the design and performance features of a single-fluid molten salt breeder reactor

THE DESIGN AND PERFORMANCEFEATURES OF A SINGLE-FLUIDMOLTEN-SALT BREEDER REACTORE. S. BETTIS and ROY C. ROBERTSON Reactor Diuisi.onOak Ridge National Laboratory, Oak RMge, Tennessee gZgS0

Received August 4, 1969Revised October Z, 1969

A conce\hnl design has been mad,e of a single-fluid 1000 MW(e) Molten-salt Breeder Reactor(I\ISBR) pouter station based on the capabiti.ties ofpresent technology. The reactor aessel is ,-,22 ftin diameter x 20 ft high and is fabricated bfHastelloy -N with graphite as the moderator andreflectar. The fuel is 233(I carried in a LiF-BeFz-ThFt mixhtre uthi.ch is molten aboue gsT"F. Th;-rium is conaerted to 233 ry in exces s of fissilebu'rrntp so that bred materiar is a ptant product.The estimated fuel yi,etd i.s J.S% \er year.

The estimated constructi.on cost of the stationis comparable to pwR totar construction cosfs.The power produ,cti.on cost, inctuding fuel-cycleand graphite replacement costs, w i th priuateutility financing, is estimated to be 0.s to r miL/hwh less than that for present-day light-waterreactors, largely due to the low fuet-cycle costand high plant thermal efficiency.

After engineering deuelopment of the fuet fuiri-fication processes and large-scale components, apractical plant simi.lar to the one described hereappears to be feasible "

INTRODUCTION

The objective of this design study is to investi-gate the f easibility of attaining low- cost electricpower, a low specific inventory of fissile rmaterial,and a reasonably high breeding gain in a molten-salt reactor. As discussed by perry and Baumanra moltensalt reactor can be designed as either abreed€r, or, with relatively few major differencesother than in fuel processing, ?s a converter. Thisparticular design study is confined to the single-fluid Molten- salt Breeder Reactor (MSBR) .

K EYWORDS: molten-solt re-ocfors, design, performonce,econom ics, urtrr ium-233, pew-er reocfors, fuels, MSBR, cosf,fuel cycle

The conceptual design of the MSBR plant wasmade on the basis that the technology required forfabrication, installation, and maintenance be gen-erally within present-day capabilities. Majofae-sign considerations were keeping the f uel- saltinventory low , accomodating graphite dimensionalchanges, selection of conditions favoring graphitelife, and providing for maintainability.

The facilities at an MSBR power station can begrouped into the following broad categories: (a)the reactor system which generates fission heat ina fuel salt circulated through primary heat ex-changers; (b) an off- gas system for purging thefuel salt of fission-product gases and colbidalnoble metal particulates; (c) a chemical proces-sing facility for continuousry removing fissionproducts from the fuel salt, reeovery of the bred"tIJ, and replenishment of fertile material; (d) astorage tank for the fuel salt which has an after-heat removal system of assured realiability; (e) acoolant-salt circulating system, steam generators,and a turbine- generator plant for converting thethermal energr into electric power; and (f) gen-eral facilities at the site which include condettiingwater works , electrical switchyard, stacks , con-ventional buildings, and services. These cate-gories are not always clearly defined and areclosely interdependent, but it is convenient todiscuss them separately. The reactor, and itsrelated structures and maintenance system, thedrain tank, the off- gas system, and the chemicalprocessing equipment, are of primary interest.The steam turbine plant and the general facilitiesare more or less conventional and will be dis-cussed only to the extent necessary to completethe overall picture as to feasibility and costs of anMSBR station.

In a single-fluid MSBR the nucrear fuel is 233u

(or other fissile material) carried in a lithium -7 -fluoride, beryllium-fluoride, thorium-fluoride salt.The mixture is fluid above - 930"F and has goodflow and heat transfer properties and very low

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t90 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O

vapor pressure. This salt is pumped in a closedIoop through a graphite-moderated and reflectedcore where it is heated to - 1300iF by fissioningof the fuel. It then flows through heat exchangerswhere the heat is transferred to a circulating heattransport salt. This fluid, in turn, supplies heat tosteam generators and reheaters to power a con-ventional high-temperature, high-pressure steamturbine-generator plant. The thorium in the flu-oride fuel-salt mixture is converted to "tU at a

rate in excess of the fuel burnup so that fissilematerial, as well as power, iS a valuable productof the plant.

A low specific inventory is obtained by design-ing the MSBR to operate at a high reactor powerdensity with a minimum of fuel salt in the circu-Iating system. The concentration of fissile-fertilematerial in the salt results from a compromisebetween keeping the inventory low and achieving a

higher breeding gain. High performance depends

on keeping the neutron parasitic absorptions lowand fuel losses to a minimum. The Li and Be inthe fuel salt are good neutron moderators, buttheir concentrations are relatively low in fluoridesalts anri additional moderation is needed. Graph-ite is the most satisfactory moderator and re-flector for the MSBR. However, radiation affectsthe graphite and its useful life varies nearly in-versely as the maximum power density in thecore. The radiation damage eff ect is also in-creased by higher temperatures. Selection of a

power density is thus a balance between a low fuelinventory and the frequency with which the graph-ite must be replaced.

The optimization studies which equate the sev-eral factors mentioned above are described by

Scott and Eathe rby." These studies indicate thatan average core power density of 22.2 kflliterresults in a useful graphite life of F' 4 years, whichis the operating condition used in this designstudy. Thus, the reactor design must provide forperiodic replacement of the core graphite withminimal plant downtime and complexity of mainte-nance equipment.

To attain a high nuclear performance it isnecessary to maintain Iow concentrations of Pa

and 135Xe in the high flux region of the core.The protactinium is kept low by processing a

small side stream of the salt for removal of thisnuclide and other f ission products. By integralprocessing on-site a minimum inventory of fuelsalt is involved in transport and storage. The

xenon is removed by helium- sparging the fuel salton a few-second cycle; the core graphite is alsosealed to decrease the rate of diffusion of the Xe

into the pores of the graphite. The plant designtheref ore includes the auxiliary systems to re-move the nuclear poisons and the bred fissile ma-

Bettis and Robertson MSBR DESIGN AND PERFORMANCE

terial from the salt, to purge the fission-productgases, and to store or dispose of the radioactivewaste products; Whatley et &I.,t and Perry and

Baumant treat the subiect of salt processing andestimate fuel-cycle costs.

Although the economic performance of a 2000

IWW(e), or larger, MSBR station would be signifi-cantly better than that of smaller sizes, a plantwith a net electrical output of 1000 ll[W(e) waschosen for t}is study because it permitted moredirect comparison with the results of otherstudies. A steam-power cycle with 1000:F and3 500-psia turbine throttle conditions, with singlereheat, was selected because this was representa-tive of current practice and because it afforded ahigh overall plant thermal efficiency of - 44V0.

The MSBR is adaptable to other steam conditionsand to additional reheats if future developmentsIead in this direction.

Att portions of the systems in contact withsalts are fabricated of Hastelloy-N. When thismaterial is modified with ^" IVo tltanium or hafniumto improve resistance to radiation embrittlement,as described by McCoy et al.,a it has good high-temperature strength and excellent resistance tocorrosion. No exothermic reactions of concernresult from mixing of the fuel and coolant saltswith each other or with air or water. It is im-portant to keep water and oxygen out of the saltsin normal operation, however. The reactor graph-ite is a specially developed type having very lowsalt and gas permeability and good resistance toradiation damage. The salt composition is a com-promise between the nuclear and physical proper-ties, chemical stabilitY, etc., &S discussed byGrimes.s Some selected physical properties of thematerials important to the design study are shownin Tables I, II, and III.

The performance features of the plant, in termsof breeding gain, graphite life, thermal efficiency,and the net cost to produce electric power, are allreported here on the basis of normal f ull-Ioadoperation at 80Vo plant factor. Very preliminaryreview of the various modes of start-up and shut-down, partial-Ioad operation, etc., has not dis-closed any maior problem areas, however-

PLANT DESCRIPTIOTU

A simplified flow diagram of the primary and

secondarysalt circulating systems is shown inFig. 1. The fluoride fuel-salt mixture is circulatedthrough the reactor core by four pumps operatingin parallel." Each pump has a capacity of - L4 000

" Flow sheet does not show duplicate equipment.

r9tNUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970


TABLE I

Properties of the Primary and Seconda.ry Salts Used inConceptual Design Study of MSBR 1000-MW(e) Station

PrimarySalt

SecondarySalt

Components

Composition , mole%o

Molecular weight,approximate

Liquidus temperature,OF

Density ,Ib/ft'Viscosity, lbl(ft h)

Thermal conductivity,Btu/(h ft "F)

Heat capacity,Btu/(tb 'F)

Vapor pressure at1150"F, Torr (mm Hg)

LiF- BeFz- ThF4- UF4

71.6- 16- 12- 0.4

64

930

205 (at 1300"F)

16.4 (at 1300"F)

0.7 to 0.8

0.32

<0.1

NaBFa- NaF

92- 8

104

725

117 (at 988'F)

2.5 (at 900'F)

0.27

0.36

252

gaL/min and circulates the salt through one of fourprimary heat exchangers and returns it to a com-mon plenum at the bottom of the reactor vessel.use of four pumps and heat exchangers corres-ponds to a pump size which represents a reason-able extrapolation of the Molten- Salt ReactorE>rperiment (MSRE) experience.

Each of the four coolant- salt circuits has apump of 22 000- gaL/min capacity which circulatesthe coolant salt through a primary heat exchangerlocated in the reactor cell and then through steamgenerating equipment instalted in an adjacent cell.The reactor can continue to operate although atreduced output, if not all of the coolant salt pumpsare operative.

A plan of the reactor plant is shown in Fig. 2;an isometric view and a sectional elevation arepresented in Figs. 3 and 4.

The reactor cell is ^-' 62 f.t in diameter and 3b ftdeep. It houses the reactor vessel, the four pri-mary heat exchangers, and four fuel- salt circu-lating pumps. The roof of the cell has removableplugs over all equipment which might requiremaintenance. The reactor cell is normally kept ata temperature of '-' 1000'F by erectric heaterthimbles to ensure that the salts will remainabove their liquidus temperatures. The estirnatedmaximum heating load is ,--'2b00 kw. This"furnace" concept for heating is preferred overtrace heating of lines and equipment because itensures more even heating, heater elements canbe replaced without reactor shutdown, there is noneed for space coolers inside the cell, and bulkythermal insulation that would crowd the cell andrequire removal for maintenance and inspection iseliminated.

TABLE IINominal Values for Properties of MSBR Graphite

Density ,Ib/ftt at room temperatureBending strength, psiYoung's modulus of elasticity, psiPoisson's ratioThermal expansion, per oF

Thermal conductivity, Btu/(hr ft 'F)( 1340'F)

Electrical resistivity, Q-cmSpecific heat, Btu/(Ib .F) at 600.F

Btu/(lb 'F) at 1200'F

1154000- 60001.7 x 106

0.272.3 x 10-6

35- 428.9-g.g x 10-n0.330.42

Biological shielding for the reactor cell is pro-vided by reinf orced concrete. The walls of thereactor cell also furnish double containment forthe fuel-salt systems. Two thicknesses of carbonsteel plate form inner and outer containment ves-sels, each designed for b0 psig, and also providegamma shielding for the concrete. Nitrogen gasflows between the plates in a closed circulatingloop to remove ^''3 vilM(th) of heat. Double bellowsare used at all cell penetrations. The normal celloperating pressure is - 26 in. Hg abs.

High-temperature thermal insuration is at-tached to the inside of the containment membraneto limit heat losses f rom the reactor cell. Theinside surface of the insulation is covered with athin stainless-steel liner to protect the insulationfrom damage, to act as a radiant heat reflector,

TABLE IIISelected Physical Properties of Hastelloy- N

'The exact composition may be different from the nominal valuesgiven here, as discussed by McCoy et al.a

Nominal Chemical Composition ofModified Alloy for Use in MSBRa wt%o

NickelMolybdenumChromiumIronTitaniumOther

75L2

7

41

1

Density , lb/ftsThermal conductivity, Btu/(h ft "F)

Specific heat, Btu/(lb 'F)Thermal expansion, per oF

Modulus of elasticity, psi

Electrical resistance, CFcm

Approximate tensile strength, psi

Maximum allowable design stress, psi

Maximum allowable design stress,bolts, psi

Melting temperature, oF

At gO"F At 1300"F

-553

6.0

0.098

5.? x 10-6

31 x 106

120.b x 1o- 6

115 000

25 000

10 000

-2 500

-55312.6

0.136

g.5 x 10-6

25 x 1oo

126.ox 1o-6

?5 000

3 500

3 500

-2 500

192 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. B FEBRUARY 19?O

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r93FEBRUARY 1970VOL. BNUCLEAR APPLICATIONS & TECHNOLOGY

Bettis and Robertson

-- 'H d;;Conloinrnenf

MSBR DESIGN AND PERFORMANCE

'$isoitemblla *ieo:." t.{Remofe} Ii :ir'

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,,s inj

iiil:llIi:::::::

ltit::,,..i:ij:

,$iii:l :.::,l::,,,',",t,,

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and to provide a clean, smooth surface for theinterior.

The stainles s - steel ' 'catch pan' ' at the bottomof the reactor cell, shown in Fig. 4, slopes to adrain line leading to the fuel- salt drain tank Io-cated in an adjacent cell. In the very unlikelyevent of a maj or salt spill, the salt would f low tothe tank. A valve is provided in the drain line toisolate the tank contents from the ceII during nor-mal conditions and to permit pressurizing thedrain tank for salt transfer.

The f our rectangular steam-generating cellsare located adjacent to the reactor cell. Thesehouse the secondarysalt circulating pumps, thesteam generator-superheaters, and the reheaters.The cell construction is similar to that of the re-actor cell but only a single containment barrier is

lnstrunn enf sf ion Ce l,l

essina't

required and heavy steel plate is not needed forshielding the concrete. The 4O-psig design pres-sure would accommodate leakage of steam into thecell.

other adjacent cells provide for storage of thefuel salt, for storage of spent reactor core as-sembly, and for other radioactive equipment re-moved for maintenance.

Through caref ul quality control of materialsand workmanship, the reactor loops containingfission products would have a high degree ofintegrity and reliability in preventing escape ofradioactive materials from the system. The cool-ant-salt system will operate at a slightly higherpressure than the fuel salt so that any leakage atheat exchanger joints would be in the direction ofthe fuel salt. The coolant-salt system would be

APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O

Drqln {,qpl{s ' ':

{Befow Sr,ound

b#;i'H;or'Exlro*di'B*ndle Sroqsg,b' Ce ll ' '

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il;c*r:lfl-,-,Prirnory. ','i",,,r,H6ot ',-"";'',i.,Exihongei'iF'

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Fig. 2. PIan view of MSBR cell complex for 1000 MW(e) power station.

194 NUCLEAR

provided with rupture disks to prevent steampressure from being transmitted to the fuel saltvia the coolant- salt system. In addition to thedouble containment around all fuel-salt equipment,mentioned previously, the building covering thereactor plant is in itseU a sealed structure to actas a confinement for airborne material.

REACTOR

The MSBR reactor core design is based on amaximum allowable neutron dose to the coregraphite of '- 3 x L02' n/cm' (for E > 50 keV). Theaverage core power density is ^" 22 kW/liter,


:::l:l:::.;::l:::::::t:::::l:l:.

affording a useful core graphite life of ^" 4 years,a fuel yield of - 3.3Vo per year, and a compoundedfuel doubling time of - 2t years. These and otherdata are given in Table IV. The dimensions of thereactor were obtained through nuclear physics op-timization studies discussed by Perry and Bau-*an. t

The coefficient of thermal expansion of Has-telloy-N is about three times that of the graphite.As a result, the clearances inside the reactorvessel increase signif icantly as the system israised from room temperature to operating tem-perature. It is necessary to keep the graphite in acompact array, however, to have control over the

Fig. 3. Cutaway perspective of MSBR reactor and steam cells. (1) Reactor, (2) Primary heat orchangers,(3) Fuel-salt zumps, (4) Coolant-salt pumps, (5) Steam generators, (6) Steam reheaters, (7) Fuel-saltdrain tank, (8) Containment structure, (9) Confinement building.

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 195


Control Rod Drive

Primaryfult Pump

SfeamGenerqlqr

Reheofer

f,t 6, in.

Fig. 4. sectional elevation of MSBR cell complo< for 1000 MW(e) power station.

Steomi cetti

Seconclarytulr PufiF

Steom

lipinb

TABLE IVPrinciple Reactor Design Data for MSBR

't

','l i

.:J

';{ Floor

fuel-to-graphite ratios, to ensure that the saltvelocities in the passages are as planned and toprevent vibration of the graphite. The fact that thegraphite has considerable buoyancy in the fuelsalt, and yet must be supported when the reactoris empty of salt, is another important design con-sideration.

Figures 5 and 6 shorv plan and elevation viewsof the reactor. The vessel is '.', 22 ft in diameterand 20 ft high, with 2-in.-thick Hastelloy-N walls.The dished heads at the top and bottom are 3 in.thick. The maximum design pressure is 75 psig.A cylindrical extension of the vessel above the tophead permits the reactor support flange and tophead closure to be located in a relatively low-temperature, low-activtty zone between two layersof roof shielding plugs. In this position the sealwelds and flange bolting are more accessible anddistortion of the flange due to high temperature iskept small. Double metal gaskets and a leak-detection system are used in the flanged closure.

The fuel salt enters the bottom of the reactorat ^" 1050oF, flows upward through the vessel, andleaves at the top at - 1300'F. The central core is*14 ft in diameter and consists of extruded graph-ite elements , 4 in. x 4 in. x 13 ft long. A *-in.-diam hole through the center of each element, andridges on each side of the elements to separatethe pieces, furnish flow passages and provide therequisite ISVo salt volume in this most active por-tion of the reactor. other shapes of graphite

i: ;i;"i : i t if#ll;l*fiffcotih,J' ' " '

,Bosin' ,,'1 .,. ,'

=*;;- 62 fi 0

"Based on SOVo plant factor

196

Useful heat generation, MW(th)Gross electrical generation, IVIW(e)Net electrical output of plant, MW(e)Overall plant thermal efficien cy, Vo

Reactor vessel i.d., ftVessel height at centerline, ftVessel wall thickness, inVessel head thickness, inVessel design pressure, psigVessel top flange opening diameter, ftCore height, ftDistance across flats of octagonal core, ftRadial blanket thickness, ftGraphite reflector thickness, ftNumber of core elementsSize of core elements, in.Satt-to-graphite ratio in core rVo of eore volumeSalt-to-graphite ratio in undermoderated region, Vo

Salt in reflector volume, 7o

Total weight of graphite in reactor, lbWeight of removable core assembty, lbMaximum flow velocity in core , ft/seeMaximunl graphite damage flux (>b0 keV),

n/(cmz sec)Graphite temperature in rnaximum flux region, oF

Average core power density, W/cmgMaximum thermal neutron flux, n/(cm'sec)Estimated graphite life, yearsaPressure drop through reactor due to flow, psiTotal salt volume in primary system, ft3Thorium inventory, kgFissile fuel inventory of reactor system and

processing plant, kgBreeding ratioYield, Vo/yearDoubling time, years

225010351000

44.422.220

2

3

75181314.3

1.32.5

L4t24x4x156

13

37<1

650 000480 000

8.5

3.2 x 1014

t28422.2

7.9 x 1014

418

l 72068 000

1 4681.063.3

2L

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O


-i';i;';,22 ft ,6+ in. O* d,r;;.-,*.,.:,-'

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j'

'iiliuul-

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'.t 'ti

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"iirrxi'l'.liii*'1;'".fi..,. ""... irlax.r;iu.=;t=ti1"rr'*;i '.'',."..'.,,.,o,rn.,,..iiir,,r

Top cut-away view of reactor vessel for MSBR 1000 MW(e) station.

ficing to obtain the desired flow distribution. By

uaryirrg the salt velocity f rom '-" I ft/ sec at the

center to '"' 2 ft/sec near the periphery' a uniformtemperature rise across the core of ^" 250'F isobtained. The overall pressure drop in the saltflowing through the core is - 26 psi.

Graphite slabs , - 2 X 10 in. in cross sectionand 13 ft long, are aryanged radially around the

active core. These slabs have larger flow pas-

sages, with '\' 37Vo of the volume being fuel salt.This undermoderated region serves to reduce the

elements can be considered, with perhaps the ex-

ception of cylinders, which would have poorlydefined flow passages between them. The graphiteprisms are maintained in a compact array, &S

shown in Figs. 5 and 6.Molded extensions on each end of the elements

provide a greater salt-to-graphite volume ratio atthe top and bottom of the main core to form an

undermoderated region which helps reduce the

axial neutron leakage from the core. The molded

extensions on the elements also provide the ori-


BEttiS ANd RObETISON MSBR DESIGN AND PERFORMANCE

Fig. 6. Sectional elevation of reactor vessel for MSBR station.

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radial neutron leakage. The slabs provide thestiffness to hold the inner core graphite elementsin a compact array as dimensional changes occurin the graphite. The slabs are confined by graph-ite rings at the top and bottom.

Eight equally spaced vertical holes, - 3 in. indiameter, are provided in the undermoderated re-

tion, mentioned above, to allow a small portion ofthe incoming fuel salt to bypass the core regionand flow through graphite passages in the top headreflector graphite to prevent overheating. Theseholes are indicated on Fig. b as t3lif ting rodports. "

since it is not expected that the core graphite

APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O198 NUCLEAR

would last the life of the plant, it is designed forperiodic replacement. It was decided to replaceall the core graphite as an assembly rather thanby individual pieces because it appeared that thismethod could be performed more quickly and withIess likelihood of escape of activity. Handling the

core as an assembly also permits the replacementcore to be caref uIIy preassembled and testedunder essentially shop conditions.

The portions of the core described above arearranged as a removable unit having an overalldiameter of - 16 ft and a total weight of 240 tons.When the reactor is empty of salt, the graphiterests on a Hastelloy-N bottom plate, which, irturn, rests on the bottom head of the vessel. (The

bottom reflector graphite is keyed to this supportplate and when the vessel is filled with salt the

*ass of the plate is suff icient to prevent the

bottom reflector from floating.) When it is neces-sary to replace the

- graphite assembly, eight

Hastetloy-N rods , - 2i in. in diameter, are in-serted through ports around the periphery of the

top head, do'wn through the lifting rods ports in the

graphite, mentioned above, and latched to the

bottom support plate. The rods are then attachedto the top head so that the entire core can be re-moved as an assembly. The reactor control rods

and the associated drive assemblies are also

removed with the head. The maintenance equip-ment and procedures are briefly described in alater section of this Paper.

The reactor vessel also contains a z.'-ft-thickIayer of radial reflector graphite at the outsidewall. This graphite receives a relatively low neu-

tron dose and does not require periodic replace-ment. The graphite is arranged in wedged- shaped

slabs , - 12 in. wide at the thicker end and - 4 fthigh. These slabs are spaced '-' i in. from the

vessel wall to allow an upward flow of fuel salt tocool the metal waII and maintain it below the

design temperature of '-" 1300:F. The slabs have a

radial cleaiance of '-'1| in. on the inside to accom-

modate dimensional changes of the graphite and to

allow clearances in removing and replacing the

core assembly. Salt flow passages and appropri-ate orificing are provided between the graphite

slabs to maintain the temperature at acceptableIevels. A system of Hastelloy-N bands and verti-cal rods key the slabs together and cause the

reflector to move with the vessel wall as the

systems expand with temperature. Reflectorgraphite is also provided in the top and bottom

heads. The amount of fuel salt in the radial and

axial reflector regions is - l7o of the reflectorvolume.

Design concepts for a single-fluid breeder re-actor diff erent f rom that described above areclearly possible. Relatively little optimization


work has been completed to date. Hydrodynamictesting will be an essential part of establishing afinal MSBR reactor design.

PRIMARY HEAT EXCHANGERS

Four counterflow, vertical, shell-and-tube heat

exchangers are used to transfer heat from the fuelsalt to the sodium fluoroborate coolant salt. Per-tinent data for the exchangers are given in Ta-ble V.

As shown in Fig. 7 , f uel salt enters the top ofeach unit and exlts at the bottom after once-

through flow through the tubes. The coolant saltenters the shell near the top, flows to the bottomin a z}-in.-diam downcomer, turns and flows up-ward through disk-and-doughnut baffling in thelower portion of the shell, and exits through 28'in. -diam concentric piping at the top.

The Hastelloy-N tubes are L-shaped and arewelded into a horizontal tube sheet at the bottomand into a vertical tube sheet at the top. The

toroidal- shaped header is stronger than a f lat

TABLE V

Primary Heat Exchanger Design Data

a Tubes are not on triangular pitch to facilitate assembly of

bent tubes into hrbe sheets.

Each of4 Units

I

Thermal duty, Bhr/h I

Tube-side conditionst IFluid I

Entrance temPerature, oF I

Exit temPerahtre, oF I

Mass flow rate,lb/h I

Volume salt in tubes, ft3 |

Number tubes I

Nominal hrbe o.d., in. I

Tube thiclmess, in. I

Tube spacing on pitch circle, in'a - |

Radial distance between pitch circles, in'" I

Length, ft I

Effective surfa ce, ftj I

Pressure droP, PSi I

Velocity in tubes, average ft/ sec I

I

Shell-side conditions, I

Fluid.,o.nlEntrance temPerahrre, oF

I

Exit temPerahtre, oF I

Mass flow rate, \b/h I

brside diameter of shelI, ftI Strell thicloress, in.

I ntme spacing, ftI eressure droP, PsiI Vetocity, typical , ftf secI

I On""tll heat transfer coefficient, based on

I -i;;;;;1r1,

hrbe o.d., Btu/(h rt2 "F1

L g22 x 106

Fuel salt I1300 |1o5o I

23.7 x 106 |64 1seoo IIo.oss I

ls*. I

I llrooo I

l:",.*"', I

laso I11150 |

| 1t-8 x 106 I

I s.+

| 2.5

I l-'I z.s

I

I e50

NUCLEAR APPLICATIONS & TECHNOLOGY VOL' 8 FEBRUARY 1970 r99

Bettis and Robertson MSBR DESIGN AND pERFORMANCE

Prf mortrr

Sclf '

header and also simplifies the arrangement forthe coolant-salt flow. The design also permits theseal weld to be located outside the heat exchanger.About 6 ft of the upper portion of the tubing is bentinto a sine-wave configuration to absorb differen-tial expansion. Maximum stresses in the tubingwere estimated to be < 8000 psi.

, Secondory-' !l ss lr

SecondcrySeli

Fig. 7. Sectional elevation of primary heat exchanger for MSBR station.

The tubes have a helical indentation knurledinto the surface to enhance the film heat transfercoefficients. In the cross-flow regions of theexchanger, a heat transfer enhancement factor of2 was used on the inside of the tubes, and a factorof 1.3 was applied to the outside. No enhancementwas assumed in the bent-tube region. The shells

.f'fi

.

:j

200 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O

of the exchangers are also fabricated of Hastel-loy-N. Disk and doughnut baffles are used in the

shetl to a height of '-" 11 ft. Coolant-salt velocityin the unit varies but has an average value of 7 to8 ft/sec.

The heat exchanger design was computer-optim tzed, and the calculations take into accountthe temperature dependence of the salt propertiesas a function of position in the exchanger. Theaverage overall heat transfer coefficient, based on

simple overall log mean temperature diff erenceand total area. is - 950 gtu/(tr tt2 "F), as given inTable V. Limited heat transfer tests are now inprogress, but a more comprehensive program isrequired to establish a final MSBR primary heatexchanger design.

The exchangers are mounted at a point neartheir center of gravity by a gimbal-type joint thatpermits rotation to accommodate the unequalthermal expansions in the inlet and outlet pipes.The entire heat exchanger assembly is suspendedfrom an overhead support structure.

Through close material control and inspection,the heat exchangers are expected to have a highdegree of reliability and to last the 30-year life of

the plant. If maintenance is required, however,provisions have been inctuded in the design forreplacement of tube bundles. No specific arrange-ments are made f or replacement of the shell,although this could be accomplished but with moredif f iculty.

STEAII GENERATORS AND REHEATERS

Each of the four primary heat exchangers has

coolant salt circulated through it by a coolant- saltpump operating in a closed loop independently of

the other three pumps. Each of the coolant- saltIoops contains f our horizontal U- shell, U-tubesteam-ger€rator Superheaters and two horizontalstraight-tube reheaters , eonnected in parallel tothe Coolant-salt sy s te m. A flow-proportioningvalve balances the salt flow between the steamgenerators and reheaters to provide 10001F outletsteam temperature f rom each. The coolant- saltpump speeds can be varied in proportion to the

btectric load on the plant. Design data for the

steam generators and reheaters are listed inTables VI and VII.

The steam generators are supplied with '-' 10 X

106 Lb/h of demine ta1tzed f eedwater at '-" 7 00"F and

3800 psia. A supercritical pressure steam cycle

was selected because it affords a high thermal ef-f iciency and also permits steam to be directlymixed with the high-pressure feedwater to raiseits temperature to ,-,?00'F to guard against fteez-ing of the coolant salt in the steam generator. Forthe same reason, prime steam is also used to


TABLE VI

steam Generator-superheater Design Data

temper the '(cold" reheat steam to 650'F before itenters the reheaters.

After encountering piping stress and otherIayout problems with vertical units, it was found

that by making both the steam generators and re-heaters horizontal, the piping was simplif ied,

stresses were lowered to within acceptable limits,and ceII heights could be decreased'

SALT-CIRCULATING PUMPS

Both the primary and secondary salt- circu-Iating pumps are motor-driven, single-stage,sump-type, centrifugal pumps with overhung im-pellers as illustrated in Fig. B. A helium covergas is used in the salt circulating systems, a por-tion of this gas being fed continuously through the

Iabyrinth seals on the pumps to prevent leakage of

fission products from the systems'A design criterion adopted for the MSBR con-

ceptual stuOy was that pressure drops in the saltcircuits be ii*it"a to a total of '-" 150 ft of head, a

pressure that can be developed by a single- stage

bn*p. Single-stage pumps, with overhung impel-lers and the upper bearing lubricated by a circu-lating oil system, have successfully operated at

Each of16 Units

I

Thermal duty, Btu/h I

I

Tube-side conditions: IF1uid I

Entrance temPerature,oF I

Exit temPerature, oF I

Mass flow rate, Ib/h ]

Number fubesNominal fube o.d., irr.Tube thickness, in.Tube pitch, in.Tube length, ft.Effective surfa ce, ftzPressure droP, PsiEntrance pressure, Psia

I

I SfreU-side conditions:I FluidI n"trance temperature, oF

I B*it temperature, oF

I vtass flow rate, Lb/h

I eressure drop, psiI naffle spacing, ftI mside diameter of shelI, ftI

I Overall heat transfer coefficient, based

| -";l;t*""r,

at, tube o.d. , Btu./l-ff "F)

4L5 x 106

Steam70010006300 x 103

384I20.0770.87573362815037 50

Coolant salt11 508503.85 x 106

6041.5

764

NUCLEAR APPLICATIONS & TECHNOLOGY VOL' 8 FEBRUARY 1970 201

Each of8 Units

Thermal duty, Btu/h

Tube-side conditions:FluidEntrance temperature, oF

Exit temperature, oF

Mass flow rate, Ib/hNumber hrbesNominal ttrbe o.d., in.Tube thickness, in.Tube pitch, in.Tube length, ftEffective surface, ftzPressure drop, psiEntrance pressure, psia

Shell-side conditions :

FluidEntrance temperahrre, oF

Exit temperafure, oF

Mass flow rate, Lb/hPressure drop, psiBaffle spacing, in.Inside diameter of shelI, in.

Overall heat transfer coefficient, basedon log mean At, hrbe o.d. , Btu/ (h ft' "F)

L.25 x 106

Steam65010006420 x 103397340.0 351

27.32L2727570

Coolant salt11 508501.16 x 106588.422

340


TABLE VIISteam Reheater Design Data

ORNL over many thousands of hours in test loopsand in the MSRE. Altlough development will berequired to produce the high- capacity, single-stage pumps needed for the MSBR, Do major tech-nological problems are foreseen.

Multistage pumps having a lower bearing wouldbe required to produce a higher head. Salt-lubricated bearings have been tested with somesuccess at ORNL, but it was judged that thesingle- stage design is a more conservative ap-proach and requires less development effort. Afurther advantage of not using the salt-Iubricatedbearing is that tests of single- stage units havedemonstrated that, when drained of salt, significantamounts of nitrogen gas can be moved by thepumps when operating at ,\', 1200 rpm. Advantageis taken of this in planning for removal of after-heat in the reactor core in event of an unscheduledsalt drain.

DRAIN TANKS

Both the primary and secondary salt- circu-Iating systems are provided with tanks for storageof the salt during start-up and for emergency ormaintenance operations.

BELLOWS

ASSEMBLY

CONCRETESHIELDING

I BEARING ANDSEAL ASSEMBLY

-REACTOR CELL? CONTAINMENT

\--_sHtELD PLUG

./-PUMP TANK(6ft Oin.O0)

-- NORMALSALT LEVEL

IMPELLER

o 5 to 15 2025l,,Irl I I | |

INCHES

Fig. B. Sectional elevation of primary-salt pump forMSBR station.

202 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O

Design obiectives for the fuel-salt drain tanks

are that the highly radioactive contents be con-

tained with a high degree of integrity and relia-bility and that the af terheat- removal system be

Iargely independent cjt the need f or mechanicalequipment, emergency power supply, of initiatingaction by the operating personnel.

The f uel- salt inv.ntbty of - !7 20 f t3 is storedin a single tank , - 12 ft in diameter X 20 ft high'Iocated in the drain-tank cell. Since there is in-sufficient moderation a critical mass could notexist. The cell is heated to maintain the saltabove 930"F, although there would be no permanentdamage in the very unlikely event that the saltwere to freeze in the tank. To limit the tempera-ture rise af ter a drain due to f ission-productdecay heat, the tank is equipped with internal U-tubes through which sodium fluoroborate coolantsalt is circulated by thermal convection to a salt-to-air exchanger located at the base of a naturalconvection stack, as indicated in the flow sheet,Fig. 1. There are two sets of U-tubes and air-cooled exchangers, each of which alone is capable

of maintaining the tank at acceptable tempera-tures.

The fuel- salt storage tank is connected to the

circulating system by a drain Iine having a

freeze-plug-type "valve." At the discretion of the

ptant operator, the plugs could be thawed in a fewminutes for gravity drain of the salt into the tank.

The plugs would also thaw in event of a maior lossof power or failure of the valve coolant system.The salt is returned to the circulating system by

pressurization of the drain tank.A coolant-salt storage tank, - 10 ft in diameter

X 20 ft high, is provided for each of the coolant-salt- circulating loops and for the two heat re-moval systems on the f uel- salt drain tank. The

six tanks have a total storage volume of - 8400 ft3

of coolant salt. The tanks are located in a heated

cell which maintains the temperature above 800T'.

OFF.GAS SYSTEM

The volatile f ission products and colloidalnoble metals formed in the fuel salt are largelyremoved by a helium-gas purge system. The gas-

handting and -disposal system is shown sche-

maticallY in Fig. 1.

About l}vo of the fuel salt discharged f rom each

of the four circulating pumps is bypassed througha gas separator. Swirl vanes in the separatorcaose the gas bubbles to collect in a central vor-tex from which they are withdrawn with nearlyI00vo stripping efficiency. Do1nstream vanes

then kill the swirl. It is desirable that the radio-active gas flow from the separator ..carry some

saltwithit(perhapsaSmuchaSs}vdtoactasasink for the decay heat. The separated gas (- 9


scfm) then flows through an entrainment separatorto remove residual salt before it enters a particletrap and 1000-fts decay tank. The salt removed inthe entrainment separator returns to the pump

bowl.The main bypass flow of salt leaving the gas

separator described above then passes through abubble generator where the salt is recharged withthe helium gas used to strip xenon and other fis-sion product gases from the reactor core. The

bubble generator is a venturi-Iike section in thepipe capable of producing bubble diameters in the

range of 15 to 20 mils. , The fraction of xenon

remaining in the reactor core depends upon theratio of the surface area of the bubbles to the sur-face area of the graphite-and on the permeabilityof the graphite to xenoll. Based on the belief thatthe graphite coating reduces the permeability toxenon as effectively as the tests now indicate, thevolume f raction of bubbles circulating with the

fuel salt probably need be no more than '- 0.\Vo.

The particle trap for the gas leaving the gas

separator also receives about a 2- scf m flow of

helium gas from the pump bowl. The helium isused to purge the reactor of volatile fission pro-ducts and "smoke" of extremely small-sizednoble metal fission-product particles which tend

to accumulate above the liquid interface in the

bowl. The particle trap tank is designed to re-move a fission-product heat load of '-'18 MW(th).The entering gas impinges on the surface of aflowing lead-bismuth alloy where heavy particlesare trapped while the gas passes over the surfaceand is cooled. The lead-bismuth is circulated by

two 900- gaL/min motor-driven pumps to the shellside of heat exchangers cooled bya closed-circuit,boiling-water system coupled to an air- cooledcondenser. The liquid metal is circulated to achemical processing ceII f or removal of the

captured fission-product particles.The highly radioactive gas decays for '-' t h in

the tank and then passes to one of two water-cooled charcoal hotdup traps. These water-cooledtraps absorb and hold up the gases' with the xenon

being retained for - 47 h. On leaving the charcoalbeds, - 9 scfm of the gas passes through a waterdetector and trap to a gas compressor which re-circulates it to the gas purge system bubblegenerator. The remaining 2 scfm is held up forfurther decay and stripping of tritium and othergases before it is returned to the off -gas systemfor purging of the PumP seals, etc.

With the exception of the lead-bismuth particletrap, the components proposed f or the off -gassystem have been tested in principle in the MSRE

or in experimental loops at ORNL. The off-gassystem is described in more detail by Scott andEatherly.'



FUEI-SALT PROCESSING

The nuclear perf ormance of the MSBR isenhanced by continuous removal of the f issionproducts and nuclear poisons. There are alsoadvantages in plant safety to continuous reductionand disposal of the inventory of fission products inthe system. Methods of treating the fuel salt aredescribed in an article by whailey et al.t verybriefly, tJre process consists of diverting a 3 -gat /min sidestream of circulating f uel salt to anadjacent cell for reduction extraction using bis-muth. The Pa is removed on about a 3-day cycleand the rare earth fission products on about a b0-day cycle. It may be noted that the reactor cancontinue to operate f or relatively long periods evenif the fuel processing facility is out of service.

STEAiI-POITER SYSTEIII

The MSBR can produce high-temperature, high-pressure steam to make use of advanced-designsteam turbine- generator equipment.

The steam system flow sheet, Fig. g, considersa supercritical pressure, 1000T', with reheat to1000iF, steam turbine very similar to that used inthe TVA's BulI Run plant. This is a cross-compounded unit of 103 b nrrw(e) gross capacitywith the load about equally divided between thesingle-flow, 3600- rph, high-pressure and two-

flow intermediate pressure turbines on one shaftand the 1800-{pm, four-flow arrangement of low-pressure turbines on the other shaft. It seemscertain, however, that tandem- compounded unitsof this capacity will be available to a future MSBRpower industry. Credit for this has been takenin the cost estimates.

Eight stages of regenerative feedwater heatinghave been shown, with extraction steam takenfrom the high- and low-pressure turbines andfrom the turbines used to drive the two boiler feedpumps. The system is conventional except for theprecautions taken to preheat the entering steam orwater above the liquidus temperature of the cool-ant salt. The 550'F steam from the high-pressureturbine exhaust is preheated to ^,6b0T before itenters the reheaters. This is accomplished in theshell side of an exchanger heated by high-pressuresteam at throttle conditions in the tubes. Thehigh-pressure heating steam leaving the tube sideis then mixed directly with the high-pressure550oF feedwater leaving the top extraction heaterto raise its temperature to - ?00oF. Motor-drivenpumps then boost the feedwater pressure from'\' 3 500 psia at the mixing , rtee, to the steamgenerator pressure of ,^-' 3800 psia. These pumpsare similar to canned- motor supercritical pres -sure boiler recirculation pumps in current use.

The allowable temperature margins to guardagainst significant freezing of the coolant salt in

r l2J:f

950"F

Booste r

Pum ps4.6 MW(e)

(eoc h)

695'FMixing "Tee',

54 psio 1000'F 5.1 x 106 lb/hi-- ^--- ---r

I

r- - - f- Iq- Ps'L lgoo:F l o ' l!'--1b4 |

Mixing "Tee" I FW Heoting

',zt:'1irlil",fiffn'l ^ i i

| < l_ ..... ..Jrit p::h"or",. . I

l.,l i f .ilror l,twtrr-') r=l- -=-L---1 E-r-^^_^"_ | :; I i liL _ _ 4e p,ro s52"F _850"F1>l ! -----+ i i r

steo,i --# .so'r l i --tl9p"l-----j I t

Generotor L-=::J i l-----------J _

j 172 psio 706"Ft932Mw(th), I i i rltll

3500 psio _ago'?Jr.lgLlrun i ffi508Mw(e)Hg obs

rJlQ_sjgsg: i

.L _-J rrI

I

llJ--J

I

I

tn.

I

rJFW Heoting

Reoctor heot =Gross output =Net output =Net eff, =

22s0 Mw(rh)1035 MW(e)1000 MW(e)44.4 o/o

FW Heot ing FW Heoting

Fig. 9. Simplified steam-system flow sheet for 1000 MW(e) MSBR station.

204 NUCLEAR APPLICATIONS & TECHNOIOGY VOL. 8 FEBRUARY 19?O

the heat exchangers are yet to be determined ex-perimentally. No optimization studies of the feed-water and reheat steam systems have been made

and there are obviously many possible variations.For example, if 580"F feedwater and 550"F "cold"reheat steam could be used, thus eliminating theheating and mixing processes suggested above, theflow through the steam generator could be re-duced by one-third and the overall plant thermalefficiency increased from - 44.4 to 44.9V0.

Four control rods of medium reactivity have

been provided. The effective portions of the rodsare graphite, which on removal from the core de-crease the neutron moderation and the reactivity.

The basic principle of MSBR control takes intoaccount its load-following characteristics. An in-crease in turbine load would increase the steamflow rate and tend to lower the steam outlet tem-peratures f rom the steam generators 'and re-heaters. The coolant-salt pump control wouldincrease the pump speed to maintain constant1000'F steam outlet temperatures, thereby tendingto lower the salt temperatures in both the primaryand secondary salt loops. The combination of the

negative temperature coefficient of reactivity andpossible control rod movement would then in-crease the reactor Power.

The MSBR control system has not been studiedin detail, but no major problems are foreseen.Preliminary analog computer studies indicate thatthe primary, secondary, and steam-system loopshave greatly diff erent time constants so thatpower oscillating and control rod position huntingproblems can be avoided.

AFTERHEAT REIi/IOVAL

The normal shutdown procedure after takingthe reactor subcritical would be to continue tocirculate both the salts so as to transfer afterheatto the steam system for rejection to the condens-ing water. The afterheat load would decrease by afactor of 10 in two to four days. After severaldays of circulation, the fuel salt could be drainedto the storage tank without concern f or over-heating of the primary loop due to the depositedfission products remaining in the system.

In the event of excessive salt leakage or othermalfunction which would require quick drainage ofthe fuel salt, nitrogen could be introduced and cir-culated at - 50 psig in the f uel- salt loop. Pre-liminary estimates indicate that three of the fourf uel- salt- circulating pumps (operating at 1200

rpm) can circulate enough gas to keep the temper-atures within tolerable levels. This assumes thatat least one of the four secondary-salt pumps isoperative.


MAINTENANCE

Any portions of the system containing fuel saltwill become suff iciently radioactive to requirethat all maintenance operations be accomplishedby other than direct methods. The coolant saltwiII e>iperience induced activity, primarily due toactivation of the sodium atoms. The coolant- saltpumps, steam generators and reheaters, however'witl not become radioactive, and after the coolantsalt is drained and flushed, could probably be ap-proached for direct maintenance.

CeII roof access openings will be provided overaII equipment that might need servicing. Theplugs are replaceable with a work shield throughwhich long-handled tools can be manipulated. Abridge-type manipulator will be mounted beneaththe work shield to perform more sophisticatedoperation. Viewing can be accomplished throughcell windows, periscopes, and closed-circuit tele-vision.

Replacement of the reactor core graphite willbe a maintenance operation that can be planned inconsiderable detail. The assembly to be removedis - 16 ft in diameter, 3 0 ft high (including thecylinder extension), ild weighs - 240 tons. After10 days of decay time the activity level at theouter circumference would be '^'' 5 x 105 R/h oncontact. The f irst step in replacing the coregraphite would be to disconnect the attachmentsand holddown devices. A transition piece and

track-mounted transport cask would then be posi-tioned on the operating floor directly above thereactor, and a lifting mechanism inside the caskused to draw the core assembly up into it. Thecask would then be closed by means of slidingshields, moved into position over the reactorstorage cell, and the spent core assembly loweredinto it.

A replacement core assembly would have pre-viously been made ready for lifting into the trans-port cask and subsequent lowering into the reactorveSSeI. Since this arrangement requires two topheads and control rod drive assemblies for thereactor vessel to be on hand, these have been in-cluded in the direct construction cost of the plant.Very preliminary estimates indicate that the capi-tal cost of special maintenance equipment wouldbe in the range of $6 million. These studies givepromise that the core graphite can be replacedwithin the usual downtime allowance f or otherplant maintenance.

Although there is some background of exper-ience in maintenance of the MSRE and other re-actor types, obviously the above commentsregarding MSBR maintenance procedures andcosts are speculative and wiII require f urtherevaluation in continued study and testing.

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O 205


COST ESTIIIIATES

Methods used in making the cost estimates foran MSBR power station conform largely to thoseused by the USAEC in reactor cost evaluationstudies. It is important to note that the MSBRestimated construction cost is not that of a firstplant but that of a new plant in an establishedmolten-salt reactor industry. It is further as-sumed that there is a proven design essentiallythe same as the one described here, that develop-ment costs have been largely absorbed, and thatthe manufacture of materials, plant constructioo,and licensing are routine. on this basis the con-tingency and other uncertainty factors assumedf or MSBR costs are little different from thoseapplied to other reactor types.

costs are based on the early 1969 value of thedollar. It is assumed that it would require fiveyears to construct a plant. Escalation of pricesduring the construction period was not included inthese estimates but could amount to - !2Vo addi-tional cost. Interest (at 6/) on capital during the5-year construction period was included, however.Expenses of an extended (b4-h) work week wereapplied, as was a 3Vo Local sales tax.

TABLE VIIIEstimated Construction Cost of 1000-MW(e)

MSBR Power Station

tBased on early 1969 prices.

26

TABLE D(

Explanation of lrdirect costs used in Table vIII(For direct cost of 91) Percent Total Cost

General and administrativeMisc ellaneous constructionArchitec t-engineer feesNuclear engineering feesStart-up costsContingencyhterest during constnrctionLand ($300 000)

4.71.05.12.00.72.7

tu_.u

L.047L.0571.1111.134L.L42L.L7 21.3311.335

substantive cost data are available for makingMSBR estimates on about half the items in theMSBR plant. The remaining costs are less certainbecause of the preliminary nature of the reactorsystem designs, and because the special graphiteand Hastelloy-N materials have aceumulated tittlelarge-scale production experience. The fuel cyclecosts are discussed in a companion paper of thisseries. The estimated construction costs areshown in Tab1e VIII and the breakdown of indirectcosts in Table D(. The power production costs aregiven in Table X.

The reactor vessel, heat exchangers, draintanks, etc., were estimated on the basis of an in-stalled cost of Hastelloy-N parts of $g/tn forsimple cylindrical shapes, $r2/rb for vesselheads, etc.,, and $20/Lb f or tubing and other specialshapes. The assumed instalted graphite cost wasestimated at $8/tu for simple slabs and g 10/lb forthe extruded core elements. The cost of an alter-nate reactor head assembly is included.

Fixed charges f or the capital, which includeinterest depreciation, routine replacements (otherthan reactor core), taxes, and insurance, total13.7Vo for a plant owned by a private utility.

The estimated cost of replacing the coregraphite every 4 years is given in Table xI. It isassumed that the replacement can be accomplished

TABLE XEstimated Power Production Cost in 1000-MW(e)

MSBR Station

Depreciating capitala I S.1 mills/kwhFuel cycle costb | 0.?Operating cost | 0.8

Total | +t mils/kwh

"Based on fixed charges of 1gTo/year and 80Vo plant fac-tor.

blncludes graphite replacement cost of 0 .L miL/kwh(e),based on A-year life of graphite and capitalizationcosts totaling 8Vo for replacement expenses.

$1ouu

Stmchrres and improvements

ReactorVesselGraphiteShielding and containmentHeating and eooling systemsControl rodsCranes

Heat transfer systemsDrain tanksWaste treatment and disposalInstnrmentation and controlsFeedwater supply and treatmentSteam pipingRemote maintenance and equipmentTurbine- generator plant equipmentAccessory electrical equipmentMiscellaneous

Tota1 + 6Vo allowance for 54-h work weekTotal + 3Vo loca1 sales ta>rTotal + 33.570 indirect costs (see Table

D()

11

5.96.03.11.31.00.2

22.24.30.54.L5.25.66.6

25.05.2L.4

109J

$115. 6

$r.1e.1

$15e.0

5

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O

Bettis

TABLE XI

Explanation of Graphite Replacement CostUsed in Table X

$106

Cost of graphite per rePlacement

Cost of Hastelloy-N per replacement

Special labor costs per replacement

Replacement cost factor for 4-year life toconvert to present worth (at 6Vo\, based on30-year plant life

Total 30-year replacement cost

Production cost of graphite replacement,based on fixed charges of 8c/o, 80Vo plantfactor, mills/kWh

$ 3.3

0.5

0.3

$4J

(3.04)

$L2A

0.1

within the usual plant downtime for other mainte-nance that is accommodated within the 80Vo plantfactor.

The general conclusion has been reached thatthe direct construction cost of an MSBR is similarto that of a PWR. This cost position of the MSBRis largely due to the high thermal efficiency of theplant and the use of modern high-speed turbine-generators having a relatively low first cost perkilowatt. The above featur€s, combined with a lowfuel-cycle cost, permit MSBR stations to operatewith low power production costs and with high loadfactors throughout their life.

In summ&ry, this study of a conceptual design

and Robertson MSBR DESIGN AND PERFORMANCE

for a 1000-NIW(e) MSBR po\Mer station has indi-cated that it is feasible to design practical ther-mal breeders that would contribute significantly toconservation of f ertile f uel resources and be

capable of generating low-cost electric power bothin the near and distant future.

ACKUOTLEDGTETTTS

This research was sponsoredEnergy Commission under contractbide Corporation.

by the (J.S. Atomicwith the Union Car-

REFERENCES

1. A. M. PERRY and H. F. BAUMAN, "Reactor Physicsand Fuel Cycte Analysis," Nucl. Appl. Tech., 8, 208(1970).

2. DUNLAP SCOTT and W. P. EATHERLY, "Graphiteand Xenon Behavior and Their Influence on Molten-SaltReactor Design,t' Nucl. Appl. Tech., 8rL79 (1970)'

3, M. E. WHATLEY, L. E. McNEESE, 'W'. L. CARTER'L. M. FERRIS, and E. L. NICHOLSON, "EngineeringDevelopment of the MSBR FueI Recycla," Nucl. Appl.Tech., ,8, L70 (1970).

4. H. E. McCOY, W. H. COOK, R. E. GEHLBACH, J. R.WEIR, C. R. KENNEDY, C. E. SESSIONS,R.L. BEATTY,A. P. LITMAN, and J. W. KOGER, "Materials forMolten-Salt Reactors," Nucl. Appl. Tech., 8, 156 (1970).

5. 'W. R. GRIMES, "Molten Salt Reactor Chemistry,t'Nncl. Appl. Tech., 8, 137 (1970).

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. B FEBRUARY 1970 207

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