25 results on '"Dai, Kai"'
Search Results
2. RADIATIVE TRANSFER IN AN ABSORBING-SCATTERING MEDIUM
- Author
-
Dai-Kai Sze, H. C. Hottel, and A. F. Sarofim
- Subjects
Materials science ,Scattering ,Radiative transfer ,Molecular physics - Published
- 2019
3. Tritium Control for Flibe/V-Alloy Blanket System
- Author
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Teruya Tanaka, Takeo Muroga, Dai-Kai Sze, Zaixin Li, and Akio Sagara
- Subjects
Nuclear and High Energy Physics ,Neutron transport ,Materials science ,020209 energy ,Mechanical Engineering ,Nuclear engineering ,FLiBe ,02 engineering and technology ,Partial pressure ,Fusion power ,Blanket ,01 natural sciences ,010305 fluids & plasmas ,Surface coating ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Nuclear fusion ,General Materials Science ,Tritium ,Civil and Structural Engineering - Abstract
One of the critical issues of Flibe/V-alloy blanket with REDOX control by Be is a large tritium inventory in V-alloy structures. Among the possible solutions to this issue would be to control REDOX not by Be but by addition of MoF 6 or WF 6 enhancing the reaction from T 2 to TF. The present study investigated feasibility of this procedure by thermodynamic and neutronics calculations. Using the blanket dimensions of Force Free Helical Reactor (FFHR), tritium inventory in V-alloy structure and Flibe were estimated based on the calculated equilibrium partial pressures of T 2 and TF in various cases of REDOX control by MoF 6 or WF 6 . Also carried out were neutronics examinations for the impact of Mo or W doping in the blanket. The results showed that the tritium inventory in the blanket area would be less than 100g at the TF level of 0.1 and 1 ppm in Flibe with addition of WF 6 and MoF 6 , respectively. WF 6 doping is far more advantageous than MoF 6 doping for low activation purposes.
- Published
- 2007
4. Chemical treatment of carbon nanotubes as electrodes in electrochemical double-layer capacitors
- Author
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Yu Bing-kun, Zhang Dengsong, Fang Jian-hui, Dai Kai, and Shi Liyi
- Subjects
Materials science ,Carbon nanofiber ,General Mathematics ,Inorganic chemistry ,General Engineering ,Electrolyte ,Carbon nanotube ,Chemical vapor deposition ,Electrochemistry ,law.invention ,Potential applications of carbon nanotubes ,law ,Frit compression ,Carbon nanotube supported catalyst - Abstract
Multi-walled carbon nanotubes with homogeneous diameters (40–60 ran), produced by chemical vapor deposition of hydrocarbon gas, are purified by nitric acids. Infrared and Raman studies indicate that oxygen containing surface groups, which are predominately carboxylic, phenolic and lactonic groups, are introduced into purified carbon nanotubes. Then three kinds of block-form porous tablets of carbon nanotubes are fabricated as electrodes in electrochemical double-layer capacitors. Using mounded mixture comprising carbon nanotubes and binder powders provides these tablets. Comparison of the effect of different processing on the structural performance of the capacitors is specifically investigated. Using chemically treated electrodes, electrochemical double-layer capacitors with a specific capacitance of about 33 F/g are obtained with 38 wt% H2 SO4 as the electrolyte.
- Published
- 2005
5. Thermo-Physical Properties and Equilibrium Vapor-Composition of Lithium Fluoride-Beryllium Fluoride (2LiF/BeF2) Molten Salt
- Author
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A. René Raffray, Mofreh R. Zaghloul, and Dai-Kai Sze
- Subjects
Nuclear and High Energy Physics ,Materials science ,Vapor pressure ,020209 energy ,Mechanical Engineering ,FLiBe ,Vapour pressure of water ,Lithium fluoride ,02 engineering and technology ,Enthalpy of vaporization ,01 natural sciences ,010305 fluids & plasmas ,Beryllium fluoride ,Surface tension ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,Chemical engineering ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Physical chemistry ,General Materials Science ,Molten salt ,Civil and Structural Engineering - Abstract
An assessment of Flibe thermo-physical properties relevant to the prompt x-rays ablation of the liquid wall is presented with emphasis given to the equilibrium vapor composition and vapor pressure....
- Published
- 2003
6. Nuclear Performance of the Thin-Liquid FW Concept of the CLiFF Design
- Author
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Mahmoud Z. Youssef, Neil B. Morley, and Dai-Kai Sze
- Subjects
Toroid ,Materials science ,020209 energy ,FLiBe ,Nuclear engineering ,General Engineering ,chemistry.chemical_element ,02 engineering and technology ,High power density ,Plasma ,Blanket ,Radiation ,01 natural sciences ,010305 fluids & plasmas ,chemistry.chemical_compound ,chemistry ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Neutron ,Beryllium - Abstract
The nuclear performance of the thin Convective Liquid Flow First Wall (CLiFF) concept is investigated. Liquid walls offer the advantage of protecting solid structure behind them from excessive damage from neutrons originated in the plasma and thus have the capability for high power density applications; the central research focus of the Advanced Power Extraction (APEX) study. In the present parametric and scoping work, several combinations of liquid breeder and structure type where investigated. The aim is to maximize local tritium breeding ratio (TBR), power multiplication, and ensuring that the vacuum vessel and toroidal coils are protected from excessive radiation. The candidate liquid breeders considered are Li, Flibe, and Sn-Li. Vanadium-alloy is deployed with Li while either Ferritic steel or SiC is deployed with Flibe and Sn-Li. Deployment of other refractory alloys and their impact on TBR was also studied. The introduction of a beryllium multiplier zone in the blanket was shown to enhance tritium production capability, particularly for those liquid breeders whose TBRs are marginal.
- Published
- 2001
7. Design and development of the Flibe blanket for helical-type fusion reactor FFHR
- Author
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O. Motojima, H. Yamanishi, T. Terai, Atsushi Suzuki, Satoru Takahashi, Hideki Matsui, Takeo Muroga, Tanaka Satoru, Tetsuji Noda, T. Uda, Y. Hosoya, Akira Kohyama, Ken-ichi Fukumoto, Shinsaku Imagawa, H. Hasizume, Saburo Toda, Dai-Kai Sze, Akihiko Shimizu, T. Yamamoto, and Akio Sagara
- Subjects
Thermal efficiency ,Materials science ,Mechanical Engineering ,FLiBe ,Nuclear engineering ,Blanket ,Fusion power ,Nuclear reactor ,law.invention ,chemistry.chemical_compound ,Breeder (animal) ,Nuclear Energy and Engineering ,chemistry ,law ,Heat exchanger ,General Materials Science ,Molten salt ,Civil and Structural Engineering - Abstract
Blanket design is in progress in helical-type compact reactor FFHR-2. A localized blanket concept is proposed by selecting molten-salt Flibe as a self-cooling tritium breeder from the main reason of safety: low tritium solubility, low reactivity with air and water, low pressure operation, and low MHD resistance which is compatible with the high magnetic field design in force-free helical reactor (FFHR). Numerical results are presented on nuclear analyses using the MCNP-4B code, on thermal and stress analyses using the ABAQUS code, and heat exchange efficiency from Flibe to He. R&D programs on Flibe engineering are also in progress in material dipping-tests and in construction of molten salt loop. Preliminary results in these experiments are also presented.
- Published
- 2000
8. The ARIES-RS power core—recent development in Li/V designs
- Author
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Laila El-Guebaly, Dick Cole, X.R. Wang, Lester M. Waganer, Mark S. Tillack, H.Y. Khater, Thanh Q. Hua, Michael C. Billone, Dai-Kai Sze, I.N. Sviatoslavsky, E. A. Mogahed, Siegfried Malang, Jeffrey A. Crowell, James Blanchard, Farrokh Najmabadi, and Dennis Lee
- Subjects
Tokamak ,Materials science ,Power station ,Mechanical Engineering ,Nuclear engineering ,Magnetic confinement fusion ,Fusion power ,Blanket ,law.invention ,Nuclear physics ,Breeder (animal) ,Nuclear Energy and Engineering ,law ,Beta (plasma physics) ,Energy transformation ,General Materials Science ,Civil and Structural Engineering - Abstract
The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study.
- Published
- 1998
9. Tritium recovery from lithium, based on a cold trap
- Author
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Richard F. Mattas, Hiroshi Yoshida, Dai-Kai Sze, James L. Anderson, O. Kveton, and Rem Haange
- Subjects
Air separation ,Materials science ,Mechanical Engineering ,Radiochemistry ,chemistry.chemical_element ,Fusion power ,Blanket ,Alkali metal ,Nuclear physics ,Nuclear Energy and Engineering ,chemistry ,General Materials Science ,Tritium ,Lithium ,Saturation (chemistry) ,Civil and Structural Engineering ,Cold trap - Abstract
A concept to recover tritium from lithium, based on a cold trap, has been developed as part of the U.S. contribution to ITER. The cold trap process can only reduce the tritium concentration to about 400 appm, which is far above the ITER design goal of reducing the tritium concentration in lithium to about 1 appm. To achieve this lower goal, protium is added to the lithium to a concentration higher than the saturation concentration of the hydrogen isotope at the cold trap temperature. Thus, LiH and LiT will precipitate out together at the cold trap. The tritium from the cold trap can be recovered by heating the Li(H + T) to 600 °C for decomposition. The H and T then can be separated by a cryogenic distillation process.
- Published
- 1995
10. Recent Designs for Advanced Fusion Reactor Blankets
- Author
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Dai-Kai Sze
- Subjects
Materials science ,Tokamak ,Divertor ,Nuclear engineering ,General Engineering ,Fusion power ,Blanket ,Coolant ,law.invention ,Nuclear physics ,law ,Beta (plasma physics) ,Neutron ,Waste disposal - Abstract
A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li{sub 2}ZrO{sub 3} was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li{sub 2}O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability more » regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D-{sup 3}He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view. « less
- Published
- 1994
11. Materials Recycling Considerations for D-T Fusion Reactors
- Author
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J.A. Sommers, E.T. Cheng, O.T. Farmer, and Dai-Kai Sze
- Subjects
Materials science ,Gamma dose ,020209 energy ,Shield ,Nuclear engineering ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,General Engineering ,02 engineering and technology ,Fusion power ,Blanket ,01 natural sciences ,010305 fluids & plasmas - Abstract
Materials recycling aspects including contact gamma dose rates and cooling times were investigated for the first wall, blanket, and shield components of future fusion power reactors. Candidate stru...
- Published
- 1992
12. MHD Considerations for a Self-Cooled Liquid Lithium Blanket
- Author
-
A. B. Hull, Dai-Kai Sze, Richard F. Mattas, D.L. Smith, and B. F. Picologlou
- Subjects
Pressure drop ,Liquid metal ,Materials science ,Tokamak ,020209 energy ,Nuclear engineering ,General Engineering ,02 engineering and technology ,Blanket ,engineering.material ,Fusion power ,01 natural sciences ,010305 fluids & plasmas ,law.invention ,Coating ,Physics::Plasma Physics ,law ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,engineering ,Magnetohydrodynamic drive ,Magnetohydrodynamics - Abstract
The magnetohydrodynamic (MHD) effects can present a feasibility issue for a self-cooled liquid metal blanket of magnetically confined fusion reactors, especially inboard regime of a tokamak. This pressure drop can be significantly reduced by using insulated wall structure. A self-healing insulating coating has been identified, which will reduce the pressure drop by more than a factor of 10. The future research direction to further quantify the performance of this coating is also outlined.
- Published
- 1992
13. Activation Product Safety in the ARIES-I Reactor Design
- Author
-
Edward T. Cheng, S.P. Grotz, J. Stephen Herring, C.P.C. Wong, and Dai-Kai Sze
- Subjects
Zirconium ,Materials science ,020209 energy ,Nuclear engineering ,Divertor ,General Engineering ,Radioactive waste ,chemistry.chemical_element ,02 engineering and technology ,Tungsten ,Blanket ,01 natural sciences ,010305 fluids & plasmas ,Breeder (animal) ,chemistry ,Activation product ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Waste disposal - Abstract
The ARIES design effort has sought to maximize the environmental and safety advantages of fusion through careful selection of materials and careful design. Three goals are that the reactor achieve inherent or passive safety, that no public evacuation plan be necessary and that the waste be disposable as 10CFR61 Class C waste. The ARIES-I reactor consists of a SiC composite structure for the first wall and blanket, cooled by 10 MPa He. The breeder is Li2ZrO3, although Li2O and Li4SiO4 were also considered. The divertor consists of SiC composite tubes coated with 2 mm of tungsten. Due to the minimal afterheat of this blanket design, LOCA calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. Two primary safety issues are the zirconium in the breeder and tungsten on the divertor. Li2ZrO3 was chosen because of its demonstrated high-temperature stability. The other breeders have lower afterheat and activation. Use of zirconium in the breeder will necess...
- Published
- 1991
14. Thermo-structural design of the ARIES-III divertor with organic coolant in subcooled flow boiling
- Author
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G.E. Orient, M. Valenti, C.P.C. Wong, Dai-Kai Sze, Farrokh Najmabadi, Shahram Sharafat, M.Z. Hasan, and E.E. Reis
- Subjects
Subcooling ,Materials science ,Heat flux ,Critical heat flux ,Pressurizer ,Divertor ,Heat transfer ,Mechanical engineering ,Mechanics ,Nucleate boiling ,Coolant - Abstract
The thermal-hydraulic and structural design of the ARIES-III divertor plate is presented. The divertor plate is made of small-diameter W-3Re tubes laid along the radial direction and coated with 4 mm of plasma-sprayed tungsten on the plasma-facing side to withstand one hard disruption. The plate is contoured to have the constant heat flux of 5.44 MW/m/sup 2/ on the entire surface. The total divertor thermal power of 629 MW is removed by the organic coolant HB-40 with the same inlet/exit temperatures (340 degrees C/425 degrees C) as in the first-wall/shield coolant circuit. The principal mode of heat transfer is by subcooled flow boiling. The inlet pressure is 5.34 MPa and the exit pressure is 4.3 MPa, which, by passing through an orifice, is reduced to 1 MPa, equal to the first-wall/shield exit pressure. The total coolant flow rate is 3.35 m/sup 3//s and the circulation power is 18 MWe. The maximum plate temperature is 821 degrees C. The safety factor with respect to the critical heat flux is >or=2, and it is approximately 3 with respect to the maximum allowable plate temperature. Maximum equivalent total, thermal, and pressure stresses are also given. >
- Published
- 2002
15. Liquid-Metal Corrosion
- Author
-
Dale L. Smith, Peter F. Tortorelli, Omesh K. Chopra, Dai-Kai Sze, and Jackson H. DeVan
- Subjects
Austenite ,Liquid metal ,Materials science ,Structural material ,020209 energy ,Metallurgy ,General Engineering ,Halide ,02 engineering and technology ,01 natural sciences ,Chemical reaction ,010305 fluids & plasmas ,Corrosion ,chemistry.chemical_compound ,chemistry ,Operating temperature ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Fluoride - Abstract
A review of corrosion and environmental effects on the mechanical properties of candidate structural alloys for use with liquid metals in fusion reactors is presented. The corrosion/mass transfer behavior of austenitic and ferritic steels and vanadium-base alloys is evaluated to determine the preliminary operating temperature limits for circulating and static liquid-lithium and Pb-17Li systems. The influence of liquid-metal environment on the mechanical properties of structural materials is discussed. Corrosion effects of nitrate and fluoride salts are presented. Requirements for additional data are identified.
- Published
- 1985
16. Neutronics Studies of the Gas-Carried Lithium Oxide Cooling-Breeding Fusion Reactor Blanket and Shield
- Author
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E. T. Cheng, Dai-Kai Sze, and T. Y. Sung
- Subjects
Nuclear and High Energy Physics ,Neutron transport ,Materials science ,020209 energy ,Nuclear engineering ,Isotopes of lithium ,Radiochemistry ,Pellets ,chemistry.chemical_element ,02 engineering and technology ,Fusion power ,Condensed Matter Physics ,020303 mechanical engineering & transports ,0203 mechanical engineering ,Nuclear Energy and Engineering ,chemistry ,Shield ,0202 electrical engineering, electronic engineering, information engineering ,Tritium ,Loss-of-coolant accident ,Helium - Abstract
The use of Li/sub 2/O pellets suspended in a stream of helium gas flowing through a fusion reactor blanket can serve the dual functions of tritium breeding and cooling. The neutronics properties, such as tritium breeding ratio and total nuclear heating, have been studied as functions of the enrichment of /sup 6/Li, the breeding and graphite zone thickness, and solid concentration in the suspension. The performance of the shield and the thermal reaction of the magnet in case of loss of coolant accident (LOCA) was also investigated. The radioactivity and afterheat in such blankets after shutdown were calculated.
- Published
- 1977
17. The INPORT concept — an improved method to protect ICF reactor first walls
- Author
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W.F. Vogelsang, I.N. Sviatoslavsky, Gregory A. Moses, Dai-Kai Sze, Mohamed E. Sawan, J. Sapp, and G.L. Kulcinski
- Subjects
Nuclear and High Energy Physics ,Liquid metal ,Materials science ,Nuclear transmutation ,Neutron spectra ,Improved method ,Metal ,Nuclear Energy and Engineering ,visual_art ,visual_art.visual_art_medium ,General Materials Science ,Vacuum chamber ,Composite material ,Porosity ,Displacement (fluid) - Abstract
A method to protect the first metallic walls of ICF reactors from X-rays and target debris has been developed. The concept utilizes porous, flexible tubes of woven C or SiC fibers to contain liquid metals inside the vacuum chamber of an ICF system. These porous tubes allow for ablation and recondensation of liquid metal films. The tubes also moderate the neutron spectra and reduce the displacement and transmutation damage in metallic walls.
- Published
- 1981
18. Thermal and Mechanical Design of a Double-Walled Steam Generator
- Author
-
Dai-Kai Sze, L. Pong, J. H. Huang, and D.C. Schluderberg
- Subjects
Double walled ,Materials science ,Nuclear engineering ,Thermal ,General Engineering ,Boiler (power generation) ,Mechanical design ,Fusion power ,health care economics and organizations - Abstract
A double-walled steam generator is used in the fusion power cycle to replace the intermediate loop. This will save the cost associated with a complicated system and the associated temperature degra...
- Published
- 1983
19. Design of Self-Cooled, Liquid-Metal Blankets for Tokamak and Tandem Mirror Reactors
- Author
-
Dale L. Smith, Saurin Majumdar, Basil F. Picologlou, Yung Sheng Cha, Yousry Gohar, Ahmed Hassanein, and Dai-Kai Sze
- Subjects
Neutron transport ,Liquid metal ,Materials science ,Tokamak ,Hydraulics ,020209 energy ,Nuclear engineering ,General Engineering ,chemistry.chemical_element ,02 engineering and technology ,Blanket ,01 natural sciences ,010305 fluids & plasmas ,law.invention ,Coolant ,Thermal hydraulics ,chemistry ,law ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Lithium - Abstract
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 engineeringmore » 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.« less
- Published
- 1985
20. Tritium Recovery from Liquid Lithium-Lead by Vacuum Degassing
- Author
-
Dai-Kai Sze, L. C. Wittenberg, K.E. Plute, and Edwin M. Larsen
- Subjects
Inert ,Liquid metal ,Materials science ,020209 energy ,Radiochemistry ,Ultra-high vacuum ,General Engineering ,02 engineering and technology ,Partial pressure ,01 natural sciences ,Purge ,010305 fluids & plasmas ,Nuclear physics ,Torr ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Tritium ,Inert gas - Abstract
Various techniques for tritium removal from the liquid eutectic Li/sub 17/Pb/sub 83/ under vacuum are considered as candidates for the tritium removal system (TRS) for the Mirror Advanced Reactor Study (MARS). The TRS baseline parameters require the removal of 60% of the tritium contained in the liquid metal at a tritium partial pressure of 1.0 x 10/sup -4/ torr (0.013 Pa). Degassing from a droplet spray was chosen as the preferred design option, although removal from thin films is a feasible alternative. Vacuum removal from a stirred pool was rejected because of the size and relatively poor transport conditions. The use of an inert purge gas was also rejected due to the large purge gas flow rate and the problem of separating tritium from a large quantity of inert gas.
- Published
- 1983
21. Water-Cooled Solid-Breeder Blanket Concept for ITER
- Author
-
Michael C. Billone, R.C. Clemmer, Yousry Gohar, Carl E. Johnson, Saurin Majumdar, Dai-Kai Sze, Richard F. Mattas, H. Attaya, Charles C. Baker, Ahmed Hassanein, Patricia A. Finn, H.C. Stevens, L.R. Turner, and D.L. Smith
- Subjects
Materials science ,020209 energy ,Water cooled ,Nuclear engineering ,General Engineering ,chemistry.chemical_element ,02 engineering and technology ,Blanket ,Key features ,01 natural sciences ,010305 fluids & plasmas ,Reliability (semiconductor) ,Breeder (animal) ,chemistry ,Tritium breeding ratio ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Beryllium ,Decay heat - Abstract
A water cooled solid-breeder blanket concept was developed for ITER. The main function of this blanket is to produce the necessary tritium for the ITER operation. Several design features are incorporated in this blanket concept to increase its attractiveness. The main features are the following: (a) a multilayer concept which reduces fabrication cost; (b) a simple blanket configuration which results in reliability advantages; (c) a very small breeder volume is employed to reduce the tritium inventory and the blanket cost; (d) a high tritium breeding ratio eliminates the need for an outside tritium supply; (e) a low-pressure system decreases the required steel fraction for structural purposes; (f) a low-temperature operation reduces the swelling concerns for beryllium; and (g) the small fractions of structure and breeder materials used in the blanket reduce the decay heat source. The key features and design analyses of this blanket are summarized in this paper.
- Published
- 1989
22. A Molten Salt Cooling/17Li-83Pb Breeding Blanket Concept
- Author
-
E. T. Cheng and Dai-Kai Sze
- Subjects
Materials science ,Metallurgy ,General Engineering ,Tritium ,Blanket ,Fusion power ,Molten salt ,Solubility ,Reactor design ,Corrosion - Published
- 1985
23. Mechanical and Thermal Design Aspects of the Blanket, and Maintenance Considerations for the Central Cell in MARS
- Author
-
Dai-Kai Sze, I.N. Sviatoslavsky, and Y. T. Li
- Subjects
Materials science ,Nuclear engineering ,Energy transfer ,Thermal ,Heat transfer ,General Engineering ,Mars Exploration Program ,Blanket - Abstract
This paper describes the mechanical and thermal design features of the MARS (Mirror Advanced Reactor Study) blanket and presents a credible concept for maintaining the central cell components of the reactor.
- Published
- 1983
24. Research in microsphere fabrication and coolant loop technology for lithium oxide moving bed fusion power plant
- Author
-
T.A. Thornton, D.C. Schluderberg, F.A. Zenz, and Dai-Kai Sze
- Subjects
Nuclear and High Energy Physics ,Materials science ,Nuclear engineering ,Magnetic confinement fusion ,Blanket ,Fusion power ,Coolant ,Corrosion ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,Particle ,General Materials Science ,Lithium oxide ,Inertial confinement fusion ,Nuclear chemistry - Abstract
A gravity-flow bed of lithium oxide microspheres has been proposed as the tritium breeding and blanket cooling medium for both magnetic confinement and inertial confinement controlled fusion reactors using the D-T cycle. The use of the lithium oxide moving bed could help eliminate MHD, corrosion, and materials compatibility problems associated with the use of liquid lithium. Adequate particle transport mechanisms which alleviate particle attrition and erosion seem attainable. Moving bed heat transfer coefficients an order of magnitude greater than for fluidized beds are indicat, ed. The fabrication of microspheres is complicated by the aggressive caustic nature of the lithium oxide but should be possible for commercial-scale production.
- Published
- 1979
25. Helium-Cooled Lithium Compound Suspension Blanket Concept for ITER
- Author
-
Saurin Majumdar, Carl E. Johnson, R.C. Clemmer, H.C. Stevens, Ahmed Hassanein, Richard F. Mattas, L.R. Turner, Michael C. Billone, D.L. Smith, Yousry Gohar, Patricia A. Finn, Dai-Kai Sze, H. Attaya, and Charles C. Baker
- Subjects
Alkaline earth metal ,Materials science ,020209 energy ,Nuclear engineering ,General Engineering ,chemistry.chemical_element ,02 engineering and technology ,Blanket ,Fusion power ,01 natural sciences ,010305 fluids & plasmas ,Coolant ,chemistry ,0103 physical sciences ,Heat exchanger ,0202 electrical engineering, electronic engineering, information engineering ,Tritium ,Beryllium ,Helium - Abstract
This blanket concept uses a dilute suspension of fine solid breeder particles (Li/sub 2/O, LiAlO/sub 2/, or Li/sub 4/SiO/sub 4/) in a carrier gas (He) as the coolant and the tritium breeding stream. A small fraction of this stream is processed outside the reactor for tritium recovery. The blanket consists of a beryllium multiplier and carbon/steel reflector. A steel clad is used for all materials. A carbon reflector is employed to reduce the beryllium thickness used in the blanket for a specific tritium breeding ratio. The breeder particle size has to exceed few microns (greater than or equal to2 microns) to avoid sticking problems on the cold surfaces of the heat exchanger. The helium gas pressure is in the range of 2 to 3 MPa to carry the blanket and the heat exchanger loop. The solid breeder concentration in the helium stream is 1 to 5 volume percent. A high lithium-6 enrichment is used to produce a high tritium breeding ratio and to reduce the breeder concentration in the helium gas. At a lithium-6 enrichment of 90%, the local tritium breeding ratio is 2.03 based on a one-dimensional poloidal model. The total thickness of the helium stream is only 4more » cm out of the 50 cm total blanket thickness. The blanket uses a 35 cm of beryllium for neutron multiplication. A simple multi-layer design is employed where the blanket sector has the helium coolant flowing in the poloidal direction. The blanket concept has several unique advantages which are very beneficial for fusion reactors including ITER. 10 refs., 2 tabs.« less
- Published
- 1989
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