Bromberg, L.

Cohn, D.R.

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PFC/RR-81-16USE OF HIGH PERFORMANCE RESISTIVEMAGNET TOKAMAKS ASADVANCED TEST REACTORS ANDFISSILE FUEL BREEDERSL. Bromberg and D. R. CohnApril 1981Use of High Performance Resistive Magnet Tokamaks as AdvancedTest Reactors and Fissile Fuel BreederstL. Bromberg and D. R. CohnPlasma Fusion CenterMassachusetts Institute of TechnologyCambridge, Mass.AbstractTokamaks with resistive toroidal field magnets operatedsteady state at relatively high stress and magnetic field levelscould provide very high values of nTe and fusion power density.Illustrative machine parameters are given for an advanced testreactor which has these characteristics. This device could pro-vide ignited long pulse DT plasma operation with large physicsmargins and could produce high wall loadings for possible use asa materials test facility. It could be operated with Q > 1 overa wide range of tritium-assisted semi-catalyzed deuterium (SCD-T)fuel mixtures. Somewhat larger high performance resistive magnettokamaks might be used as fissile fuel production reactors witheither SCD-T or DT operation. SCD-T operation could result inhigh (fissile fuel production rate)/(blanket thermal power)ratios, making possible the support of a large number of fissionreactors by a fusion reactor with moderate thermal power production.t Work supported by U.S.D.O.E.-2-I. INTRODUCTIONHigh performance tokamaks with resistive toroidal field(TF) magnets operated steady state at relatively high stressesand magnetic field levels can be used as advanced test reactors.Somewhat larger devices could be employed as fissile fuel pro-duction reactors. In this paper we discuss general features ofthese high performance resistive magnet tokamaks and describeillustrative parameters for an advanced test reactor.Relative to superconducting TF magnets the advantages ofhigh performance resistive TF magnets are:e Ease of operation at higher magnetic field, making possiblemuch higher plasma performance in terms of fusion powerdensity and the confinement parameter nT e The highervalues of fusion power density and nT e can be used tofacilitate advanced fuel cycle operation; tritium-assisted,semi-catalyzed deuterium (SCD-T) operation1 is of particularinterest.* Reduced magnet shielding requirements, allowing for a morecompact design.* Reduced complexity.* Improved maintainability.The high performance tokamak reactor design conceptsdescribed here represent an extrapolation of the ZEPHYR ' andCITR4, 5 pulsed ignition test reactor design concepts to larger-3-size and steady state, water cooled operation. These conceptsalso build directly upon the TORFA design, which utilized steadystate resistive TF magnets and placed emphasis on relatively6near term applications of fusion neutrons, and upon relatedmagnet design studies.7 Although the size of the plasma in theadvanced test reactor concept would be similar to that in TORFA,the magnetic field strength, the fusion power density, and nT ewould be much larger. It should be noted that a high performanceTF magnet can also be used to increase the fusion power/magnetresistive power loss ratio and to provide a more compact designfor constant values of nT e and fusion power density.II. SCD-T OPERATIONIn semi-catalyzed deuterium (SCD) operation, the tritiumproduced by the D(D,p)T reaction is assumed to be completelyburned up by the DT reaction. The He3 generated in the D(D,n)He3reaction, which is equally probable, leaves the plasma beforeit is burned. In tritium-assisted semi-catalyzed (SCD-T) opera-tion extra tritium is provided by an external source, such as atritium breeding blanket. The tritium breeding ratio require-ment (tritons produced external to the plasma/fusion neutrons)is nonzero but less than one. SCD-T operation provides a rangeof tradeoffs between plasma performance requirements and tritiumbreeding requirements. SCD-T operation is useful in fusionreactor operation for a number of reasons:* Since requirements on neutron economy can be reduced, thenumber of blanket design options in an electricity producing-4-fusion reactor can be significantly increased. Blanketscan be more readily optimized for safety, low activation,low tritium inventory, ease of maintenance and reduced size." SCD-T operation can be used to provide makeup tritium forfusion reactors with tritium breeding ratios which are lessthan one." The tritium burn-up fraction is higher than that in DToperation.* The availability of neutrons for nonelectrical applications,such as fissile fuel breeding and synfuels production, canbe significantly increased. This increased availabilitycould result in very high (fissile fuel production rate)/(blanket thermal power) ratios, making possible the supportof a large number of fission reactors with a fusion reactorwith moderate thermal power production.III. ILLUSTRATIVE ADVANCED TEST REACTOR PARAMETERSThe goals of an advanced fusion test reactor would include:" Long pulse, ignited DT operation with large physics margins." Q > 1 operation over a wide range of SCD-T fuel mixtures.* Use as a testbed for blankets and possibly as a high fluencematerials test facility.A parametric code was used to obtain illustrative para-meters based upon an extrapolation of the engineering design ofZEPHYR with Bitter type magnets 3; an engineering design has-5-not yet been performed. These parameters are also consistentwith parameters developed for TORFA devices operated at lowermagnetic fields.6, 7 The parameters are given in Table 1. Themajor radius is 4.1 m and the maximum magnetic field on axis is6.8 T. The blanket/shield thickness would vary depending uponthe mode of operation of the machine. For DT operation the in-board blanket shield thickness would be 55 cm and the outboardmaximum blanket/shield thickness would be 85 cm. The blanket/shield would be designed to reduce the integrated neutron fluenceon the magnet insulation to 1020 nt/cm 2; studies of Bitter typemagnet insulation failure indicate that such fluences may betolerable.3 For SCD-T operation, the blanket/shield thicknesswould be reduced to increase the plasma size and < > t (assumingthat < >t ~ 1/aspect ratio); the neutron flux, due to a lowerfusion power density, would be reduced.The resistive power, stored energy and stress in thetoroidal field magnet vary with Bt (axial toroidal magneticfield) as B .The illustrative test reactor parameters are based upona water cooled, copper-stainless steel magnet design. Reductionin resistive power requirements might be obtained by operation cfresistive magnets with helium cooling. 8There are three main possibilities for fulfilling main-tenance requirements for the Test Reactor:* The entire magnet could be modularized so that separatemodules might be removed. 9'1 0 The removal of a modulewould be required for most repairs and maintenance. The-6-estimated time lapse required for removal and installationof a new module is 2-4 weeks, depending on the level of auto-matization of the process.10e The magnet could be made up of discrete TF coils, assembledfrom Bitter plates. The TF coils would provide sufficientaccess between outer legs to remove a module consisting ofblanket, shielding and first wall. In order to provide thenecessary access, the size of the magnet, the resistivepower and the stored energy would increase by 50%, andthe stresses in the throat would increase by 30%. Thetime required for major maintenance of the tokamak may notbe reduced from the previous approach.* The same type of configuration as that described above canbe used with the additional characteristics that the coilsmay be demountable as in TORFA.6 This approach is attrac-tive in the case that there are internal poloidal fieldcoils or a poloidal divertor. For high performance magnets,however, it is not clear whether access to the blanket,shield and first wall is eased by this approach.Engineering design efforts will be required to examine the fea-sibility of these different approaches. Irrespective of thepreferred maintenance approach, the resistive TF magnet hasadvantages over a superconducting magnet in terms of remotemaintenance, arising from the absence of cryogenic environment.The OH transformer would be designed to provide sufficientdrive for 200 seconds of operation at maximum field and plasmacurrent.-7-Pump limiters would be used for impurity control andhelium ash removal.Table II lists possible modes of DT operation. In orderto make some projection about the energy confinement time, Te , it2eis assumed that T scales as T e Cna2 where n is the averageplasma density, a is the minor radius, and the proportionalityfactor C is determined by results from PLT. A margin of ignition,MI, is defined as:MI = (nT )/(nre)ignwhere (ne )ign is the value of nTe required for ignition. Para-bolic density and temperature profiles are assumed. It is assumedthat power loss from impurity radiation is negligible.The average fusion power density Pfl, the neutron wallloading Pw, and the total fusion power Pt are proportional toMI. For MI > 1, some additional loss mechanism such as impuri-ties or ripple transport loss must be introduced in order toobtain thermal equilibrium.It is assumed that the plasma is elongated and that anaverage value of toroidal beta of 0.064 can be maintained. Thedensity averaged values of the ion and electron temperaturesare <T.> = 13 keV and <T > = 11 keV.i n e nA power multiplier Qm can be defined as:Qm t mwhere Pm is the steady state electrical power required for thetoroidal field magnet.For MI = 1.3, a neutron wall loading of 2 MW/m and a-8-value of Qm of 1.8 can be obtained with a toroidal field magnetpower requirement of 260 MW.Very high neutron wall loadings (_ 10 MW/m2 ) and powerproduction levels can be obtained by operating at 6.5 T, some-what less than the maximum magnetic field. The use of these highwall loadings could be very useful if the device were operated asa materials test reactor. The design of the vacuum vessel andthermal hydraulics system must be evaluated in order to determinewhether such high wall loadings could be tolerated.The test reactor would be heated with ICRF at the secondharmonic of deuterium. Assuming a parabolic power depositionprofile, approximately 40 MW would be needed to heat to DTignition in approximately 3 seconds. It is estimated that themachine could be designed to provide adequate access for up to120 MW of auxiliary heating power.Table 3 shows possible modes of SCD-T operation forillustrative machine parameters at maximum magnet field. Thesemodes of operation are based upon the.assumptions that the T na2scaling holds. Table 3 lists tritium to deuterium ratios in theplasma, NT/ND' and the required external tritium productionrequirements per fusion neutron. The tritium production require-ment is determined by the breeding ratios y DT and yDD; theseparameters represent the ratios of tritons from the externalsource to DT and DD neutrons produced in the plasma. Forillustrative purposes it will be assumed that yDD = YDT =y ranges from 0 in SCD operation to one in DT operation.-9-For fixed values of nT <T> < t>, and Bt, there is acontinuum of tradeoffs between y, Q (fusion power/heatingpower), and P The auxiliary heating power requirements, Paux'represents the minimum startup power.Improvement in SCD-T performance can,of course, be obtainedby increasing the test reactor size. Increased performance mayalso be realized by refinements in the magnet design.IV. FISSILE FUEL PRODUCTION REACTORSThe use of high performance resistive magnet tokamaksas practical fissile fuel production reactors would generallyrequire machine parameters which would be somewhat larger thanthe illustrative test reactor parameters. More space would beallocated for blankets and shielding.Operation of these tokamak devices with the SCD-T fuelcycle could lead to very high (fissile fuel production rate)/(blanket thermal power) ratios. At these very high ratios avery large number of fission reactors might be supported by afusion reactor with moderate thermal power production. The over-all system cost (fission reactors plus the fusion reactor whichsupports them) might then be quite insensitive to the capitalcost of the fusion reactor. In addition, the net electricalpower of the overall system (electrical power produced by thefission reactors minus the electrical power required by fusionreactor) might be relatively insensitive to the electrical require-ments of fusion reactor; the fusion breeder could be run in a-10-mode where it did not produce any of the electricity required forits operation. The use of tokamak reactors with SCD-T operationfor fissile fuel production is relatively sensitive to the valuesof < t> that can be achieved.The use of DT operation for fissile fuel breeding reducesfusion breeder electrical requirements, decreases the sensi-tivity to < t> and could lead to smaller machine designs. How-ever, the attainment of moderately high (fissile fuel productionrate)/(blanket thermal power production) ratios becomes verysensitive to blanket design.V. CONCLUSIONSA high performance resistive magnet tokamak could be usedas a test reactor which would provide a wide range of physics andtechnology data needed for fusion reactor development. Some-what larger devices could be used as fissile fuel productionreactors with either DT or SCD-T operation. SCD-T operationcould facilitate the attainment of high (fissile fuel productionrate)/( blanket thermal power) ratios. High performance resistivemagnet tokamaks might also be employed for other nonelectricalfusion applications, such as synfuel production.-11-REFERENCES:1. Greenspan, E., Blue, T., and Miley, G.H., Proc. of ANSTopical Meeting on Technology of Controlled Dusion, Kingof Prussia, PA (October 1980).2. ZEPHYR Design Report, Max Planck Institut fur Plasmaphysik,1980.3. J. Williams et.al., Bitter Magnet Design for ZEPHYR, tobe published as an MIT Plasma Fusion Center Report.4. Bromberg, L., Cohn, D.R., Williams, J.E.C. and Becker, H.,"Compact Ignition Test Reactor (CITR)", Plasma Fusion CenterReport PFC/RR-80-23 (December 1980) (to be published in theJournal of Fusion Energy).5. Bromberg, L. and Cohn, D.R., MIT Plasma Fusion Center ReportPFC/RR-81-12 (March 1981).6. Jassby, D.L., TORFA, Princeton Plasma Physics LaboratoryReport PPPL-1700.7. J. Kalnavarns and D.L. Jassby, Proc. of 8th Symposium onEng. Problems of Fusion Research, San Francisco, CA 1979,148-153; J. Kalnavarns and D.L. Jassby, Princeton PlasmaLaboratory Report PPPL 1685, 1980.8. Yang, T., et.al., MIT Plasma Fusion Center ReportPFC/RR-80-26 (1980).9. Cohn, D.R., et.al., HFCTR Final Design, MIT Plasma FusionDenter Report PFC/RR-78-4.-12-10. Westinghouse Electric Corp., Fusion Power Division, "FusionReactor Remote Maintenance Study", Electric Power ResearchInstitute Report ER-1046 (April 1979).11. Bromberg, L. and Cohn, D.R., Proc. of ANS Topical Conf. onTechnology of Controlled Fusion, King of Prussia, PA(October 1980) .-13-TABLE 1ILLUSTRATIVE TEST REACTOR PARAMETERSPlasma major radius:Plasma minor radius:Plasma elongation:Inboard blanket/shield thickness:Outboard blanket/shield thickness:TF magnet plate dimensions:Maximum magnetic field on plasma axis:Maximum magnetic field at TF wagnet;TF Tnagnet current density at max field:Electrical power dissipation inmagnet at max field (copper/stainlesssteel ratio = 2:1)TF magnet stored energy at max field:Max TF magnet stress at max mag field(in copper plates):DT4.1 m1.25 m1.50.55 m0.806 m x 7.7 m6.8 T13.6 T. 2 kA/cm 2650 MW12140GJMPASCD-T4.1 m1.45 m1.50.35 m0.606 m x 7.7 m6.8 T13.6 T2 kA/cm2650 MW12 GJ140 MPAI-14-TABLE 2DT MODES OF OPERATION FOR ILLUSTRATIVE TEST REACTOR PARAMETERS<Ti>n = 13 keV a = 1.25 mMI = 1.36.5 T2300 MW10 MW/m 2600 MW4.2 T460 MW2 MW/n 2260 MW1.8< > = 0.064MI = 6.5BtPtPwPmmn3.8-15-TABLE 3SCD-T MODES OF OPERATION FOR ILLUSTRATIVE TEST REACTOR PARAMETERSt= 0.074Bt = 6.8 Ti n= 17 keVnT = 5.1 x 10 20m-3a = 1.45 mP = 650 MW0 f 3 w 2p (MW/m ) (MW/m00 3.25. 1.53. 1.12.41.10.8120 1. 0.5 0.4Y0.030.0110.007Paux(MW)5075900.930.640.53Pt(MW)8303)02901 30.600.450.0006 0.09 130 0.20

Use of high performance resistive magnet tokamaks as advanced test reactors and fissile fuel breeders

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Use of high performance resistive magnet tokamaks as advanced test reactors and fissile fuel breeders

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