013-08-01

    Full Text Available We report on studies of an inorganic electrolyte: LiAlCl4 in liquid sulfur dioxide. Concentrated solutions show a very high conductivity when compared with typical electrolytes for lithium ion batteries that are based on organic solvents. Our investigations include conductivity measurements and measurements of transference numbers via nuclear magnetic resonance (NMR and by a classical direct method, Hittorf’s method. For the use of Hittorf’s method, it is necessary to measure the concentration of the electrolyte in a selected cell compartment before and after electrochemical polarization very precisely. This task was finally performed by potentiometric titration after hydrolysis of the salt. The Haven ratio was determined to estimate the association behavior of this very concentrated electrolyte solution. The measured unusually high transference number of the lithium cation of the studied most concentrated solution, a molten solvate LiAlCl4 × 1.6SO2, makes this electrolyte a promising alternative for lithium ion cells with high power ability.

  15. Liquid-metal flow through a thin-walled elbow in a plane perpendicular to a uniform magnetic field

    International Nuclear Information System (INIS)

    Walker, J.S.

    1986-04-01

    This paper presents analytical solutions for the liquid-metal flow through two straight pipes connected by a smooth elbow with the same inside radius. The pipes and the elbow lie in a plane which is perpendicular to a uniform, applied magnetic field. The strength of the magnetic field is assumed to be sufficiently strong that inertial and viscous effects are negligible. This assumption is appropriate for the liquid-lithium flow in the blanket of a magnetic confinement fusion reactor, such as a tokamak. The pipes and the elbow have thin metal walls

  16. 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

  17. Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

    2016-11-01

    Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

  18. Thermal analysis of a helium-cooled, tube-bank