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Thermal design and analysis of steady state graphite blankets

Description: A description is given of the graphite blanket design for either an experimental power reactor or commercial reactor. A thermal analysis was made of the blanket. It is shown that the introduction of a low conductivity surface layer permits one to go to higher first wall loadings for commercial reactors. (MOW)
Date: January 1, 1975
Creator: Fillo, J.A. & Powell, J.R.
Partner: UNT Libraries Government Documents Department

Low activity blankets for experimental power reactors

Description: Results of current studies aimed at the development of low activity blankets for Tokamak experimental power reactors are presented. First wall loadings in the range of 0.5 to 1.0 MW(th)/m$sup 2$ have been assumed. Blanket designs are developed for both circular plasma reactors (R = 6.25m, a = 2.1m) and non-circular plasma reactors (R = 4.0m, a = 1.0m, b = 3.0m). For each of these two reactor choices, two blanket options are described. 1) In the first option, the blanket is thick graphite block structure (approximately 50cm thickness) with SAP coolant tubes carrying helium imbedded deep within the graphite to minimize radiation damage. The neutron and gamma energy deposited in the graphite is radiated along internal slots to the coolant tubes where approximately 80 percent of the fusion energy is carried off by He at 380$sup 0$C. The remaining 20 percent of the fusion energy is removed by a separate He stream at a slightly lower temperature. The maximum graphite surface temperature is relatively low (approximately 1700$sup 0$C at 1 MW(th)/m2). 2) In the second blanket option, the blanket is composed of aluminum modules. The aluminum shell (5000 series alloy) is maintained at a low temperature (approximately 200$sup 0$C) by a water coolant stream. Approximately 40 percent of the fusion energy is removed in this circuit. The remaining 60 percent of the fusion energy is deposited in a thermally insulated hot interior (SiC and B$sub 4$C) where it is transferred to a separate He coolant, with exit temperature of 700$sup 0$C. (auth)
Date: January 1, 1975
Creator: Benenati, R.; Fillo, J.; Lazareth, O.W.; Majeski, S.; Powell, J.R. & Tichler, P.
Partner: UNT Libraries Government Documents Department

Nonsteady heat conduction code with radiation boundary conditions

Description: A heat-transfer model for studying the temperature build-up in graphite blankets for fusion reactors is presented. In essence, the computer code developed is for two-dimensional, nonsteady heat conduction in heterogeneous, anisotropic solids with nonuniform internal heating. Thermal radiation as well as bremsstrahlung radiation boundary conditions are included. Numerical calculations are performed for two design options by varying the wall loading, bremsstrahlung, surface layer thickness and thermal conductivity, blanket dimensions, time step and grid size. (auth)
Date: January 1, 1975
Creator: Fillo, J.A.; Benenati, R. & Powell, J.
Partner: UNT Libraries Government Documents Department

Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report, 1 August 1974--30 April 1975

Description: A brief report of research progress is given for the following areas: (1) completion of UWMAK-I ''balance of plant'' design, (2) UWMAK-II reactor design, (3) the carbon curtain concept, (4) the Internal Spectral Shifter and Energy Converter (ISSEC) blanket design concept as a method to minimize the radiation damage to CTR first walls from D-T neutrons, (5) beam driven tokamak studies, (6) plasma wall experiments, (7) neutron cross section measurements for the production of long lived radioisotopes, H, and He atoms, and (8) experimental testing of magnet conductor design. (MOW)
Date: January 1, 1975
Creator: Conn, R.W.; Kulcinski, G.L. & Maynard, C.W.
Partner: UNT Libraries Government Documents Department

Optimizing the mirror (fusion--fission) hybrid reactor for plutonium production

Description: An analytic model of the fusion components is used to generate a consistent set of fusion parameters, and component costs as parameters are varied. A model of the blanket, based on neutronic and thermal hydraulics, is then used to analyze the trade-offs of energy production vs plutonium production dictated by blanket type and management. An economic discussion of fuel cost is also given. (MOW)
Date: November 17, 1975
Creator: Lee, J.D.; Bender, D.J. & Moir, R.W.
Partner: UNT Libraries Government Documents Department

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

Description: The many possible combinations of blanket materials, tritium generation and recovery systems, and power conversion systems were surveyed and a comprehensive set of designs were generated by using a common set of ground rules that include all of the boundary conditions that could be envisioned for a full- scale commercial fusion power plant. Particular attention was given to the effects of blanket temperature on power plant cycle efficiency and economics, the interdependence of the thermodynamic cycle and the tritium recovery system, and to thermal and pressure stresses in the blanket structure. The results indicate that, of the wide variety of systems that have been considered, the most promising employs lithium recirculated in a closed loop within a niobium blanket structure and cooled with boiling potassium or cesium. This approach gives the simplest and lowest cost tritium recovery system, the lowest pressure and thermal stresses, the simplest structure with the lowest probability of a leak, the greatest resistance to damage from a plasma energy dump, and the lowest rate of plasma contamination by either outgassing or sputtering. The only other blanket materials combination that appears fairly likely to give a satisfactory tritium generation and recovery system is a lithium-beryllium fluoride-Incoloy blanket, and even this system involves major uncertainties in the effectiveness, size, and cost of the tritium recovery system. Further, the Li$sub 2$BeF$sub 4$ blanket system has the disadvantage that the world reserves of beryllium are too limited to support a full-blown fusion reactor economy, its poor thermal conductivity leads to cooling difficulties and a requirement for a complex structure with intricate cooling passages, and this inherently leads to an expansive blanket with a relatively high probability of leaks. The other blanket materials combinations yield even less attractive systems. (auth)
Date: January 1, 1975
Creator: Fraas, A.P.
Partner: UNT Libraries Government Documents Department

Chemical equilibrium studies of tritium--lithium and tritium--lithium alloy systems

Description: In deuterium-tritium fusion reactors currently under design, the production of tritium is accomplished by utilizing a lithium-bearing blanket. Lithium metal is presently the leading candidate for the blanket material, although molten Li$sub 2$BeF$sub 4$, solid Li--Al (50-50 at. percent) alloy and other lithium-containing materials are distinct possibilities. This paper summarizes progress of ongoing studies of the thermodynamics of some of these lithium containing systems. The individual solubilities of hydrogen, deuterium, and tritium in lithium as a function of temperature (700 to 1000$sup 0$C) and pressure are presented. Recent work with the solid alloy Li--Al (50-50 at. percent) has shown that the tritium solubility between 400 and 600$sup 0$C is low. When the tritium pressure was between 0.14 and 0.52 torr, the Li--Al samples contained only 1 to 4 ppm tritium. (auth)
Date: January 1, 1975
Creator: Smith, F.J.; Land, J.F.; Talbot, J.B. & Bell, J.T.
Partner: UNT Libraries Government Documents Department

Design of an EPR blanket

Description: A blanket concept is presented which meets typical requirements anticipated for an Experimental Power Reactor. Design alternatives are reviewed. One-dimensional neutronic and thermal hydraulic results are presented for the ORNL reference design. Design consideration was given for remote maintenance and assembly requirements. Modifications of the reference design first wall are necessary because of high thermal stresses. (auth)
Date: January 1, 1975
Creator: Bettis, E.S.; Huxford, T.J.; McAlees, D.G.; Santoro, R.T.; Watts, H.L. & Williams, M.L.
Partner: UNT Libraries Government Documents Department