A_"_LORNL-3996 _';-IUC-SO Reactor Technology " ‘BREEDER REACTORS --‘-.:Pa'ul R -Kcst'en E. S. Beths Royic Roberrson ‘ '._?’;‘.‘},;.DESIGN ;STUDIE_S OF lOOO-Mw(e) MOLTEN sALT._’I-_ff_ff T _;.‘.\‘\rv)|< AR " Wil i loe ) s - - Pnnfed in USA. Pnce $5 00 Y Avu[lcbie from the Cleormghouse for Federcl '5'7""7 s Sc:enhhc and Techpical informchon, Natienal Bureuu of Standards,® ' < VU.S. Depcrtrne,nf of Commer_ce__,_,_Spnngfn-cld,, V|rgmu_:r_ 22151 - ;nor the Commlsslcn, nor uny person nchng on behalf of fhe Comm:ssion. or hus empleymeni vmh such confrector. SR .Thls reporf was prepm‘ed as on account of Govefnment sponsored work Nelther fhe Umfed Sfctes, ' A“ Mokes any warrcnfy or represenioflon, sxpressed of |mphed “with reepecf to the cccuracy, . completeness, ‘or - usefulness -of the informuhon ct:mh:unet:l m this report, or Ihat the vse of ~ any mformchon, cppcratus, rmefhod or process dnsclosed in Ihls report moy nof lnfrmge privately owned nghfs, of oo e e : T i "B, _':Assumes any Imb:lmes w:th respect to the ise of or for dqmages :esultmg from fhe use of ~ony mforrnchon, cppcrcfus, methed, ‘or procéss dlsclosed in this report. © T 0 - i As used in the ‘abave, 'parson acting ‘on behalf of the Commlsslon mcludes any emp!oyee or 7 : '_;‘Vcontracfor of the Commlssmn, or employee of such confrucfor, to fl-ne extent thet - such employee :ior confroctor of 'he Commlssion, or emplcyee of " such confracfor prepcres, dwsemmotes, or ' _:;provrdes access to, any Infermahon pursucnt to. I-ns employmenf or conircct w[!h tho Commasslon, S Ay G v - ( CFSTI PRICES /) o0 ne s 35.90; mugZ ORNL-3996 Contract No. W-7405-eng-26 DESIGN STUDIES OF lOOO-Mw(e) MOLTEN-SALT BREEDER REACTORS Paul R. Kasten E. S. Bettis Roy C. Robertson Molten-Salt Reactor Program R. B. Briggs, Director W RELEASED FOR ANNOUNCEMENT | IN NUCLEAR SCIENCE ABSTRACTS e b gt A AUGUST 1966 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION «t - iii SUMMARY Design and evaluation studies have been made of thermal-energy molten-salt breeder reactors (MSBR) in order to assess their economic and nuclear performancehand to identify important design and development prob- lems. The reference reactor design presented here is related to molten- salt reactors in general. The reference design is a two-region two-fluid system, with fuel salt separated erm the_blanket salt by graphite tubes. The fuel salt consists of uranium fluoride dissolved in a carrier salt of 1lithium and beryllium fluorides, and the blanket salt contains thorium fluoride dissolved in a similar carrier salt. The energy generated in the reactor fluid is trans- ferred to a secondary coolant-salt cifcuit, which couples the reactor to a supercritical steam cycle. On-site fuel-recycle processing is employed, with fluoride-volatility and vacuum-distillation operations used for the fuel fluid, and direct-protactinium-removal processing applied to the blanket stream. The resulting power cost for the reference plant, termed MSBR(Pa), is less than 2.7 mills/kwhr(e); the specific fissile-material inventory is only 0.7 kg/Mw(e), the fuel doubling time is about 13 years, and the fuel-cycle cost is 0.35 mill/kwhr(e). The associated power dou- bling time based on continuous investment of bred fuel is less than 9 years. Reference MSBR Plant Design Flowsheet Figure 1 gives the flowsheet of the lOOO-Mw(e) MSBR power plant. Fuel flows through the reactor at a rate of about 44 OOO gpm (velocity of about 15 fps); it enters the core at 1000°F and leaves at 1300°F. The prlmary fuel C1rcuit hag four loops, and each loop has a pump and a pri- mary heat exchanger. Each of these pumps has a capaC1ty of 'about 11,000 - gpm. The four blanket-ealt pumps and heat exchangers, although smaller, are similar to corresponding components in thegfuei system. The blanket salt enters the reactor vessel at 1150°F and leaves-at 1250°F. The blanket-salt pumps have a capacity of about 2000 gpm. REACTOR VESSEL 5.134 # ORNL-DWG 66-7022 T 7 - 10.067 # 7.152¢ | 1518.5h-540p- 1000° 1550~ 1000*F - - | 1518.5h ! | i | 1424 h-3515p-1000° i [ € l;— 1 ! i r | 300p-1000°F | i GEN. o ! 150 ft%sec L |20t Vs ! | : 52'(!;.2 Mwe 73ty 9.7 ¥ | BOILER | : | o o = i2soer 1300°F REHEATERS< | | SUPERHEATERS 2s6In| 4 4 1150°F B50°F A J8s0°F Y 4 [ r 551.7° ' | BLANKET SALT HEAT COOLANT SALT COOLANT SALT! ! GEN. EXCH. AND PUMPS ' ; ;. PUMPS PUMPS ' TURBINE [ TuriNg [] 3077 Mwe FUEL SALT HEAT i ., | _ Gross [uzsee EXGH. AND PUMPS g‘ @_“.f.'_”.a“ i ! r . 5. 5- | | I T b T REHEAT STEAM oo IIPE — i LL'_[‘ T ! I | | PREHEATERS CONDENSER B FEEDWATER | | ' = |' 3800p-7b0'F E 1307.8h SYSTEMS | S | [te%2n 3500p-866°F < o S| L __J i 3500p - 550.9° J 1125°F hd v seash_ A Ao 234 | a0 f 0 570 p-GSO‘fi 3475p- 695°F ! eg—————— " j— P — 766.4h BOOSTER MIXING TEE PUMPS BLANKET SALT FUEL SALT COOLANT SALT PERFORMANCE DRAIN TANKS DRAIN TANKS DRAIN TANKS NET OUTPUT 1000 Mwe ‘ LEGEND - GROSS GENERATION 1,034.9 Mwe BF BOOSTER PUMPS 9.2 Mwe FUEL == STATION AUXILIARIES 257 Mwe BLANKET === wwe REACTOR HEAT INPUT 2225 Mwi COOLANT —--— NET HEAT RATE 7,601 Blu/kwh STEAM ———=—— NET "EFFICIENCY 449 % o 10% 1b /hr Pmmmmmn P8I0 Mo Bt /1b. e .. Fraeze Valve Fig. 1. Reference MSBR Flow Diagram v &4 o oy & - s - Four 14,000-gpm pumps circulate the coolant, which consists of a mix- ture of sodium fluoride and sodium fluoroborate. The coolant enters the shell side of the primary heat exchanger at 850°F and leaves at 1112°F. After leaving the primary heat exchanger, the coolant salt is further heated to 1125°F on the shell side of the blanket heat exchangers. The coolant then circulates through the shell side of 16 once-through super- heaters (four superheaters per pump). In addition, four 2000-gpm pumps circulate a portion of the coolant through eight reheaters. The steam system flowsheet is essentially that of the new Bull Run plant of the Tennessee Valley Authority system, with modifications to in- crease the rating to 1000 Mw(e) and to preheat the working fluid to 700°F prior to entering the heat exchanger-superheater unit. A supercritical power-conversion system is used that is appropriate for molten-salt appli- cation and takes advantage of the high-strength structural alloy employed. Use of a supercritical fluid system results in an overall plant thermal efficiency of about 45%. Reactor Design Figure 2 shows the plan and elevation views of the MSBR cell arrange- ment. The reactor cell is surrounded by four shielded cells containing the superheater and reheater units; these cells can be individually iso- lated for maintenance. The fuel processing plant, located adjacent to the reactor, is divided into high-level and low-level activity areas. The elevation view in Fig. 2 indicates the position of equipment in the various cells. Figure 3 gives an elevation view of the reactor cell and shows the - location of the reactor, pumps, and fuel and blanket heat exchangers. The Hastelloy N reactor vessel has a side-wall thickness of about 1.25 in. and a head thickness of about 2.25 in.; it is designed to operate at 1200°F and up to 150 psi. The plenum chambers at the bottom of the ves- sel commmnicate with the external heat exchangers by concentric inlet- outlet piping. The inner pipe has slip joints to accommodate thermal expansion. Bypass flow through these slip joints is about 1% of the total flow. As indicated in Fig. 3, the heat exchangers are suspended vi from the top of the cell and are located below the reactor. Each fuel pump has a free fluid surface and a storage volume that permit rapid drainage of fuel fluld from the core upon loss of flow. In addition, the fuel salt can be drained to the dump tanks when the reactor is shut down for an.extended time. The entire reactor cell is kept at high tem- perature, while cold "fingers" and thermal insulation surround structural support members and all special equipment that must be kept at relatively ' REHEAT STEAM ORNL-DWG 66=7111 1. e “ " '. o | i ’ H.P. STEAM -I‘l n " WASTE GAS N m n FEEDWATER FEEDWATER 1| CELL o nn HP STEAM , fN0f. FUEL nn LP. STEAM 28 OOLANT SALT nil- ) ¢ PUMPS ,’f-, « .{qn | HEAT EXCH [ rr{' fl : et anl L o Y AN DI et 4 2 = ; ‘ . i B 0 X o 2 e s v N 3 2 4 I = A 1 L2 — = ' 54 12 ) ¢ : & 7 1 N + 8 REHEATERS or . AW L 1 é ‘7 ' ‘.” 2 ? v ! DEC(;I:’{)AQ_:%%‘Z@E ~. {j ) i 16 SUPERHEATERS 4 . ¢ BLANKET , ; g - HEAT EXCH. 30 ; ) B X CONTROL AREA—* py S g e Tt s | T ¥ - 144 - ORNL-DWG 66-7110 / CONTROL ROD DRIVE COOLANT SALT— FUEL CIRCULATING—, / BLANKET CIRCULATING PUMPS —\ PUMP// ) ; / / PUMP SUPERHEATERS—. _ ’ N - FUEL HEAT EXCH. Z REHEATERS HEAT EXCH. Fig. 2. Reactor and Coolant-Salt Cells — Plan and Elevation. s ) ORNL~DWG 66-7109 BLANKET PUMP MOTCR FUEL PUMP MOTOR CONTROL ROD DRIVE - CONSTANT — £.07s SUPPORT |~%: HANGERS w) FUEL DUMP [ _ TANK WITH _. HERTTTT > Sl COOLING = COILS FOR AFTER HEAT REMOVAL LANKET et HEAT ' EXCH. < _____ e, - - . 2 AR : 2. w 3 4 . . ' ) ~ L TT—I0 FT. DIA . CORE '+ VESSEL ‘ a 4 . l -‘ e .'.': o “ A "'"' b ‘l' ’ ’ o’ L d. .'” . L .' ':. : e R ’ 'l * L . « " “REACTOR CELL HEATERS |, P I T e T, e ‘A_-. 'n_'.-.!\:l,l 04”\“ * P s * v _,s‘:-"'r 'Fig. 3. Reactor Cell — Elevation. Ty viii low temperatures. The control rod drives are located above the core, and the control rods are inserted into the central region of the core. The reactor veséel, gbout 14 £t in diameter and about 19 ft high, contains a 12.5-ft-high 10-ft-diam core assembly composed of reentry- type graphite fuel elements. The graphite tubes are attached to the two plenum chambers at the bottom of the reactor with graphite-to-metal transition sleeves. Fuel from the entrance plenum flows up fuel passages in the outer region of thé fuel tube and down through a single central passage to the exit plenum. The fuel flows from the exit plenum to the 0 heat exchangers and then to the_pumprand back to the reactor. An 18-in.- thick molten-salt blanket plus a 3~in.-thick graphite reflector surround . the core. The blanket salt also permeates the interstices of the core lattice, and thus fertile material flows through the core without mixing with the fissile fuel salt. | The MSBR requires structural integrity of the graphite fuel element. In order to reduce the effect of radiation damage, the fuel tubes have been made small to reduce the fast flux gradient across the graphite wall. Also, the tubes are anchored only at one end to permit axial movement. The core volume has been made large in order to reduce the flux level in the core. In addition, the reactor is designed to permit replacement of the entire graphite core by remote means if required. _ Figure 4 shows a cross section of a fuel element. Fuel fluid flows upward through the small passages and downward through the large central . passage. The outside diameter of a fuel tube is 3.5 in., and there are 534 of these tubes spaced on & 4.8-in. triangular pitch. The tube as- - semblies are surrounded by hexagonal blocks of moderator graphite with blanket salt filling the interstices. The nominal core composition is 75% graphite, 18% fuel salt, and 7% blanket salt by volume. In determining the design parameters of the MSBR, two different methods were considered for removal of bred fuel from the reactor. The designation MSBR(Pa) represents a plant in which protactinium is removed directly from the blanket stream, whereas the designation MSBR corre- sponds to remcval of uranium per se from the blanket. With the exception of the blanket-processing step, the MSBR(Pa) and the MSBR plants have (;J essentially the same design. Development of an MSBR(Pa) plant is the ' ') ) Y MODERATOR GRAPHITE) FUEL PASSAGE (UP ix £ > ' ORNL~-DWG 66-7139 BLANKET PASSAGE FUEL PASSAGE (DOWN) 35 OD. FUEL TUBE MODERATOR HOLD DOWN NUT GRAPHITE) N Al o § 3 REACTOR | Wi S - —SPACER I 1 METAL TO GRAPHITE - SLIP-JOINT i-.‘___\. —————METAL TO GRAPHITE t - BRAZED JOINT | | ——BRAZED JOINT l —/ = .\ : r. Ny b= = 2 \&E ~\. ——FUEL INLET : i s PLENUM = - 1 3 FUEL OUTLET e re PLENUM Fig. 4. MSBR Graphite Fuel Element. present goal of the molten-salt reactor program. A summary of the parameter values determined for the MSBR(Pa) and MSBR designs is given in Table 1. Fuel Processing The primary objectives of fuel processing are to purify and recycle fissile and carrier components and to minimize fissile inventory while holding 1osses to a low value. The fluoride volatility—vacuum distilla- tion process fulfills these objectives through simple operations. The process for direct protactinium removal from the blanket also appears to be a simple one. o | The core fuel for both the MSBR and the MSBR(Pa) is processed by fluoride volatility and vacuum distillation operations. For the MSER, blanket processing is accomplished by fluoride volatility alone, and the processing cycle time is short enough to maintain a very low concentraé tion of fissile material. The effluent UFg is absorbed by fuel salt and reduced to UF, by treatment with hydrogen to reconstitute a fuel-salt mixture of the desired composition. For the MSBR(Pa), the blanket stream is treated with molten bismuth containing dissolved thorium; the thorium displaces the protactinium from solution (as well as uranium). The metal- lie protactinium and uranium are deposited on a metal filter and hydro- fluorinated or fluorinated for recyéle of bred fuel. Molten-salt reactors are inherently suited to the design of process- ing facilities integral with the reactor plant; these facilities require only a small amount of cell space adjacent to the reactor cell. Because all services and equipment available to the reactor are available to the processing plant and shipping and storage charges are eliminated, inte- gral processing facilities permit significant savings in capital and dperating costs. Also, the processing plant inventory of fissile mate- rial is very low. The principal steps in core and blanket stream processing of the MSBR(Pz.) and the MSBR are shown in Fig. 5. A small side stream of each fluid is continuously withdrawn from the fuel and blanket loops and circu- lated through the processing system. After processing, the decontaminated fluids are returned to the reactor system. Fuel inventories retained in 2 O (,\ w) ) xi Teble 1. Reactor Design Values MSBR(Pa.) MSBR Power, Mw Thermal 2225 Electrical 1000 Thermal efficiency, fraction 0.449 Plant load factor 0.80 Reactor wvessel Outside diameter, ft 14 Overall height, ft ~19 Wall thickness, in. 1.5 Head thickness, in. 2.25 Core Height of active core, ft 12.5 Diameter, ft 10 Number of §raphite fuel passage tubes 534 Volume, ft 282 Volume fractions : Fuel salt 0.169 0.169 Blanket salt 0.073 0.074 Graphite moderator 0.758 0.757 Atom ratios Thorium to uranium 42 40 Carbon to uranium 5800 5440 Neutron flux, core average, neutrons/cmzosec Thermal 7.2 x 1014 6.7 x 1014 Fast 12.1 x 10*#4 12.1 x 104 Fast, over 100 kev 3.1 x 10*4 3,1 x 10t Power den31ty, core average, kw/liter Gross 80 In fuel salt 473 Blanket Radial thickness, ft 1.5 Axial thickness, ft 2.0 Volume, £t3 1120 Volume fraction, blanket salt 1.0 Reflector thickness, in. 3 Fuel salt Inlet temperature, °F 1000 Outlet temperature, °F 1300 Flow rate, ft’/sec (total) 95.7 £pm 42, 950 Nominal volume holdup, £t3 Core 166 Blanket 26 Plena 147 Heat exchangers and plping 345 Processing plant .33 717 - Total xii Table 1 (continued) MSBR(Pa) MSBR Fuel salt (continued) Nominal salt composition, mole % LiF 63.6 BeFa 36.2 UF; (fissile) 0.22 Blanket salt Inlet temperature, °F 1150 Outlet temperature, °F 1250 Flow rate, ft3/sec (total) 17.3 g£pm 7764 Volume holdup, £%3 Core 72 Blanket - 1121 Heat exchanger and piping 100 Processing : 24 Storage for protactinium decay 2066 Total 1317 3383 Salt composition, mole % LiF 71.0 T11F4 27.0 UF; (fissile) 0.0005 System fissile inventory, kg 681 769 System fertile inventory, kg 101,000 260, 000 Processing data Fuel stream Cycle time, days 42 47 Rate, ft3/day 16.3 14.5 Processing cost, $/ft> 190 203 Blanket stream Equivalent cycle time, days Uranium-removal process 55 23 Protactinium-removal process 0.55 Equivalent rate, ft3 per day - Uranium-removal process 23.5 144 Protactinium-removal process 2350 Equivalent processing cost (based on 65 7.3 uranium removal), $/ft> Fuel yield, %/yr 7.95 4.86 Net breeding ratio- 1.071 1.049 Fissile losses in processing, atoms per 0.0051 0.0057 fissile absorption S Specific inventory, kg of fissile material 0.681 0.769 per megawatt of electricity produced Specific power, Mw(th)/kg of fissile material 3.26 2.89 Fraction of fissions in fuel stream 0.99% 0.987 Fraction of fissions in thermal-neutron group 0.815 0.806 Net neutron production per fissile 2.227 2.221 absorption (n¢) » @ ORNL-DWG 66-7668 SORBER Y S/ &’ / /| [ < // y %« _ STORAGE /] NaF /MgF 5/ UFG+ (s VOLATILE FP MAKE UP : LiF/ BeFy/ ThFg, r NN, .' s 772 7 7. 7/////7 40NT|NU0U5/ FERT'L‘V/ 7, FLUORIDE 7/} - MAKE U/ 1/ | voLamiLiTy N N\ ey LiF/BeF, /ThF,/FP Fy— (7777 77 DISCARD FOR [/ FP REMOVAL L L Ll LSS FERTILE STREAM RECYCLE Fig. 5. MSBR(Pa) UFg RECYCLE TO REACTOR =) S S 7z 7 /S S coLp /] TRAP V //N 100 V' /S S 7% 7 o, 77777 D CONVERSION LSS ) ANPO REMOVAL z v LA et e e e e — ERTILE STREAM RECYCLE v SALT DISCARD FOR FP REMOVAL EXCESS /] | PRODUCTION v oo UFg + VOLATILE FP AKE U LiF/BeF, v v 77777 77T — |- - HOLDUP, ,/connuuous; ? 7// A ///// 7///// / FLUORIDE ~, / v 7 ur—Ur, 7|/ 7 [ FP DECAY | » VOLATILITY 7} /| DISTILLATION,, /) repuction | |FiLTRATION ] . e ), N atFy=URg v //// / N LA 7 0 » A - L., REQUCED METALS Cr, Fe; N LiF /BeFy /UF4 RECYCLE Fuel- and Fertile-Stream Processing for the MSBR and MSBR(Pa). TTIX xiv the processing plant are estimated to be about 5% of the reactor system for core processing and less than 1% for blanket.processing. Heat'Exdhangg_and Steam Systems The structural material is Hastelloy N for all components contacted by molten salt in the fuel, blanket, and coolant systems, including the reactor vessel, pumps, heat exchangers, piping, and storage tanks. The- primary heat exchangers are of the tube-and-shell type, with fuel salt on the tube side. Each sheil contains two concentric tube bundles at- tached to fixed tube sheets. Fuel flows through the two bundles in series; it flows downward in the inner section of tubes, enters a plenum at the bottom of the ekchanger, and then flows upward to the pump through the outer section of tubes. The coolant salt enters at the top of the ex- changer and flows on the baffled shell side down the outer annular re- gion; if then flows upward in the inner annular section before exiting through a pipe centrally placed in the exchanger. ' Since a large temperature difference exists in the two tube sections, the design permits differential tube expansion. Changes in tube lengths due to thermal conditions are accommodated by the use of a sine-wave type of construction, which permits each tube to adjust to thermal changes. The blanket heat exchangers increase the temperature of the coolant leaving the fuel heat exchangers. The design of these units is similar to that used in the fuel heat exchangers. The superheater is a U~tube U-shell heat exchanger'that has disk and doughnut baffles with varying spacing; it is a long, slender exchanger. The baffle spacing is established by the shell-sidé pressure drop and by the temperature gradient across the tube wall; it is greatest in the central portion of the exchanger where the temperature difference between the fluids is high. The supercritical fluid enters the tube side of the superheater at 700°F and 3800 psi and leaves at 1000°F and 3600 psi. The reheaters transfer energy from the coolant salt to the working fluid before its use in the intermediate pressure turbine. A shell-and- tube exchanger is used that produces steam at 1000°F and 540 psi. Since the freezing temperature of the secondary coolant salt is about 700°F, a high working fluid inlet temperature is required. Preheaters, ') *® wy - oy XV along with prime fluid, are used in raising the temperature of the work- ing fluid entering the superheaters. Prime fluid goes through a pre- heater exchanger and leaves at a pressure of 3550 psi and about 870°F. It is then injected into the feedwater in a mixing tee to produce fluid at 700°F and 3500 psi. The pressure is increased to about 3800 psi by a pressurizer (feedwater pump) before the fluid enters the superheater. Capital Cost Estimates Reactor Power Plant Preliminary estimates of the capital cost of a 1000-Mw(e) molten- salt breeder reactor power station indicate a direct construction cost of about $80.7 million. After applying the indirect cost factors asso- ciated with reactor construction, an estimated total plant cost of $114.4 million is obtained for private-financing conditions and $110.7 million for public financing. A summary of plant costs is given in Table 2. The relatively low capital cost estimate obtained is due to the small physical size of the reactors and associated equipment, the high thermal efficiency, and the simple control requirements. The operating and maintenance costs of the reactor power plant were estimated by standard procedures and were modified to reflect present-day salaries. These costs amount to 0.34 mill/kwhr(e). Fuel-Recycle Plant The capital costs associated with fuel-recycle equipment were ob- tained by itemizing and costing the major process equipment required and estimating the costs of site, buildings, instrumentation, waste disposal, and building services associated with fuel recycle. Table 3 summarizes direct construction costs, indirect costs, and total costs associated with an intégrated processing facility having approximately the capacity required for a 1000-Mw(e) MSBR plant. The total construction cost was estimated to be about $5.3 million; in ob- taining this figure, the indirect charges amounted to about 100% of the direct construction cost. The high value used for the indirect charges XVl Table 2. Preliminary Cost-Estimate Summary® for a 1000-Mw(e) Molten-Salt Breeder Reactor Power Station [MSBR(Pa} or MSBR] Federal Power Costs Commission (in thousands of dollars) Account 20 Land and Land Rights ] 360 21 Structures and Improvements ! 211 Ground improvements 866 . 212 Building and structures .1 Reactor building® o 4,181 .2 Turbine building, auxiliary building, and feedwater 2,832 heater space ) .3 Offices, shops, and laboratories 1,160 2 .4 Waste disposal building 150 .5 Btack 76 .6 Warehouse 40 .7 Miscellaneous 30 Subtotal Account 212 8,469 Total Account 21 9,335 22 Reactor Plant Equipment ’ 221 Reactor equipment .1 Reactor vessel and internals 1,610 .2 Control rods 250 «3 Bhielding and containment 2,113 +4 Heating-cooling systems and vepor-suppression system 1,200 .5 Moderator and reflector 1,089 .6 Reactor plant crane 265 Subtotal Account 221 6,527 222 Heat transfer systems .1 Reactor coolant system 6,732 +2 Intermediate cooling system 1,947 .3 Steam generator and reheaters 9,853 +4 Coolant supply and treatment 300 Subtotal Account 222 18,832 223 Nuclear fuel handling and storage (drain tanks) 1,700 = 224 Nuclear fuel processing and fabrication (included in (c) fuel-cycle costs ) 225 Redioactive waste treatment and disposal (off-gas 450 < system) 226 Instrumentation and controls 4,500 227 TFeedwater supply and treatment 4,051 228 Steam, condensate, and feedwater piping 4,069 229 Other reactor plant equipment (remote maintenance) 5,000d Total Account 22 45,129 aEstimates are based on 1966 costs for an established molten-salt nuclear power plant industry. bContainment cost is included in Account 221.3. ®see Table 3 for these costs., dThe allowance for remote maintenance may be too high, and some of the included replacement equipment allowances could be classified as operating expenses rather than first capital costs. o " wh w) Xvii Table 2 (continued) Federal Power Costs Commission (in thousands of dollars) Account 23 Turbine-Generstor Units 231 Turbine-generator units 19,174 232 Circulating-water system 1,243 233 Condensers and auxiliaries 1,690 234 Central lube-oil systen 80 235 Turbine plant instrumentation 25 236 Turbine plant piping 220 237 Axuiliary equipment for generator 66 238 Other turbine plant eguipment 25 Total Account 23 22,523 24 Accessory Electrical 241 Switchgear, main and station service 500 242 Switchboards 128 243 Station service transformers 169 244 Auxiliary generator 50 245 Distributed items 2,000 Total Account 24 2,897 25 Miscellaneous 800 Total Direct Construction Cost® 80,684 Private Financing Total indirect cost 33,728 Total plant cost 114,412 Public Financing Total indirect cost 30,011 Total plant cost 110,695 ®Does not include Account 20, Land Costs. However, land costs were included when computing indirect costs. Land is treated as a nondepreciating capital item. xviii Table 3. Summary of Processing-Plant Capital Costs for a 1000-Mw(e) MSBR Installed process equipment $ 853,760 Structures and improvements 556,770 Waste stofage 387,970 Process piping 155,800 Process instrumentation 272,100 Electrical auxiliaries | 84,300 Sampling connections 20, 000 Service and utility piping | 128,060 Insulation 50,510 Radiation monitoring 100, 000 Total direct cost $2,609, 270 Construction overhead 782,780 (30% of direct costs) — e Subtotal construction cost $3,392,050 Engineering and inspection 848,010 (25% of subtotal construction cost) —_— Subtotal plant cost | $4, 240,060 Contingency (25% of subtotal 1,060,020 plant cost) Total capital cost $5,300, 080 should more than compensate for the higher rates of equipment replacement in the fuel-processing plant as compared with the power plant as a whole. The operating and maintenance costs for the fuel-recycle facility include labor, labor overhead, chemicals, utilities, and maintenance mate- rials. The total annual operating and maintenance costs for a processing facility having a throughput of 15 ft3 of fuel salt per day plus 105 ft> of fertile salt per day is estimated to be about $721,000. A breakdown of these charges is given in Table 4. These capital and operating costs were used as base points for ob- taining the costs for processing plants having different capacities. For each fluid stream the capital and operating costs were estimated separately 3] [ (" wh ) . Xix Table 4. Summary of Annual Operating and Maintenance Costs for Fuel Recycle in a 1000-Mw(e) MSBR Direct labor $222,000 Iabor overhead 177,600 Chemicals 14,640 Waste containers 28,270 Utilities 80, 300 Maintenance materials Site 2,500 Services and utilities 35,880 Process equipment 160, 040 Total annual charges $721,230 as a function of plant throughput based on the volume of salt processed. The results of these estimates, given in Fig. 6, were used in calculating the nuclear and economic performance of the fuel cycle as a function of fuel-processing rate. For the MSBR(Pa) plant, the processing methods and costs were the same as those for the MSBR, except for blanket-stream processing. The cost of direct protactinium removal from the blanket stream was estimated to be c(Pa) = 1.65R0°45 , (1) where C(Pa) is the capital cost of protactinium-removal equipment, in millions of dollars; and R is the blanket-stream processing rate for prot- actinium removal, in thousands of cubic feet of blanket salt per day. Thus, the cost of fuel recycle in the MSBR(Pa) was estimated to be equiva- lent to the costs given by Eg. (1) and Fig. 6 based on uranium being re- moved from the blanket stream by the fluoride volatility process and the rate of uranium removal being influenced by the rate of protactinium re- moval. ) ORNL-DWG 66-T455 BLANKET STREAM PROCESSING RATE (ft*/day) 2 5 10° 2 ’ 5 10 AL COST OF CORE PROCESSING TING COST OF CORE PROCESSING N CORE STREAM PROCESSING COST (:S7£3) BLANKET STREAM PROCESSING COST (S7ft%) COST OF OPERATING COST OF ——, BLANKET PROCESSING BLANKET PROCESSI CINO 2 5 10 2 5 100 CORE STREAM PROCESSING RATE (ft¥day) Fig. 6. MSBR Fuel-Recycle Costs As a Function of Processing Rates. Fluoride volatility plus vacuum distillation processing for core; fluo- ride volatility processing for blanket; 0.8 plant factor; 12%/yr capital charges for investor-owned processing plant. Fuel-Cycle Performance The objective of the nuclear desigh calculations was primarily to find the conditions that gave the lowest fuel-cycle cost and, then, with- out appreciably increasing this cost, the conditions that gawve highest fuel yield. Analysis Procedures and Basic Assumptions The nuclear calculations were performed with a multigroup, diffusion, equilibrium reactor program, which calculated the nuclear performance, the equilibrium concentrations of the various nuclides, including the fission products, and the fuel-cycle cost for a given set of conditions. 'y xxi The 12-group neutron cross sections were obtained from neutron spectrum calculations, with the core heterogeneity taken into consideration in the thermal-neutron-spectrum computations. The nuclear designs were optimized by parameter studies, with most emphasis on minimum fuel-cycle cost and with lesser weight given to maximizing the annual fuel yield. Typical parameters varied were the reactor dimensions, blanket thickness, frac- tions of fuel and fertile salts in the core, and the fuel- and fertile- stream processing rates. The basic economic assumptions employed in obtaining the fuel-cycle costs are given in Table 5. The processing costs afe based on those given in the previous section and are included in the fuel-cycle costs. A fis- sile material loss of 0.1% per pass through the fuel-recycle plant was applied. The effective behavior used in the fuel-cycle-performance calcula- tions for the various fission products was that given in Table 6. A gas- stripping system is provided to remove fission-product gases from the fuel salt. In the calculations reported here, a *3°Xe poison fraction of 0.005 was applied. Table 5. Economic Ground Rules Used in Obtaining Fuel-Cycle Costs Reactor power, Mw(e) 1000 Thermal efficiency, % 45 Load factor 0.80 Cost assumptions B Value of 33U and 233Pa, $/g 14 Value of 235U, $/g 12 Value of thorium, $/kg 12 Value of carrier salt, $/kg _ 26 Capital charge, %/yr Private financing Depreciating capital 12 Nondepreciating capital 10 Public financing | Depreciating capital 7 Nondepreciating capital 5 Processing cost: given by curves in Fig. 6, plus cost given by Eq. (1), where applicable xxii Table 6. Behavior of Fission Products in MSBR Systems Behavior | Fission Products Elements firesent as gases; assumed to be Kr, Xe removed by gas stripping (a poison - fraction of 0.005 was applied) Elements that form stable metallic colloids; Ru, Rh, Pi, Ag, In removed by fuel processing | Elements that form either stable fluorides Se, Br, Nb, Mo, Tec, 3 or stable metallic colloids; removed by Te, I fuel processing _ Elements that form stable fluorides less . Sr, Y, Ba, Ia, Ce, volatile than LiF; separated by vacuum Pr, Nd, Pm, Sm, distillation , Eu, G4, Tb ' Elements that are not separated from the Rb, C4, Sn, Cs, Zr carrier salt; removed only by salt discard The control of corrosion products in molten-salt fuels does not appear to be a significant problem, and the effect of corrosion products was neglected in the nuclear calculations. The corrosion rate of Hastel- loy N in molten salts is very low; in addition, the fuel-processing operations can control cbrrosion-product buildup in the fuel. The important parameters describing the MSBR and MSBR(Pa) designs are given in Table 1. Many of the parameters were fixed by the ground : rules for the evaluation or by engineering-design factors that include the thermal efficiency, plant factor, capital charge rate, maximum fuel velocity, size of fuel tubes, processing costs, fissile-loss rate, and the out-of-core fuel inventory. The parameters optimized in the fuel- cycle calculations were the reactor dimensions, power density, core compo- sition (including the carbon-to-uranium and thorium-to-uranium ratios), and processing rates. Nuclear Performance and Fuel-Cyclé Cost The general results of the nuclear calculations are given in Table 1; the neutron-balance results are given in Table 7. The basic reactor QEJ ! L3 1] ¥ xxiii Table 7. Neutron Balances for the MSBR(Pa) and the MSBR Design Conditions MSBR(Pa) MSBR Neutrons per Fissile Absorption Neutrons per Fissile Absorption Material Total %:igr:?d Neutrons Total g:sgrb?d Neutrons Absorbed Fi u.lng Produced Absorbed ? u?lng Produced ission Fission 232my, 0.9970 0.0025 0.0058 0.9710 0.0025 0.0059 233p, 0.0003 0.0079 233y 0.9247 0.8213 2.0541 0.9119 0.8090 2.0233 234y 0.0819 0.0003 0.0008 0.0936 0.0004 0.0010 2357y 0.0753 0.0607 0.1474 0.0881 0.0708 0.1721 236y 0.008% 0.0001 0.0001 0.0115 0.0001 0.0001 2375p 0.0009 0.0014 238y 0.0005 0.0009 Carrier salt 0.0647 0.0186 0.0623 0.0185 (except 6Li) 614 0.0025 0.0030 Graphite 0.0323 0.0300 135%e 0.0050 0.0050 149gn 0.0068 0.00692 1539m 0.0017 0.0018 Other fission 0.0185 0.019 products Delayed neutrons 0.0049 0.0050 lost® LeakageP 0.0012 0.0012 Total 2.2268 0.8849 2.2268 2.2209 0_. gaz28 2.2209 aDelayed neutrons emitted outside core. b Leakage, including neutrons absorbed in reflector. design has the advantage of zero neutron losses to structural materials in the core other than the moderator. Except for the loss of delayed neutrons in the external fuel circuit, there is almost zero neutron leak- age from the reactor because of the thick blanket. The neutron losses to fission products are low because of the low cycle times associated with fission-product removal. The components of the fuel-cycle cost for the MSBR(Pa) and the MSBR are sumarized in Table 8. The main components are the fissile inventory XXiv Table 8. Fuel-Cycle Cost for MSBR(Pa) and MSBR Plants® MSBR(Pa) Cost (mill/kwhr) MSER Cost [mill/kwhr(e)] el T oatoray om Rl feriile g gnd Fissile inventory? 0.1125 0.0208 0.1333 0.1180 0.0324 0.1504 Fertile inventory 0.0000 0.0179 0.0179 0.0459 0.0459 Salt inventory 0.0147 0.0226 0.0373 0.0146 0.0580 0.0726 Total inventory 0.188 0.269 Fertile replacement 0.0000 0.0041 0.0041 0.0185 0.0185 Salt replacement 0.0636 0.0035 0.0671 0.0565 0.0217 0.0782 Total replacement 0.071 0.097 Processing 0.1295 0.0637 0.1932 0.1223 0.0440 0.1663 Total processing 0.193 0.166 Production credit (0.105) (0.073) Net fuel-cycle cost 0.35 0.46 ®Based on investor-owned power plant and 0.80 plant factor. Prncluding 23%Pa, 233y, ana 22%U. and processing costs. The inventory costs are rather rigid for a given reactor design, since they are largely determined by the external fuel volume. The processing costs are a function of the processing-cycle times, one of the chief parameters optimized in this study. As shown by the results in Tables 1 and 8, the ability to remove protactinium directly from the blanket stream has a marked effect on the fuel yield and lowers the fuel-cycle cost by about 0.1 mill/kwhr(e). This is due primarily to the decrease in neutron absorptions by protactinium when this nuclide is removed from the core and blanket regions. In obtaining the reactor design conditions, the optimization pro- cedure considered both fuel yield and fuel-cycle cost as criteria of performance. The corresponding fuel-cycle performance is shown in Fig. 7, which gives the minimum fuel-cycle cost as a function of fuel-yield rate based on privately financed plants and a plant factor of 0.8. The de- sign conditions for the MSBR(Pa) and MSBR concepts correspond to the designated points in Fig. 7. 7 Tt has good resistance to oxi- of many austenitic stainless steels. dation by air, and it retains favorable mechanical properties at tempera- tures up to about 1500°F. Results of long-term corrosion experiments (exposures of up to 20,000 hr) have demonstrated its basic inertness to molten fluoride salts at temperatures up to about 1500°F. Corrosion rates appear to be controlled primarily by impurity levels in the molten salts and by the temperature-dependent mass transfer associated with the reac- tion 2UF, + Cr = 2UF3 + CrFp . Based on experimental data from test loops, the corrosion rate of Hastel- loy N in MSBR fuel systems will be less than 0.5 mil/yr with a core outlet temperature of 1300°F, and probably will not exceed that with a 1500°F outlet temperature under equilibrium conditions. Even less corrosion should occur in the blanket-salt and secondary-coolant-salt systems, where the UF, concentration will be extremely low and zero, respectively. These test loop results have been substantiated by data obtained in the MSRE, where no significant corrosion of the Hastelloy N has taken place in 2500 hr of exposure at 1200°F (on the average, chromium was removed from a 1ayer 0.006 mil in thickness over loop surfaces, with v1rtually zero corrosion after the initial months of operation). Extensive tests of the mechanlcal and physical properties of Hastel- loy N as a function of temperature up to about 1800°F indicate charac- teristics suitable for MSBR use. The creep and stress~rupture properties are equivalent to and in most cases superior to those of Inconel. Iong- time ageing studies have shown that the material does not embrittle with *This alloy is commercially available as Hastelloy N or INCO-806; throughout this report, the designation Hastelloy N is employed. 10 time. Further, the mechanical properties of Hastelloy N are virtually unaffected by long-time exposure to the molten fluoride salts. The structural material must retain its good mechanical properties when exposed to reactor radiation. Irradiation studies have shown that the (n,o) reaction in structural materials tends to decrease ductility. This reaction and its effects on Hastelloy N have been studied in detail, and it appears that the deleterious effects can be minimized by maintain- ing a low *°B content, adjusting the concentration of minor constituents in the alloy, and improving heat-treatment practices. Development work in these areas appears dapable of producing an improved Hastelloy N whose ductility will not decrease below aéceptable values during long-term ex- posures to MSBR fluxes. The melting and casting of Hastelloy N can be carried out with the conventional practices for nickel and its alloys. Conventional methods of hot and cold forming have been used to produce it on a commercial basis in a variety of shapes, such as plate, sheet, rod, wire, and as-welded and seamless tubing. Cold working operations can be performed, such as rolling, swageing, tube reducing, and drawing. Cold forming has been successfully used for fabricating Hastelloy vessel heads. The material is readily weldable by the inert-gas-shielded tungsten-arc process. In addition to Hastelloy N, the other prime structural material used in the MSBR is graphite. This material does not react chemically with the molten fluoride mixtures under consideration, and since it is not wetted by molten-salt mixtures, there is little salt permeation of the graphite. 1In general, the graphite needs to have low permeability to salt and gases, to have adequate structural properties when.exposed to high radiation fluxes, and to be fabricated into tubes and other mod- erator shapes. These properties were obtained, at least partialiy, in the MSRE graphite, which was produced by extruding petroleum coke bonded with coal-tar pitch and applying multiimpregnations and heat treatments. The resulting product has a high specific gravity (1.86), low permeation (0.2% bulk volume penetration by molten salt — surface penetrations . only — when a 150-psi pressure was applied to the salt), and high strength (ability to withstand 1500-psi tensile strain and 3000-psi flexural strain was shown by all bars fabricated). This material represents a successful o 11 first step in developing a graphite acceptable for MSBR use. Graphite tubing having l/2—in.—thick walls has also been successfully fabricated: the product had no visible cracks. The graphite in regions of high flux in an MSBR will be irradiated to doses above 10°? neutrons/cm® in five years and will be exposed to radiation flux gradiefits. The magnitude of the graphite differential shrinkage that will occur under these conditions will depend on the graphite creep coefficient, flux gradient, and geometry of the particular structural component. Isotropic graphite has demonstrated the ability to withstand high radiation exposures. Also, the ability of the graphite to absorb the creep strain regardless of the stress intensity has been shown experimentally. Thus it appears that graphite satisfactory for MSBR use can be developed. Techniques are required for attaching graphite to metal with reliable joints. Graphite has been brazed successfully to metals, with brazing alloys that were found resistant fo corrosion by molten salts. Alloys of gold, nickel, and molybdenum and other alloys under development appear to be satisfactory brazing materials. Brazes made with these materials can be used for Jjoining graphite to graphite or graphite to molybdenum (molybdenum has a thérmal expansion coefficient near that of graphite). Metal-to-graphite joints have maintained their integrity in molten-salt enviromments at 1300°F and at pressures of 150 psi for periods of 500 hr. In addition, mechanical joints may be useful in MSBR cores, gince zero leakage between the core and blanket fluids is not required. Finally, compatibility of molten salts, Hastelloy N, and graphite appears excellent. Tests have shown no carburization of Hastelloy N under MSBR conditions. 2.3 TFuel-Processing Development? Experience in processing molten-fluoride-salt fuels at Oak Ridge National ILaboratory dates from 1954 and began with fluoride volatility processing studies. The initial laboratory and development work formed the basis for successful operation of a pilot plant. The associated process is designated the Fluoride Volatility Process after the principal 12 operation of volatilizing uranium as the hexafluoride. Although also applicable to the treatment of solid fuel elements, fluoride volatility processing is uniquely suited to molten-salt fuels because the fuel salt can be treated directly with fluorine. Elemental fluorine reacts with the UF, in the molten salt (at about 930 to 1020°F) to produce volatile UFg. The reaction is rapid and essentially quantitative for uranium; it easily reduces the uranium content of the molten salt to a few parts per million. The UFg product can be treated in absorber beds to give decontamination factors of 10° ahd more. Recycle uranium is easily con- verted to UF, dissolved in carrier salt by absorbing the UF6 in molten salt containing some UF,; and hydrogenating in the liquid phase. This treatment also reduces any corrosion product contaminants to metal that can then be filtered from the fuel solution prior to returning fuel fluid - to the reactor system. The fluoride volatility process can be used for both the core stream and the blanket stream. When applied to the core stream it is used to separate the uranium from the carrier éalt before that stream is pro- cessed (by another method) for fission-product removal. Essentially all the uranium must be recovered, and this leads to relatively severe fluori- nation conditions. Requirements for processing the blanket stream are less stringént. Uranium that is not removed during the fluorination is merely returned to the reactor blanket and is removed during subsequent passes through the processing plant. Discard of 3% annually or process- ing by other methods keeps the fission products at a very low level in the blanket salt. The ease of removal of xenon gas from molten-salt fuels has been demonstrated in both the ARE and the MSRE. It thus appears practical to obtain very low xenon poisoning by sparging the salt with an inert gas 1351, the precursor of 13°Xe, such as helium or nitrogen. In addition, can be stripped from fuel salts by sparging with HF and hydrogen. Such processing would virtually eliminate xenon poisoning in MSBR systems. The discovery that vacuum distillation permits the economic separa- tion of carrier salts from fission products has been a vital factor in improving the economic and nuclear characteristics of MSBR systems. Laboratory experiments have demonstrated that carrier salt can be readily 13 separated from rare-earth fluorides at distillation pressures of 2 mm Hg, with separation factors of 50 to 100 and 95% recovery of carrier salt. These process characteristiés appear adequate for MSBR application. Fluoride volatility processing appears well suited for keeping the uranium inventory and the fission rate in the blanket low and thereby maintaining low neutron leakage from the blanket. An even better process would be one for recovering protactinium directly from the blanket fluid. Recent work toward providing such a process has been encouraging; at least two possible methods are being considered. One involves removal of protactinium from the process stream by precipitation as the oxide through reaction with Zr0O;. After the protactinium decays, the product U0» can be recovered by reaction with ZrF, to give UF,; in solution. Even more encouraging results have been obtained by treating fluoride gsalts containing PaF,; with thorium dissolved in molten bismuth. The thorium metal reduced the protactinium to the metal which subsequently deposited on a stainless-steel-wool filter. These results indicate that inexpensive methods can be developed for removing protactinium directly from the blanket stream of an MSBR. 2.4 Component Development4? ? Nearly all molten-salt component development work has been for ex- perimental molten-salt reactors (the ARE, the planned Aircraft Reactor Test, and the MSRE). The components required for these systems were de- veloped at ORNL, including pumps, seals, valves, heat exchangers, fuel sampler-enricher units, freeze flanges, remote-maintenance tools, heaters, and instrumentation forrmeasuring pressure, fluid flow, liquid level, and temperature under molten-salt reactor conditions. A major effort has been devoted to developing pumps that have long-term reliability at temperatures of about 1300°F. These pumps are vertical-shaft sump-type centrifugal pumps with a free surfaée in the pump bowl; all‘parts wetted by molten salt are constructed of Hastelloy N. Various pump models with capacities up to 1500 gpm have been manufactured and tested, and present models have circulated molten salt continuously for more than 25,000 hr at temperatures above 1200°F without maintenance. Stopping and starting 14 of pumps does not appear to produce any corrosive attack; thermal and pressure stresses associated with thermal cycling and reactor operations do not appear excessive. For MSBR application, it appears feasible to use a vertical sump-type pump similar to present models, with the upper end of the pump shaft supported by oil-lubricated radial and thrust bearings and the lower end supported by a molten~salt-lubricated journal bearing. The present experience with molten-salt-lubricated bearings consists of 3900 hr of operation in development of the bearing and operation for 13,500 hr of a pump containing a salt-lubricated bearing at temperatures of 1000 to 1400°F. The results obtained indicate that the development of salt-lubricated bearings is feasible; testing of these bearings is continuing. | . Molten-salt heat exchangers have been designed and constructed and successfully demonstrated in the ARE and the MSRE. Numerous heat ex- changer designs have been tested, and the results show that the required performance capability and mechanical integrity can be obtained with straightforward design and fabrication methods. The use of Hastelloy N as the construction material introduced no major difficulties. Experi- ments and experience with the MSRE have shown that conventional heat- transfer-coefficient correlations with minor modification are applicable to molten-salt heat exchanger design; also the physical properties of molten fluorides make them good to excellent heat transfer media. Since the molten salts are good fluxing agents and keep all surfaces clean, scale formation does not occur on heat transfer surfaces. An important feature of molten-salt reactors is the ease of adding or removing fuel fluid from the reactor system. This permits ready com- pensation for fuel burnup, and the fluid removed can be easily transported to processing areas. The successful operation of the MSRE sampler- enricher system indicates that adjustments in fuel concentrations can be accomplished readily and reliably with relatively small and simple equipment. | The high melting point of MSBR fluoride salts provides a means of sealing a system, without the need for mechanical valves, through use of "freeze" valves in which a frozen plug of salt prevents leakage from the system. Although slow acting, the performance of freeze valves in the & 15 MSRE has been excellent. It appears that such valves will be useful in MSBR subsystems. Freeze flanges have also been developed because of their proven reliability in containing fluid salts under all anticipated thermal-cycling conditions. Such flanges appear appropriate for Joining components and piping in MSBR subsystems. Instrument development carried out for the MSRE also appears useful for MSBR systems. Liquid-level measuring devices have operated success- fully, as have instruments for fluid flow, differential pressure, and temperature measurements. Development work has also been performed on control-rod drive units capable of operating reliably for long periods while located in a strong gamma field. Since the inception of molten-salt reactors, there has been signifi- ® Remotely cant engineering development work on maintenance operations. operated tools and procedures for remote maintenance have been devised, and the required operations have been studied in a maintenance facility. The results of these studies, along with other experience, were used in developing the MSRE maintenance tools and procedures. Also, equipment for remotely cutting pipes and brazing them back together was developed for replacement of MSRE components, and the results obtained with this equipment indicate that a remotely operated cutter and welder for MSBR maintenance operations is feasible. Experience to date with maintenance . of radiocactive molten-salt systems 1s encouraging. 16 3. INITIAL DESIGN OF A 1000-Mw(e) MSBR POWER STATION The MSBR design discussed here is for a 1000-Mw(e) power station that appears technically sound, maintainable, and attractive from the power cost, reliability, and fuel utilization standpoints. This refer- ence design is not necessarily the best design for a molten-salt reactor, but it represents a logical starting point based on the information available at the time of this study. The report is intended to illus- trate the general merits of molten-salt reactors for power applications, delineate design problems and possible solutions to them, and indicate areas where research and development programs could improve MSBR per- formance. A complete power station is considered, including all major equip- ment and a fuel-processing facility that is integral with the reactor plant. Very little optimization work was done, and layouts and designs were detailed only to the extent necessary to establish feasibility and to permit preliminary estimates of construction and operating costs. The design is based only on those materials and techniques that appear feasible based on present-day technology. In addition, several alter- native molten-salt reactor designs were examined briefly (see Chapt. 5) in order to show the influence of design concept and technology require- ment on the performance characteristics of molten-salt systems. 3.1 General Design Criteria, Cost Bases, and Ground Rules The following design criteria, costs bases, and ground rules were used in making the study: 1. The power station will have a net electrical output of 1000 Mw(e) and will be used solely for the production of power. 2. The reactor will be a two-region two-fluid graphite-moderated and -reflected thermal breeder with graphite separating the fissile and. fertile materials. The reactor will be designed to achieve low power cost, high specific power, and low fuel doubling time. 3. Equilibrium fueling conditions will apply, with mixtures of BeF, and "IiF used as carrier salts for 233U and ThF,. O 17 4., Because of the present uncertainties concerning long-term ex- posure of graphite in a high neutron flux, the MSBR core size will be relatively large in order to reduce the graphite irradiation rate. The fuel cell dimensions will be small to reduce flux gradients in the graphite. The fuel velocity in the core will be limited to 15 fps. The graphite tubes will be attached to a fixed Structure at one end only to give freedom of movement for shrinkage and thermal expansion. Pro- visions will be made for removal and replacement of the core by remote- maintenance procedures. 5. A control rod will be incorporated in the design, primarily as a convenience feature. 6. The reactor core will be arranged so that the fluid will drain by gravity to make the reactor subcritical in event of loss of electric power or other scram-initiating disturbance. 7. The reactor vessel, pumps, heat exchangers, and drain tanks for the fuel- and blanket-salt systems will be housed in a heavily shielded structure. This structure, and the more lightly shielded structure housing all portions of the system containing the coolant salt, such as the boiler-superheaters and reheaters, will be housed in a shielded con- tainment vessel that meets acceptable leak-rate standards for this ser- vice. This containment vessel will incorporate a pressure-suppression system. The reactor containment vessel, but not the turbine room, will be located in a confinement-type building with controlled air-cleaning and venting systems. 8. Heat will be transported from the primary heat exchangers to the steam-power system by a circulating secondary coolant that must be compatible with the fuel- and blanket-salt systems in case of accidental mixing. This coolant must have suitably low vapor pressure and liquidus temperature. 9. The salt pumps will be limited in size to about 15,000 gpm; that is, they will be about an order of magnitude larger than the fuel- salt pump used in the MSRE. 8 10. The reactor system will incorporate an off-gas system for con- tinuous removal, retention, and disposal of the fission-product gases. 18 11. The fuel and blanket salts will be continuously processed in a processing facility that is an integral part of the reactor plant. In the initial design, the fluoride-volatility—vacuum~distillation pro- cesses will be used for the fuel salt, and the fluoride-volatility pro- cess will be used for the blanket salt. A system will be provided for cleanup of the coolant salt. 12. An afterheat removal system will be included in the design. 13. The core outlet temperature of the fuel salt will be 1300°F. The temperature of the coolant salt entering the primary heat exchangers will be above the liquidus temperatures of the fuel and blanket salts. The feedwater entering the boiler will be above the liguidus temperature of the coolant salt. The temperature of the steam entering the reheaters will not be more than 50°F below the liquidus temperature of the coolant - salt. 14. The cells in which the fuel and blanket salts will circulate will be maintained above the liquidus temperature of both salts (ébout 1040°F). The cells in which only coolant salt is circulated will be operated above the liquidus temperature of the coolant (about 700°F). The cell temperatures will be maintained by radiant heating surfaces. Thermal insulation and water cooling will be applied as required to pro- tect concrete, equipment supports, instrumentation, and other items. 15. The boiler will operate with supercritical-pressure steam in a once-through counterflow arrangement. 16. The steam~power cycle will operate with 3500-psia 1000°F steam to the turbine throttle, with single reheat to 1000°F. 17. A1l salt-containing portions of the system will be constructéd of Hastelloy N. The allowable design stress will be 3500 psi at 1300°F, 6000 psi at 1200°F, etc., in accordance with the MSRE design literature® and Ref. 2. 18. All portions of the system will conform to the applicable por- tions of the ASME Codes. Specifically, points of suspected high stresses will be examined for pfacticality of the proposed concepts. 19. All major equipment for the plant will be included in the study up to, but not including, the station high-voltage output transformer 19 and the switchyard. Iand and site development costs will be the same as those used in the advanced-converter reactor studies.??*0 20. Both capital and power production costs will, where applicable, 11 In- be estimated and presented in accordance with the AEC cost guide. direct and operating costs will be estimated on the same hases as those used in the advanced-converter reactor studies.? The plant life will be 30 years. Power costs will be estimated on the basis of both private financing (12% fixed charges) and public financing (7% fixed charges), with private financing as the base case. A plant factor of 80% will be assumed for both cases. In estimating all costs, it will be assumed that equipment and materials are obtained from a large and established molten~-salt reactor industry. 21. The reactor-plant financing rate will apply to the fuel-pro- cessing and -fabrication plant, which will be a part of the power plant. To account for a higher equipment replacement rate, the indirect costs for the fuel-recycle plant will be 100% of the direct costs. 22. Inventory charges on fissile, fertile, and carrier-salt inven- tories will be computed with a reference value of 10% per year for the base case and with 5% per year to represent public ownership. 23. The value of core and blanket fluids will be based on the following: 223U and ?33Pa at $14/g, 22°U at $12/g, Th at $12/kg, and carrier salt at $26/kg. 24. Losses of materials through fuel recycle will be based on uranium losses of 0.1% per pass, thorium and blanket-carrier-salt dis- card on a 30~year cycle time, and core-carrier-salt losses plus discard of 6.5% per fuel-cycle pass. 3.2 General Plant ILayout The MSBR site is that described in the AEC handbook for estimating costs!? and also used in the advanced-converter reactor studies.? In brief, the site is a 1200-acre plot of grass-covered level terrain ad- jacent to a river having adequate flow for cooling-water requirements. The ground elevation is 20 ft above the high-water mark and is 40 ft above the low-water level. A limestone foundation exists about 8 ft 20 below grade. The location is also satisfactory with respect to distance from population centers, meteorological conditions, frequency and in- tensity of earthquakes, and other external conditions. As shown in Fig. 3.1, the plant area proper is a 20-acre fenced-in area above the high-water contour on the bank of the stream. The usual cooling-watér intake and discharge structures are provided, along with fuel-oil storage for a startup boiler, a water-purification plant, water- storage tanks, and a deep well. This plant area also includes radiocactive waste-gas storage, treatment, and disposal systems. Space is provided for the output transformers and switchyard. A railroad spur serves for the transportation of heavy equipment, and parking lots are provided. A large single building houses the reactor and turbine plants, offices, shops, and all other supporting facilities. This building, as shown in Figs. 3.2 through 3.5, is 250 £t wide and 528 ft long; it rises 98 ft above and 48 ft below grade level. The construction is of the typical steel-frame type, with steel roof trusses, precast concrete roof slabs, concrete floors with steel gratings as required, and insulated aluminum or steel panel walls. The wall joints are caulked or otherwise sealed on the reactor end of the building. The reactor complex occupies less volume than the steam-generating equipment in a conventional plant, and the turbine floor dimensions are the same as those used in the Bull Run Steam Plant of the Tennessee Valley Authority (TVA), but there are slightly larger allowances for the shops, offices, control rooms, and other facilities of the reactor plant. The reactor end of the building is 168 ft long and consists of a high-bay portion above a reinforced-concrete reactor containment struc- ture. A single crane is pictured as serving both the turbine room and the reactor plant, but separate cranes would probably be required, and the cost estimate allows for two units. The reactor plant building is sealed sufficiently for it to serve as a confinement volume in the un- likely event of a radioactivity incident, and it is provided with posi- tive ventilation, air filtration and dilution equipment, and an off-gas stack. | | The arrangement of the reactor plant cells is shown in Fig. 3.6. The thicknesses of concrete required for shielding against reactor ORNL DWG 66-7133 RIVER FLOW DISCHARGE \ ELEV. 0' NORMAL RIVER LEVEL INTAKE 20" 60 80’ FUEL OIL STORAGE PUMP WASTE STORAGE POND STATION O WASTE GAS CHARCOAL ADSORBER TER PURIFICATION WASTE GAS TREATMENT & STACK TER STORAGE TE STORAGE TRANSFORMERS PARKING REACTOR PARKING Fig. 3.1. MSBR Plant Site Layout. e 22 ORNL DWG 66-7138 528" D ' c ROLL-UP —_| j _Tll 1 DOOR HOT SHOP ELECTRICAL&EQUIPMENT 1 8 35 STORAGE BUILDING VENTILATION SYSTEM i__ ——————— o] TURLB‘I)NE g 5 , | L e d | | ‘ | REACTOR | 130' i | | weate | oo ‘ ]' TURBINE OFFICES, FEEDWATER SYSTEM 30 TREATMENT ROOM, 4 LUNCH ROOMS, & ETC i SHOPS ' 3 FLOORS HIGH ' \ . 55 68 x 85' 336 x 55 x 30 HIGH od cd Fig. 3.2. MSBER Building — Plan AA. Fig. 3.3. MSBR Building — Elevation BB. ORNL DWG 66-7137 :52' ABOVE NORMAL RIVER LEVEL 3 LYl ' 23 ORNL DWG 66-7140 r —250 — ‘ 55' T3O'T'H - -130' 35';1 | 198’ 2-i50 TON OVERHEAD CRANES CRANE BAY TUNNEL Fig. 3.4. MSBR Building - Elevation CC. ORNL DWG 66-7130 CRANE BAY Fig. 3.5. MSBR Building — Elevation DD. radiations were estimated on the basis of previous reactor design experi- ence. A minimum of 8 ft of high-density concrete separates the reactor vessel from an occupied area. A minimm of 4 ft of concrete is used around equipment containing the coolant salt, which is at a relatively low level of activity durifig reactor operation and this level decreases a short time after power shutdown. As illustrated in Fig. 3.6, the reactor vessel is housed in a cir- cular cell of reinforced concrete. This cell is about 36 ft in diameter and 42 ft high. The four fuel- and blanket-salt primary heat exchangers and their respective circulating pumps are placed around the reactor. 24 ORNL DWG 66-7111 REHEAT STEAM H.P. STEAM “WASTE GA mh FEEDWATER I FEEDWATER CELL nn H.P. STEAM , o LP. STEAM - 28 COOLANT SALT - FUEL M PUMPS HEAT EXCH. i DECONTAMINATION: 16 SUPERHEATERS AND STORAGE BLANKET HEAT EXCH. “.CONTROL AREA—~ ORNL DWG 66~7110 CONTROL ROD DRIVE COOLANT SALT FUEL CIRCULATING— PUMPS—\\\L PUMP e BLANKET CIRCULATING PUMP SUPERHEATERS GROUND LEVEL FUEL HEAT EXCH. KET , REHEATERS HEAT EXCH. 3.6. Reactor and Coolant-Salt Cells — Plan and Elevation. 25 The wall separating the reactor cell from the adjoining cells is 4 ft thick, and the removable bolted down roof plugs total 8 ft in thickness. The pump drive shafts pass through stepped openings in the special con- crete roof plugs to the drive motors, which are located in sealed tanks pressurized above the reactor cell pressure. The special roof plugs are removable to permit withdrawal of the pump impeller assemblies for mainte- nance or replacement. The control-rod drive mechanism passes through the top shielding in a similar manner. The coolant-salt pipes passing through the cell wall have bellows seals at the penetrations. The cells are lined with 0.25- to 0.5-in.-thick steel plate haVing welded joints, which, together with the seal pan that forms a part of the roof structure, provide a cell leak rate that meets the requirement of less than one volume percent per 24 hr. The reactor cell is heated to about 1050°F by radiant heating surfaces located at the bottom. The heat is supplied either electrically or from gas-fired equipment. The liner plate and the concrete structure are protected from the high temperature by 6 in. or more of thermal insulation and cooled by either a circulating- gas or water-coil cooling system. The reactor and heat exchanger support structures are also cooled as required. The four circuits that circulate cooling salt are housed in indi- vidual compartments, or cells, having 4-ft-thick reinforced concrete walls and bolted down removable roof plugs. Each compartment contains four boiler~superheaters, two reheaters, one coolant-salt pump that serves the boiler-superheaters, and one coolant-salt pump that supplies the reheaters. A1l pipes that pass into these cells from the turbine plant have sealed penetrations and valves outside the walls. The pump drive shafts extend through the roof rlugs, and the cells are sealed and heafied in the same manner as the reactor cell. The temperature is only maintained above 750°F, however. The design pressure for the reactor cell and the four adjoining com- partments is aséumed to be about 45 psig. Pressure-suppression systems are provided, with the reactor cell system.beifig separate from the sys- tems for the other compariments. These systems consist of water-storage tanks through which vapor released into a cell would pass and be condensed to maintain the cell pressure below the design value. 26 As indicated, the reactor plant structures have not been optimized nor have they been studied in any detail. Likewise, the cell heating and cooling systems, the pressure-suppression systems, and the building ventilation, filtration, and air-disposal systems have received no de- | tailed study. However, the allowances made in the cost estimates for | these items should not require adjustments large enough to affect the overall conclusions drawn from this study. The turbine plant is standard in the utilities industry and needs little description. Space has been allowed for offices, control rooms, shops, storage, change and locker rooms, and other facilities. 3.3 Flowsheets and General Description The general flow arrangements and operating conditions of the MSBR power station at rated output are summarized in the flowsheet presénted in Fig. 3.7. The 2225-Mw(th) reactor consists of a vessel about 15 ft in diameter and 19 £t high that contains a 10-ft-diam core made up of 534 graphite fuel tubes, which are fastened to two plenum chanmbers at the bottom of the reactor vessel. As shown in Fig. 3.8, fuel salt is pumped into one plenum, flows upward through eight 0.53-in.-diam passages in each graphite tube to the top of the reactor core, and turns downward to flow through the central 1.5-in.-diam passage to the other plenum at the bottom of the vessel. The graphite tube construction is indicated in Fig. 3.9. A matrix of hexagonal graphite blocks surrounds the fuel tubes § and serves as moderator. A 1.5-ft-thick annular'space filled with the fertile, or blanket, salt surrounds the core. Outside the blanket volume is a 3-in. thickness of graphite that acts as a reflector. A 1.5-in.- wide space separates the graphite reflector from the wall of the re- actor vessel; the vessel wall is 1.5 in. thick and is-constructed of Hastelloy N. . The fuel salt is pumped into the reactor plenum at lOOO°F and about 144 psig at a rate of about 95.7 cfs (43,000 gpm).' It flows upward through the fuel tubes and then downward through the central passage, as described above, at an average velocity of about 15 fps. During its passage through the core, the fuel salt is heated to about 1300°F by CRNL DWG 66-7175. REACTOR VESSEL 5.134 # - 5500~ 1000°F - 10.067# E _ /7.I52#4_ i|5l8.5h-540p-I000' p-looo® f ) ”' - ' 5183, . i 11424 h-3615-1000° i BLANKET SALY FUEL SALT | — 3 o - | F | o1 [ GEN PUMPS PUMPS . 3600p-1000°F | bogy 1 | HP | ' . | | IISOlt'Isec 201 Y i'm' | g i Luraing [1'P TURBINE S 52-2"20”““ 173 ft¥sec | 95.7 #Ysec. I , BOILER | P 600p- : . . (150 J; _1000°F | | REHEATERS SUPERHEATERS 1 2sean | 4 B T, 1250°F T300°F j850°F | | j8sor - r r A BLANKET SALT .—’ COOLANT SALT COOLANT SALY B GEN. | 4| HEAT EXCH. iF PUNPS PUMPS 507.7 Mwe I Yy 250 FUEL SALT.; K B ,wfj-,s;c ; _ . Gross - . HEAT EXCH. L)} Tiae BT | | P | | ‘ ¥ o T PREHEATERS CONDENSER 8 FEEDWATER I 1 ] ) . : ! i I ' | b _J 38009-7100'F gy | DOTEH SYSTEMS . i i ! . - 3500p-866"F ) i . JL ' WZ5°F "'"‘1’""—" T sk 4 L 1 T [P0 [% | u) i:l ~ - 570.-650.'F§ -—Q WISp-695F 1 —~ - - ' 766.4h B00STER MIXING TEE | | . . PumPps BLANKET SALT FUEL SALT - { COOLANT SALT \ | FERFORMANCE DRAIN TANKS / ~ DRAIN_TANKS /- _DRAIN_TANKS | NET OUTPUT LO0O' Mue LEGEND P GROSS GENERATION 1,034.9 Mwe FUEL , - BF BOOSTER PUMPS 9.2 Mwe BLANKET e m et _ STATION AUXILIARIES 25.7 Mwe COOLANT ———— REACTOR HEAT INPUT 2225 Mwt STEAM ————— NET HEAT RATE 7,601 Btu/kwh ho NET EFFICIENCY 449 % 2 ok 10° 1b/he [ I psio Moo Btw /1D, —-______Freeze Volve Fig. 3.7. MSER General Flowsheet. L COOLANT— SALT TO SUPER HEATERS COOLANT SALT RETURN e . . . : S PRIMARY HEAT EXCH. CONTROL. f; ROD DRIVE £ i A N “ ‘:!.. - 36 FT. DIA, REACTOR CELL HEAT EXCH. SUPPORTS Fig. 3.8. Reactor + ¢j'-*}. i '::: G w;,fisi@&f?««’ R _ PRIMARY | , HEAT EXCH. | ! t v 1 'i o M —REAC TR PEDESTAL _ 28 ORNL DWG 66=-6968 FUEL PUMP BLANKET PUMP MOTOR CONTROL ROD MOTOR DRIVE ;‘ ;.‘.’. 4 "" . ._“\-, .‘:“ ' "‘._-' ‘.. ? '. . Vl. ~a: . N : . .". . . \.' u '.'“" ‘ :" .‘ s 0 " /\ y "‘.." . i' ' ‘ -fi consTANT | e R R G BN S SUPPORT S - ’ —f HANGERS | o | ; ] e O ;‘,‘ ,‘; Sl L ORNL DWG 66-6967 L » e ? — —— b te=mo LD W g i e " U ‘ ’ * , . , * 5 ‘ ; * <\ ' o - ; ‘ . Sy | » . B T |~ BLANKET S 4 lff . ! : HEAT EXCH o, ; 1 ' ¢ 4":.2‘, : o \ I N | ' BLANKET R N Y FUEL DUMP Ho HEAT g e | } i I EXCH. , : ¥ ; ; 7/ STEAM Cogtfnng N 'i : ‘ . ‘ N / . GENERATOR w - § ’ e CELL COILS FOR ~ T ’ D TR @ AFTER HEAT ™0 F7. D — BLANKE TS ~ : REMOVAL. — i | CORE ‘ REGION J \ ‘ . 0 . —CORE A @ 420 14+ T—REACTOR = REGION (1 . VESSEL 4.1 [ - T~—FUEL SALT | DISTRIBUTION PLENUMS REACTOR CELL Cell — Ej;evation and Plan. @ - L & ;"‘ P Vi 4 . fos o . AR . p a ;Ej‘* ; » Y 4 < K. 3 ' “ g § wal. % o | . * Q‘ - . ey e » ] Y 24 + A v s s ¥ s . T e LTIt | T e T T e o 4 1 - T AL . = P A IO E M G ST VRTINS T e . Ta " Ta At e a . . & « . . . « . e - . * b . . T e s . S RS L e N + - . . 1 ki g —tecnp 29 ORNL DWG 66-7139 MODERATOR GRAPHITE) —BLANKET PASSAGE FUEL PASSAGE {UP) FUEL PASSAGE (DOWN) 3% OD. FUEL TUBE : 1 : MODERATOR HOLD =48 PIT 8 PTCH DOWN NUT GRAPHITE) { N | REACTOR —SPACER AL TO GRAPHITE SLIP-JOINT METAL TO GRAPHITE BRAZED JOINT BRAZED JOINT ~——FUEL INLET PLENUM ~———FUEL OUTLET PLENUM Fig. 3.9. Graphite Core Tubes. 30 nuclear fission. About 95% of the heat generated, or 2114 Mw(th), is removed by the fuel salt. Concentric pipes connect the core plenum chambers to the heat ex- changers. The 1300°F fuel salt leaves the lower plenum of the reactor vessel and flows downward through the 18-in.-diam inner pipes to the top of the heat exchangers, where the pressure is about 96 psig. The fuel salt is circulated in four loops that operate in parallel. Each loop contains a vertical shell-and-tube heat exchanger about 5.5 ft in diameter and 18 ft high, with a fuel-circulating pump mounted on the tdp. Each pump impeller operates in a bowl that is integral with the ‘ shell of the associated exchanger. Above each bowl and connected to it by open passages is a salt storage volume of about 45vft3, which is suf- ficient to store about one-fourth of the fuel salt needed to fill the reactor core. The general afrangement of the four blanket heat ex- changers and the four fuel heat exchangers around the reactor is shown in Fig. 3.8. | In the heat exchanger, the fuel salt flows downward at about 11.3 fps through the outer row of 0.375-in.-diam tubes into the lower head, where the salt conditions are about 1170°F and 51 psig. It then flows upward at about 13 fps through the 0.375-in.-diam innermost tubes to the bottom of the pump bowl, where the conditions are approximately 1000°F and 5 psig. The pump discharges through the annular flow passage between the 18- and 24-in. concentric pipes, and the salt returns to the reactor plenum to repeat the cycle. Each pump is rated at 11,000 gpm at a 150-ft head and requires a 1250-hp motor. The blanket salt is pumped into the reactor vessel at about 1150°F at a flow rate of about 17.3 cfs (7700 gpm). The blanket salt flows downward through the space between the graphite reflector and the re- actor vessel to cool the wall and the top head of the vessel, and then flows upward through the blanket volume and the interstices of the core lattice (the blanket salt occupies about 7% of the core volume). About 111 Mw(th) is deposited in the blanket salt as it passes through the re- actor, and it leaves the reactor at about 1250°F through the inner pipe of the 8- and 12-in.-diam concentric pipes. / 31 The blanket salt is cooled in four circulating loops in a manner similar to that used for the core salt. The blanket salt flows downward through 0.375-in.-diam tubes in a 3-ft-diam, 9-ft-high vertical shell- and-tube heat exchanger at about 10.5 fps. After passing through the lower head, the fluid flows through another section of 0.375-in.-diam tubing at about the same velocity and enters the pump bowls at about 1150°F. (These pumps do not have the large storage volume above the bowls.) ZEach of the four blanket-salt pumps has a capacity of about 2200 gpm at a 150-ft head and uses a 500-hp motor. The salt flows to the reactor through the annular region between the 8- and 12-in.-diam concentric pipes connecting the heat exchanger and blanket volumes and repeats the above cycle. The volumes above the four fuel pump bowls have a combined capacity sufficient to hold all the fuel in the reactor core. Since the reactor is at a higher elevation than the fuel pumps, stoppage of the pumps will cause the salt to drain from the core by gravity. It is estimated that the reactor would become subcritical in 1 to 1.5 sec. Loss of one pump would also cause the core to become subcritical because of salt drainage. Thus all blanket and fuel-salt pumps need to be operative for the re- actor to generate power. {An alternate modular design, discussed in Chapter 4, permits partial power generation even though a fuel pump fails.) Afterheat generated in the salt stored in the volumes above the pump bowls is removed by coils through which a coolant is circulated. Salt remaining in the heat exchangers, piping, and reactor plenum cham- bers is circulated through the exchangers by a gas 1lift to permit after- heat removal from these volumes. The gas 1lift is provided by helium, which is normally introduced continuously at the bottom of the heat ex- changer to purge fission~product gases from the fuel salt. The fission- product afterheat is transferred to the coolant salt, which will circu- late through the primary exchangers by thermal convection and in turn transfer energy to the steam cycle. The fuel-, blanket-, and coolant-salt systems are provided with "ever-safe"” tanks for storage of the salts when the systems are drained for maintenance or other purposes. The drain valves for these lines 32 have not been specified, but they could possibly be freeze-type12 or ..mechanical valves developed for salt service. As indicated above, fission-product gases such as xenon and krypton are sparged from the fuel-salf circulating system by introduction of helium in the bottom head of the heat exchangers. The off-gas system and flowsheet for the handling of these radioactive gases are described in Section 3.6. , A helium system provides cover gas for fhe pump bowls, drain tanks, fuel-handling and -processing systems, and other equipment. This system is briefly described in Section 3.6. For processing pfirposes, small side streams of fuel salt (about 14.5 £t3/day) and blanket salt (about 144 ft3/day) are taken from the main circulating loops and sent to the fuel-processing plant located in cells adjacent to the reactor proper. The fuel-recycle system and its flowsheet are described in Section 3.5. An intermediate coolant salt is utilized to transfer energy from the primary circuit to the steam cycle. The coolant salt is pumped through the shell sides of the four fuel-salt heat exchangers and then through the four blanket~salt exchangers by a total of eight pumps. Four of these, each rated at 14,000-gpm capacity at a 150-ft head (1250- hp motor), pump the coolant salt through the 16 boiler-superheaters. The other four, individually rated at 2200 gpm at a 150-ft head (200-hp motor), pump the coolant salt through the eight steam reheaters. The coolant salt enters the shell side of each of the fuel-salt ex- changers at about 850°F and at a rate of 37.5 cfs. The salt is thus above the 842°F liguidus temperature of the fuel salt. The coolant salt flows across the tube bundle, as directed by the baffles, to the exit at the bottom. It then enters the top of the shell side of the blanket-salt heat exchangers at about 1111°F, which is above the 1040°F liquidus tem- perature of the blanket salt. It leaves the bottom of the shell at about 1250°F, | | About 87% of the coolant-salt flow, or about 32.5 cfs for each of the large coolant-salt pumps, supplies a total of 1931 Mw(th) of heat to the boiler-superheaters. The remainder of the flow, or about 5 cfs for each of the small coolant-salt pumps, supplies about 293 Mw(th) of 33 heat to the steam reheaters. The coolant salt exits from the heat ex- change equipment at 850°F. The coolant salt enters the 16 vertical U-shell-and-tube heat ex- changers, which serve as the boiler-superheaters, at the top of one leg at a temperature of about 1125?F. Tt passes downward through the 18-in.- diam baffled shell and upward through the other leg of the shell to emerge at 850°F. The high-purity boiler feedwater, at about 700°F (the estimated liguidus temperature of the coolant salt) and 3800 psia, is introduced at the tube sheet at the top of one leg, flows through the 1/2-in.-diam tubes, and exits at the top of the other leg as steam in a once-through arrangement. The steam leaves the units at 1000°F, 3600 psia, and a total rate of about 10,067,000 1b/hr. As shown in the steam system portion of Fig. 3.7, about 7,152,000 lb/hr of the steam enters the throttle of the high-pressure turbine at about 1000°F and 3500 psia. About 5,134,000 1b/hr leaves this turbine at 552°F and 600 psia and flows to the eight U-tube vertical shell-and- tube heat exchangers, which preheat the "cold" steam before it enters the reheaters. It flows through the 20-in.-diam shells and is heated to about 650°F by about 2,915,000 1b/hr of the 1000°F throttle steam, which flows through 0.375-in.-diam tubes. The supercritical steam leaves the tubes at 866°F and 3500 psia and is mixed with the 552°F 3500-psia feedwater from the No. 1 feedwater heater in the regenerative steam cycle to give a fluid temperature of about 695°F. The water is then boosted in pressure to 3800 psia and raised in temperature about 5°F by two par- allel 20,000-gpm 6200-hp motor-driven pumps. This produces the 700°F feedwater for the boiler—superheatefs, as mentioned above. The 650°F reheat steam from the preheaters flows through the tubes of the eight vertical straight-tube 28-in.-diam shell-and-tube heat ex- changers, which serve as the steam reheaters. The tubes in these units are 0.75 in. in diameter. The heat source for the reheaters is the 1125°F coolant salt mentioned above, which raises the ten@erature-of the steam to 1000°F. The steam returns to the double-flow intermediate- pressure turbine at about 540'psia; this turbine is on the same shaft as the high-pressure turbine. These two 3600-rpm prime movers drive a gen- erator on the same shaft to give a gross electrical output of 527.2 Mw. 34 The steam leaves the intermediate~pressure turbine at about 172 psia and 706°F and crosses to the 1800-rpm four-flow low-pressure'turbine, where it expands to 1.5 in. Hg abs and produces 507.7 Mw gross electrical power. The regenerative feedwater heating system empioys eight stages of feedwater heating, including the deaerator, and_two-tufbine-driven boiler feed pumps. Condensing water, boiler makeup, and condensate-polishing systems are also included. The gross electrical generatlon of the plant is 1034.9 Mw; the net statlon output is 1000 Mw(e). The overall net thermal efficiency is 4de 9%, 3.4 Reactor System 3.4.1 Description Top and sectional views of the reactor wvessel and core are shown in Figs. 3.10 and 3.11. Pertinent data on the reactor system are summarized in Table 3.1. The reactor vessel is about 14 ft in diameter and has an overall height of about 19 ft. It is constructed of Hastelloy N; it is designed for 1200°F and 150 psi; and it has walls 1.5 in. thick. The torospheri- cal heads are 2.25 in. thick. The bottom head is an integral part of the vessel, but the top head is arranged for grinding away the weld so that the head can be removed. The vessel is supported on reinforcing rings in the bottom head that rest on a structural steel stand mounted on a reinforced-concrete pedestal in the center of the reactor cell. As shown in Figs. 3.10 and 3.11, the fuel salt enters and leaves the reactor through four concentric pipes (diameters of 18 and 24 in.) in an arrangement that tends to minimize the stresses due to temperature differences. These pipes communicate with plenum.chambers in the bottom head of the reactor vessel. The fuel salt flows through the annular pas- sage between the two pipes and enters the outer plenum chamber. It then flows upward through the fuel-salt passages to the top of the reactor and downward to the inner plenum chamber, where it leaves through the 18-in.-diam pipe. 3 ] Ac—_1 BLANKET RETURN ORNL DWG 66-7128 CONGENTRIC FUEL LINES r..,"' CORE CELL REACTOR HOLD DOWN CLAWPS _ V@QO . O‘ e JV. \ (02610060 (000,00 \\ @@QOO@. .. %)! Q. o “\ REFLECTOR (GRAPHITE) CONCENTRIC BLANKET LINES ’ . ’ P \ ", Fig. 3.10. Reactor — Plan BB. ~I7' 0" T TT T T TTe 3 TO FUEL HEAT EXCHANGER 36 ORNL DWG 66-7129 TO BLANKET HEAT EXCHANGER 3. oll ’ S 10 FT. DIA. CORE Fig. 3.11. Reactor — Elevation AA. “ . - A k1 ) A . j o | |- | ———Core cELL - — !‘-_— | | . : b /—REFLECTOR (GRAPHITE) A ! 1 : g ! —— Bl REACTOR VESSEL ' A z - FUEL SALT Al DISTRIBUTION PLENUMS 1 K | Al _~——BLANKET o’ 4 ~ Volume fractions Fuel salt Blanket salt Graphite moderator Blanket Radial thickness, ft Axial thickness, ft Volume, ft3 Volume fraction, blanket salt Reflector thickness, in. Fuel salt Inlet temperature, °F Outlet temperature, °F Flow rate, ft3/sec (total) 1b/hr gpm Volume holdup, ft3 Core Blanket Plena Heat exchangers and piping Processing plant Total Salt composition, mole % LiF BeF; UF, (fissile) Blanket salt Inlet temperature, °F Outlet temperature, °F Flow rate, ft?/sec (total) 1b/hr gpm Volume holdup, ft3 Core Blanket , Heat exchanger and piping Processing Storage for Pa decay Total 2114 111 2225 14 ~19 2.25 12.5 534 282 0.169 0.0735 0.7575 2.0 1120 1000 1300 95.7 43,720,000 42,950 1lé6 26 147 345 33 717 63.6 36.2 0.22 1150 1250 17.3 17,260,000 7764 72 1121 100 24 2066 3383 38 Table 3.1 (continued) Blanket salt (continued) Salt composition, mole % LiF UF, (fissile) System fissile inventory, kg System fertile inventory, 1000 kg Processing data Fuel stream Cycle time, days Rate, ft3/day Processing cost, $/ft? Blanket stream Cycle time, days Rate, ft?/day Processing cost, $/ft3 Fuel yield, % per annum Net breeding ratio Fissile losses in processing, atoms per fissile absorption Specific inventory, kg of fissile material per megawatt of electricity produced Specific power, Mw(th)/kg of fissile material Core atom ratios 0 Fraction of fissions in fuel stream Fraction of fissions in thermal-neutron group Mean 7 of 233y Mean 7 of ?3°U Net neutron production per fissile absorption (Me) Power density, core average, kw/liter Gross In fuel salt Neutron flux, core average, neutrons/cmz-sec Thermal Fast Fast, over 100 kev Core thermal flux factor, ratio of peak to mean Radial Axial Qverall plant data Net electrical output, Mw Gross electrical generation, Mw Boiler feedwater pressure-booster pump power, Mw(e) Station auxiliary load, Mw(e) Net heat rate, Btu/kwhr Net efficiency, % Assumed plant factor 71.0 2.0 27.0 0.0005 769 260 47 14.5 203 23 144 7.33 4.86 1.0491 0.0057 0.769 2.89 41-7 5800 0.987 0.806 2.221 1.958 2.221 80 473 6.7 X 1014 12.1 X 1014 3.1 X 1014 o 39 The active portion of the reactor core is 10 ft in diameter and about 12.5 ft high. It contains 534 graphite tube assemblies through which the fuel salt flows and around which the blanket salt circulates. Each tube assembly, as shown in Fig. 3.9, consists of a 3.5-in.-0D graph- ite tube with eight 0.53-in.-ID holes regularly spaced on a 2.62-in.- diam circle. The fuel salt flows upward through these eight tubes at about 15 fps. The salt reverses direction at the top of the fuel as- sembly, flows downward at about 15 fps through the 1.5-in.-ID central passage, and enters the inner plenum at the bottom of the reactor. The 3.5-in.-0D graphite tubes are slipped into hexagonally shaped passages inside hexagonal graphite tubes that are approximately 5 in. across the outer flats. Blanket salt circulates in the passages between the circu- lar and hexagonal graphite tubes. Thin portions of each outside face of the hexagonally shaped graphite are cut away, as indicated in Fig. 3.9, to form passages for circulating blanket-salt. The fuel tubes are con- tinuous along their lengths, whereas the hexagonal tubes are made up of stacked graphite pieces. The upward and downward fuel flow passages com- municate at the top of the fuel tube, where a threaded-graphite plug tightly closes the top end of the tube, as shown in Fig. 3.9. This plug is provided with a threaded-graphite stud, washer, and nut assembly for holding the hexagonal pieces in place. Stubs of 4-in.-0D Hastelloy N tubes that vary in length from about 6 to 15 in. are welded to the upper diaphragm in the lower head of the reactor vessel. This diaphragm is about 0.75 in. thick. Metal transi- tion pieces with an outside diameter of 4 in. and a length of about 8 in. are brazed to each of the stubs; previous to this, the 3.5-in.-0D graph- ite tubes for the fuel salt are brazed under carefully controlled shop conditions to shoulders on the inside of the metal transition pieces. The hexagonally shaped graphite tubes rest on top 4-in.-diam by about 4=in.-long metal spacers, which in turn rest on top the metal transition - pleces. Other Hastelloy N stubs, 2 in. OD and varying in length from 8 to 30 in., are welded to the 0.25-in.-thick top of the inner plenum chamber at the bottom of the reactor vessel. These stubs neck down to about 1.62 in. OD at the top and are a sliding fit into the bottom of the inner 40 passage of the graphite fuel tube (the tubes are machined at the bottom end to permit this fit). Any salt leakage through this joint constitutes only a small bypass of the core. | The blanket salt leaves and enters the reactor vessel through con- centric 8- and 12-in. pipes located near the top of the reactor vessel (see Fig. 3.8). The inner pipes of these concentric connections, like those in the fuel-salt system, are provided with slip joints near the heat exchanger nozzles to allow for the relative movement between pipes due to temperature differences. Small leakage through the joints is inconsequential. ' The blanket on the sides and top of the core averages 1.5 ft in thickness. Outside this blanket, and 1.5 in. from the vessel wall and top head, is a 3-in. thickness of graphite which serves as a reflector for neutron economy and also helps to protect the wvessel from irradiation damage. The annular space between the reflector graphite and the wall is a flow passage for the incoming blanket salt; the stream enters at a temperature of 1150°F and serves to cool the vessel wall and the top hesd. | The basic design of the reactor has the advantage of low neutron losses to structural materials other than the graphite. Except for some unavoidable loss of delayed neutrons in the external fuel-salt circuit, there is almost no neutron leakage through the thick blanket. Neutron losses to fission products are minimized by the continuous treatment of a side stream of the fuel salt in a processing plant that is part of the MSBR power station. The nuclear performance is discussed in more detail in Section 3.5. The reactor system described above provides for support of the graphite when the vessel is empty of salts, prevents the graphite from floating during normal operation, and allows for thermal expansion and growth or shrinkage of the graphite. The core can be removed as an assembly after the holddown clamps are unbolted and removed and the seal weld is cut (see Fig. 3.11). The upper plenum diaphragm, which carries the load of the graphite in the reactor core, can then be removed for replacement should this prove necessary. Tools must be developed for seal-weld cutting, joint preparation, and rejoining. The drawings do m i 41 not indicate a means of guiding a new core assembly into position, but this could be readily provided. Replacement of a graphite tube with the core in place may also be practical. This could be accomplished by first cutting off and removing the top head of the reactor vessel to expose the tops of the fuel pas- sage tubes. Removal of the graphite nut at the top of the defective or suspect tube would permit withdrawal of the graphite hexagonal section surrounding the tube. The Hastelloy N spacer at the bottom could then be lifted out to make it possible to lower an induction coil heater and break the metal-to-metal brazed joint between the metal stub and transi- tion pieces. A replacement tube could be installed with a reverse pro- cedure. 3.4.2 Reactor Materials Fuel and Blanket Salts. The chemical compositions of the fuel and blanket salts and the pertinent physical properties employed in the de- sign are shown in Table 3.2. The phase diagrams of these salts and a general discussion of the chemistry, physical properties, and behavior of fluoride salts are given in Ref. 1. The feagibility of the use of these salts in reactors is well established on the basis of many experi- mental studies® and MSRE experience. Table 3.2. Physical Properties of MSBR Fuel, Blanket, and Coolant Salts® Fuel Salt Blanket Salt coolant , Salt Reference temperature, °F 1150 1200 988 Salt components ‘ , - LiF-BeFy=UF, LiF-ThF,-BeF; NaF-NaBF, Nominal salt composition, mole % 68.3-31.2-0.5 71.0-27.0-2.0 61.1-38.9 Molecular weight, approximate 34 o 103 68 Iiquidus temperature, °F 842 1040 700 Density, 1b/ft> 127 277 125 Viscosity, 1b/hr-ft 27 38 12 Thermal conductivity, Btu/hr-ft® (°F/ft) 4 1.5 1.3 Heat capacity, Btu/lb.°F 0.55 . 0.22 0.41 %The values listed are those used in the MSBR heat-power cycle studies to es- tablish heat transfer coefficients, flow rates, etc. Many of the properties are not known with certainty, and a few, such as the viscosity of the cooclant salt, are little better than rough estimates. In addition, the values used for thermal con- ductivity appear at present to be slightly high (Ref. 1). 42 Coolant Salt. The coolant tentatively selected for the MSBR is a sodium fluoroborate salt that appears to be compatible with the materials in the system and with the fuel and blanket salts; it has a liguidus tem- perature of about 700°F and appears to have heat transfer and fluid flow Properties that make it generally suitable for MSBR application. Several of the physical properties shown in Table 3.2 need to be verified but are believed to bé sufficiently accurate for the purposes of this study. Graphite. The MSBR core graphite would be an improved grade of that used in the MSRE (properties of the MSRE core graphite are given specifically in Ref. 2). The MSBR graphite tubes should have no signifi- cant cracks and should be able to withstand high radiation exposures ex- ceeding 10°2 neutrons/cm? (neutron energies above 300 kev). Hastelloy N. The salt-containing portions of the MSBR are fabri- cated of Hastelloy N, since it has excellent compatibility with molten fluorides at high temperatures and under severe radiation conditions. The chemical composition, mechanical and physical properties, and corro- sion resistance of this material are discussed in Ref. 2. The mechanical- property values given in Ref. 2 were used in conjunction with ASME Code requirements in specifying equipment. Although Hastelloy N has exhibited radiation embrittlement when irradiated to MSBR exposures, major improve- ments in the radiation stability of the material can be obtained by minor changes in composition and by modifying the heat treatment.? 3.4.3 MSBR Load-Following Characteristics The negative temperature coefficient of reactivity makes the MSBR independent of the need for control rods for load following.' The pre- liminary nature of this report did not permit a study of reactor safety, but on the basis of MSRE studies 15 and experience,® converter reactor‘safety studies,l7 it appears that the fuel, blanket, and molten-salt and coolant-salt temperatures will be quickly self-adjusting with no oscillations or reactivity perturbations of consequence following changes in turbine-generator load. In recognition of the need to control the _ throttle-steam superheat temperature at 1000°F and the reheat steam tem- perature at 1000°F independently of eaéh other and of turbine load, separate variable-speed coolant-salt pumps were specified for the boiler~ superheaters and the reheaters. 43 3.5 DNuclear Fuel-Cycle Performance It is desirable that the rate of fissile fuel yield be maximized consistent with low fuel-cycle costs. Since two nuclear designs can have about the same fuel-cycle cost but significantly different fuel doubling times, MSBR nuclear design optimization studies were performed to find conditions corresponding to both low fuel-cycle costs and high fuel~yield rate. ' An important feature of the MSBR concept is the fuel-recycle plant, which is an integral part of the reactor plant. Fuel-recycle costs play an important role in determining the rate at which fuel can be economi- cally processed and thus significantly influence the breeding ratio and fissile inventory of the MSBR. In order to properly consider this in- fluence, a detailed design and cost study was made of the fuel-recycle plantl® and is summarized below. The costs obtained, including those for capital and operation of the fuel-processing plant, have been kept separate from the costs of the main reactor plant.¥* This was done in order to show a fuel cost that can be more readily compared with fuel costs of other reactor plants utilizing off-site fuel-recycle facilities, where fabrication and processing charges include such facility costs. 3.5.1 Design and Cost Study of Processing Plant for Fuel Recycle The MSBR core fuel consists of fissile UF, dissolved in an inert carrier salt containing 7LiF4 and BeFp. The blanket salt contains the fertile material, ThF4,.which is also dissolved in a carrier salt con- taining "LiF,; and BeF,. The flowsheet for the MSBR processing plant for recycling the fuel is shown in Fig. 3.12. ' ' The fuel stream is processed by the well-established fluoride vola- tility process to separaté the uranium from the carrier salt and fission - products. The valuable carrier salt is separated from the fare—earth fission products by the vacuum-distillation process; about 6.5% of the *¥An exception to this is the capital cost of the building for the fuel-recycle plant. This has been included with the reactor plant, since the fuel-recycle system is housed within the reactor building. ORNL-DWG 65- 6194 Al UFg RECYCLE TO REACTOR WYl | 7 soaaens/ s e N 000005 °C > /| A V¢ A A | 7 R sl . g e ¥ | PRODUCTION / UFg + Nor MR/ " waste /] (L LS VOLATILE FP i MAKE UP | aF MR/ TP LIF/BeE, /ThE, ok + vz lr VOLATILE FP MAKE UP _ LiF/BeF, ) | fonTinudus Vioots Vegiiscos A A 7 AN e ke Rl AN ik Mo el /| MAKE B , VOLATILITY /] _[FP DECAY |/ voLATILITY |/} DisTiLLATION | - LiF/BeF, /|7 ) REDUCTION || FILTRATION 7/// 7// //// // //// c NN A~s80%C /// ~1000°c [ /|”~500°c /|, /|, 550-600°¢C // // / ~550°C / /”1.5?/ UFq'+F2=UF6 7 {mm Hg //////// /%/ 000 N 0 00 0 . & A A LIF/BeF, / ThE, /FP F— F— LiF + RARE —H; EARTH FP REDUCED METALS 7777 77 7 Cr, Fe. N /DISCARD FOR/ WASTE / - FP REMOVAL [ STORAGE L L L L L2272 / FERTILE STREAM RECYCLE LiF/BeF,/UF4 RECYCLE Fig. 3.12. Fuel- and Blanket-Salt Processing for the MSER. 45 carrier salt is either discarded or unrecovered in the distillation pro- cess in order to control fission-product buildup and reduce recovery costs. | ':.HV - The fuel salt is reconstituted by absorbing UFg in uranium-containing carrier salt”and.feducing it to UF, by bubbling hydrogen through the melt. Excess urénium.from fihE‘reactor is sold as an equilibrium mixture of the fuel isotopes. _ - | The blanket‘saitisproéessed by the fluoride volatility process alone. Any uranium nbt removed during blanket processing returns to the blanket and isvremovedfby‘subsequent processing. Smell side streams of about 14.5 £t3/day of fuel salt and 144 t3/day of blanket salt are contifiupusij withdrdwn from the reactor circulating systems and routed to fihe\fiquéssing plant located within the same build~ ing. The inventoriesuretained in'the'processing plant are estimated to be about 10% of the reactor system fuel-salt inventory. The correspond- ing value for the blanket system is about 1%. An_imporfiant factor affecting both the breeding gain and the fuel cost is the loss of fisSile‘material in processing. There is considerable engineering experiénce_in fluéride volatility prbcessing that indicates an MSBR fissile material loss of 0.1% per pass or less through the pro- cessing plant. Therefore le.l%'loSS*per pass has been assumed in this study. | o | Based on the fuele}ecycle proceSsing schemes indicated above, capital cost studiesl® were made of an MSBR integrated processing plant. The plant throughput was assfimed to be 15vft3/day of fuel salt and 105 ft3/day of blanket salt, with'each stream being treated separately. These through- put rates correspond roughly to the needs of a 1000-Mw(e) station. In performing the proéessing plant cost study,l8 the equipment flow- sheet given in Fig. 3.13 was developed, the required equipment was de- signed, and cost estimates were made for the process equipment and asso- ciated structures. The basic processes considered involve fluorination, purification of UFg, vacuum distillation, reduction of UFe and reconsti- tution of the fuel, off-gas processing, waste storage, flow control of the salt streams, removal of decay heat, provisions for sampling of the salt and off-gas streams, and provisions for shielding, maintenance, and ANV NI i e TP | @compnessoa 46 \‘ Fa.FP Dwg. 58025D fz,UFp“FP o FLUSH AXALATE Sh. ANT WATER T et TR PR 288 “OOLING Py CLIN (R AR Al RS £ LSy ALCTUE L3y SALY t .,.,’;‘;":s'f‘,. - —‘?’@AE'L’ __.‘__J. .Tfliihd @ © ‘ PO W 0 PRCER® | CLNE L _ c\,_,._ffs ' ™" PArCERS ; VESRLLY : ‘ ® cF ! 2.Y¥e. ; 1! FL W ! CONTRAL @ z HEATER £ {@ £ £ (_" ¥ z o Ty < & 2 . H2 MF t i i 1 TO SLANKI® SALT DUMP YANA TRIL TO WASTE STORAGE LiF, ThFy o .u@ur :;: ncézn ®P TRAP REFEIG@I‘H ON H2 REFRIGERATIN —3 | f ! £2 o F, FP €5, Ufg FT urn , D ;{t \ : ey . - CQQLING uk%fl Mg Fz..l " ® 9 wo M Irp rap 3 = Ve ) o @"; | g I A 5 o : %5:& z o . ; 2 ¥, 2 . ; w ! " ) 3 | | ' > T WA TE ‘ STORMGE ! LIF.BeFp CORE, BLANKET 1 | fr: i ! : LJ w e 'u@-uv o g/' rLow ¢ HE@ER up SURGE T Tank " FLAW SCONTROL q FLOW FPFo IiF CONTRL NAF WA~-TF TANKT AND -J l CoALING STITIN SULL TC WASTE ™ x % STARLGE L——@—»u&kfi FiLTER TRAP & FLOW CONTROL Hp MF,UFg uamro I K 2ot SCRUBBING : COLUMN M#xz;up P_U‘!JiPiBD . o 30 £ e L . VACUUM PUMP - ufe \ 10 WASTE - NS,UF, STORAGE l . _ €& v ; :: HE@ER} 2z | 13 & | £7 3 = i }-‘-‘.'.':5— F N ' HEATER '6 KOH,FP T WASTE STORAGE pésp - © REFFIGTRA™: "N STRVERS UFg 47 repair of equipment. Based on these considerations and associated opera- tions, a total direct capital cost for the plant was obtained along with. a. direct operating cost. From these results, and consideration of in- direct costs, the total fuel-recycle processing costs were obtained. The major novel pieces of processing equipment are the fluorinator, fuel reduction equipment, and distillation unit. The designs considered are shown in Figs. 3.14, 3.15, and 3.16. Figure 3.14 illustrates the fluorinator, which utilizes a frozen wall of salt and performs continuous fluorination of a flowing stream of uranium-containing molten salt. The NaK coolant flowing through the jacket, as shown, freezes a layer of salt on the inner surface of the column to protect the structural material (alloy 79-4) from corrosive attack by the molten-salt—fluorine mixture. Figure 3.15 illustrates the equipment for reducing UFg to UF,. Barren salt and UFg enter the bottom of the column, which contains circulating LiF-BeF5-UF,. The UFg dissolves in the salt, aided by the presence of UF,, and moves up the column, where it is reduced by hydrogen. Reconsti- tuted fuel is taken off the top of the column and sent to the reactor core. Figure 3.16 illustrates the design of the vacuum-distillation unit. Barren fuel-carrier salt flows continuously into the still, which is held at about 1000°C and 1 mm Hg. LiF-BeF, distillate is removed at the same rate that salt enters, and thus the volume is kept constant. Most of the fission products accumulate in the still bottoms and are drained to waste storage when the heat-generation rate reaches a prescribed limit. The fuel~-recycle processing plant is located in two cells adjacent to the reactor shield; one contains the high-radiation-level operations and the other contains the lower radiation-level operations. Each cell is designed for top access through a removable biological shield having s thickness equivalent to 6 ft of high-density concrete. Both cells are served by a crane used in common with the reactor'plant. Process equip- ment is located in the cell for remote removal and replacement from above. No access into the cells with be required; however, it is possible with proper decontamination to allow limited access into the lbwer radiation level cell. A general plan of the processing plant and a partial view of the reactor system are shown in Fig. 3.17. The highly radioactive opera- tions involved in fuel-stream processing are carried out in the smaller 48 ORNL DWG €5-3037 Fig. 3.14. Continuous Fluorinator for Salt Processing. ORNL OWG. 65-3036 Fig. 3.15. Fuel-Reduction Column for Salt Processing. LiF-BeF2 PRODUCT i Z, NaK OUTLET -« 817 TUBES fe———————34" |D—————=- 50 ORNL DWG 65-1802R2 AR COOLANT ;—\ /\ 770N ]lil’é b o T B I l l AIR OUTLET 1/2" x 16 GAUGE 1/2" INOR-§ TUBE SHEET FUEL SALT INLET Fig. 3.16. N INOR-8 1/2" WALL amal) —+ NaK COOLANT LiF WASTE DRAIN 30" ID——e- Vacuum Still for Carrier-Salt Recovery. 51 ORNL-DWG 66-7459 v . C 't - : v 4 v ’ , ¥ PROCESSING < " CELLS - AR-NaK - R HEAT 4 ; EXCMANGER ¥ 3 L » » = - b v v KN Y ep Ve #F i, T v r Ly e o . b .‘..,_m,......___z.z.._g___‘ TS o A "Tg. S R NP e T A e i PN d & L Nof -MgF2 WASTE CANS SR N ,. JReacTor MAKEUP § CELL AREAN B g WL VA g g Wk W R PLAW LA L y 1 i s = r . o ® @ Fflf o @K 7 3: Hz & @ ‘.‘z £ swePLYE @ [ i < T 5 ®op 5 i, A o g 2o 1 ! SUPPLYSD © K‘% k 3 @ i ) COLD TRAPS B O ! * . ; wre O B i : PRODUCT (Y [ z i L ThF4-LiF, » ’ MAKE- b ; 4 N e <, ol L 6 ® YEOK i r Y .j_nig% — Fig. 3.17. Arrangement of Salt-Processing Equipment. cell (upper left). The other cell houses equipment for the fertile-stream and the "cooler'" fuel-stream operations. The highly_radioactive cell contains only fuel-stream processing equipment: the fluorinator, still, waste receiver, NaF and MgF2 sorbers, and associated vessels. The other cell houses the blanket-processing equipment and fuel- and fertile-stream makeup vessels. 52 A detailed cost estimate for the fuel-recycle plant was made and is reported in Ref. 18. The results for the total capital costs are sum- marized in Table 3.3. The operating and maintenance costs for this plant were also estimated and are shown in Table 3.4. The direct operating Table 3.3. Summary of Cost Estimate for a Typical Fuel-Recycle Processing Plant for a 1000-Mw(e) MSBR Stationa Installed process equipment : | | $ 853,760 ' Structure and improvements | 556,770 Interim waste storage 387,970 Process piping 155,800 Process instrumentation 272,100 Electrical auxiliaries 84,300 Sampling connections 20,000 Utilities (15% of installed process equipment) 128,060 Insulation (6% of installed process equipment) 51,220 Radiation monitoring 100,000 Total direct plant cost $2,609,980 Construction overhead (30% of total direct 782,990 plant cost) - Subtotal construction cost $3,392,970 Engineering and inspection (25% of total con- 848,240 struction cost) Subtotal plant cost | $4,241,210 Contingency (25% of subtotal plant cost) 1,060,300 Total construction cost $5,301,510 Inventory® cost of NaK coolant (at $100/ft3) 40,000 Total capital cost $5,341,510 %Based on throughput of 15 ft3/day of fuel salt and 105 £t3 /day of blanket salt. bInventory of fuel and blanket salts is considered as part of the reactor inventory. o 53 Table 3.4. Summary of Annual Operating and Maintenance Costs of Fuel-Recycle Processing Plant for 1000-Mw(e) MSBR Station® Direct labor | $222, 000 Labor overhead 177,600 Chemicals 14,640 Waste containers 28,270 Utilities 80, 300 Maintenance materials Site , 2,500 Services and utilities 34,880 Process equipment : 160, 040 Total annual charges $721,230 ®Based on throughput of 15 ft3/day of fuel salt and 105 £t3/day of blanket salt. cost includes the cost of immediate supervisory, operating, maintenance, laboratory, health physics, clerical, and janitorial personnel; also in- cluded are costs of chemicals, waste containers, utilities, and mainte- nance materials. These capital and operating costs were used as base points for ob- taining the costs for salt-processing plants having different through- puts. Specifically, the capital and operating costs were estimated separately for each fluid stream as a function of plant throughput, based on the volume of salt processed.l9 The results of these estimates are given in Fig. 3.18, and were used in calculating the nuclear and economic performance of the MSBR fuel cycle. | It may be noted that in Table 3.3 the indirect charges (overhead, ‘ehgineering; and éontingencies) amount to a total of about 100% applied against the direct construction cost of the processing plant. This compares with a similar value of about 41% used in the cost estimate of the MSBR reactor and’tufbine;generator plant (see Sect. 3.11). The high value used here should more than compensate for the higher rates of 54 ORNL-DWG 66-7455 BLANKET STREAM PROCESSING RATE (ft%day) 2 5 10° 2 5 10 AL COST OF CORE PRCCESSING TING COST OF CORE PROCESSING N CORE STREAM PROCESSING COST (:S7f13) BLANKET PROCESS! 'BLANKET STREAM PROCESSING COST (S/t3) COST OF OPERATING COST OF \BLANKET PROCESSING 2 5 10 2 5 100 CORE STREAM PROCESSING RATE (ff:'f’doy) Fig. 3.18. MSBR Fuel-Recycle Costs As a Function of Processing Rates. Fluoride volatility plus vacuum distillation processing for core; fluoride volatility processing for blanket; 0.8 plant factor; 12%/yr capital charges for investor-owned processing plant. equipment replacement in the fuel-processing plant as compared with the power plant as a whole. 3.5.2 Nuclear Design Method Values of the MSBR nuclear design parameters, which were largely fixed by the design criteria in conjunction with nuclgar—economic calcu- lations, are listed in Table 3.1. The criteria helped to establish the design of the salt-circulating loops external to the reactor (the volumes associated With these loops constitute the largest portion of the total volume of salt holdup). Additional parameters which were optimized by the fuel-cycle-performance calculations were the reactor dimensions, the » G o 55 power density, the core composition, including the carbon-to-uranium and thorium-to-uranium ratios, and particularly the fuel-recycle rates through - the processing plant. Table 3.1 also lists the parameter values obtained through nuclear design optimization. The fuel-cycle calculations were performed with OPTIMERC, a combina- tion of an optimization code and a multigroup, diffusion, equilibrium reactor code. Details of the program are summarized in Ref. 20. 1In brief, the program initially calculates the nuclear performance, the equilibrium concentrations of the various nuclides (including the fission products), and the fuel-cycle costs for a given set of conditions; fol- lowing this, performance optimization is done by permitting up to 20 re- actor parameters to be varied, within limits, in order to determine the most desirable values based on the method of steepest ascent. Typical input parameters were the reactor dimensions, blanket thickness, frac- tions of fuel and fertile salts in the core, and fuel- and blanket-stream processing rates. These parameters were varied in a logical fashion, with final values based on designs optimized primarily for minimum fuel cost, with lesser emphasis given to maximizing the annual fuel yield. In addition to fuel-cycle cost per se, OPTIMERC includes several equations for approximating certain capital and operating costs that vary with nuclear design values, such as the size of the reactor vessel and the cost of graphite. These costs were automatically added to the fuel- cycle cost in the optimization routine so that the optimization search would take into account all known economic factors. However, costs other than the fuel-cycle cost are reported undér capital investment (Sect. 3.11). Standard neutron-cross-section libraries were used in obtaining the broad-group cross sections for the MSBR physics calculations (1.2 groups were employed, with one effective thermal group). The cross sections were_evaluatéd and modified where necessary to be consistent with present information (see also Sect. 3.5.4). In obtaining the nuclear constants for nonthermal neutron groups and for a particular region, a transport- type multigroup calculation was performed (B-1 approximation to the Boltzmann equation for a single region); the three specific regions con- sidered were the homogenized core, the blanket, and the reflector regions. 56 The effective thermal-neutron reaction rate was based on transport calcu- lations, which generated the thermal-neutron spectrums in the various reactor regions. In the core, the thermal-spectrum calculation considered the core lattice cell to consist of concentric cylindrical regions; the - resulting neutron reaction rates were used to determine the effective thermal-group cross sections for the various nuclides. The broad-group cross sections were employed in a one-dimensional multigroup diffusion program modified so as to approximate a two-dimen- sional calculation. The concentrations of the various nuclides were based on equilibrium neutron-reaction rates, which were consistent with criticality considerations, the fuel-processing rate, the assumed be- havior of fission products and higher isotopes, and the sale of uranium having an isotopic composition equal to the average in the reactor plant. These reactor-physics calculations were incorporated in the fuel~ cycle-performance optimizations carried out by the OPTIMERC program, in which various parameters were allowed to vary within specified limits. 3.5.3 DNuclear Performance and Fuel-Cycle Cost The nuclear performance of the MSBR is significantly influenced by the physical behavior of the fission products. In particular, the be- havior of 13%Xe and other fission gases is important. A gas-stripping system is provided to remove these gases from the fuel salt. However, part of the xenon could diffuse into the moderator graphite. In the calculations reported here, a *>°Xe poison fraction of 0.005 was assumed. The disposition of the various fission products in the reactor and processing system, based on their estimated physical, chemical, and thermodynamical properties, was assumed to be as shown in Table 3.5. Another factor to consider is the behavior of corrosion products. However, the control of corrosion products in the MSBR does not appear to be a significant problem, so the effect of corrosion products was ne- glected in the nuclear calculations. Not only is the corrosion rate very low, but the fuél-processing methods considered here can remove corrosion products from the molten salts (by reduction with hydrogen followed by filtration). te i s i 57 Table 3.5. Disposition of Fission Products in Reactor and Processing Systems Group Assumed Fission-Product Behavior Fission Products 1 Elements present as gases; assumed to be Kr, Xe removed by gas stripping, with a small fraction absorbed by graphite 2 Elements that plate out on metal sur- Ru, Rh, Pd, Ag, In faces; assumed to be removed in- stantaneously 3 Elements that form volatile fluorides; Se, Br, Nb, Mo, Te, assumed to be removed in the fluoride Te, I volatility process 4 Elements that form stable fluorides less ©Sr, Y, Ba, Ia, Ce, volatile than LiF; assumed to be Pr, Nd, Pm, Sm, separated by vacuum distillation Eu, Gd, Tb 5 Elements that are not separated from the Rb, Cd, Sn, Cs, Zr carrier salt; assumed to be removed only by salt discard The calculation of fuel-cycle cost involves economic factors as well as those given above. The economic ground rules used here are given in Table 3.6. The values of the fissile isotopes were taken from the current AEC price schedule. The capital charges of lZ%/yr for depreciat- ing items and 10% for nondepreciating materials correspond to those for a privately owned plant; the corresponding values used for publicly owned plants were 7 and 5%/yr, respectively. The processing costs are based on the specific fuel-recycle plant design and cost study given above and are included in the fuel-cycle costs. The results, given in Fig. 3.18, were used to estimate the pro- cessing cost as a function of fuel-processing rate. FProcessing losses corresponded to a fissile material loss of 0.1% per pass through fuel- recycle processing. The results of the fuel-cycle calculations for the MSBR design are sumarized in Table 3.7 and the neutron balance is given in Table 3.8. The reactor has the advantage of no neutron losses to structural mate- rials in the core other than the moderator. Except for some unavoidable 58 Table 3.6. Basic Economic Assumptions Used in Nuclear Design Studies Reactor power, Mw(e) | | 1000 Thermal efficiency, % 45 Load factor . 0.80 Cost assumptions Value of 233U and 233Pa, $/g 14 Value of 23°U, $/g 12 Value of thorium, $/kg 12 Value of carrier salt,'$/kg 26 Capital charge, %/yr Plant 12 Nondepreciating capital, including 10 fissile inventory Processing cost, $/ft> salt Fuel (at 10-ft3/da processing rate) 252 Blanket (at 100-ft?/day processing rate) 9.3 Processing-cost scale factor See Fig. 3.18 Table 3.7. MSBR Fuel-Cycle Performance Fuel yield, % per year | 4,86 Breeding ratio 1.0491 Fissile losses in processing, atoms per fis- 0.0057 sile absorption Neutron production per fissile absorption (Tme) 2.221 Specific inventory, kg of fissile material per 0.76%9 megawatt of electricity produced Specific power, Mw(th)/kg of fissile material 2.89 Power density, core average, kw/liter Gross 80 In fuel salt 473 Fraction of fissions in fuel stream 0.987 Fraction of fissions in thermal-neutron group 0.806 Mean 1 of 433y 2.221 Mean 1 of 233U : 1.958 “) tm » 59 Table 3.8. MSBR Neutron Balance Neutrons per Absorption in Fissile Fuel Material Total Absorbed Giving Neutrons Absorbed Fission Produced 232M 0.9710 0.0025 0.0059 233pg 0.0079 233 0.9119 0.8090 2.0233 234y 0.0936 0.0004 0.0010 235 0.0881 0.0708 0.1721 <36y 0.0115 0.0001 0.0001 23TNp 0.0014 238y 0.0009 Carrier salt (except 6Li) 0.0623 0.0185 611 0.0030 Graphite 0.0300 135%e 0.0050 149gm 0.0069 151gpm 0.0018 Other fission products 0.0196 Delayed neutrons lost® 0.0050 Leakage® 0.0012 Total 2.2209 0.8828 2.2209 aDela,yed neutrons emitted outside the core. bLeakage, including neutrons absorbed in the reflector. loss of delayéd'neutrons in the external fuel circuit, there is almost zero neutron leakage from the reactor because of the thick blanket. The neutron losses to fission products are minimized by the availability of rapid and inexpehéive integrated processing. | The fuel-cycle cost for the MSBR is given in Table 3.9. The main items are the fissile inventory and processing costs. The inventory costs are rather rigid for a given reactor design, since they are largely determined by the fuel volume external to the reactor core region. The processing costs are, of course, a function of the processing-cycle times, one of the chief parameters optimized in this study. 60 Teble 3.9. Fuel-Cycle Cost for MSBR* Costs [mill/kwhr(e)] Fuel Fertile Sub- Grand Stream Stream total Total Fissile inventory® 0.1180 0.0324 0.1504 Fertile inventory 0.0459 0.0459 Salt inventory 0.0146 0.0580 0.0726 Total inventory 0.269 Fertile replacement 0.0185 0.0185 Salt replacement 0.0565 0.0217 0.0782 Total replacement 0.097 Processing 0.1223 0.0440 0.1663 Total processing 0.166 Production credit (0.073) Net fuel-cycle cost 0.46 %Based on investor-owned power plant. bIncluding 233Pa, 233U} and #3°U, The fuel costs in Table 3.9 are based on use of private financing. Fuel-cycle and power-production costs based on public financing are also of interest. With public ownership, the fixed annual charge on depre- ciating capital is taken asr7% and on nondepreciating items as 5%. This difference in the financial conditions results in slightly different optimization points for the fuel cycle that affect the volume fractions of fuel, the thorium-to-uranium and carbon-to-uranium ratios, etc. Re- optimizing such parameters has only minor effects on the nuclear per- formance. However, the difference between the 12 and 7% annual fixed charges on the cost of the fuel processing plant and the lower charges on nondepreciating items (5% versus 10%) results in lowering the esti- mated fuel cost from 0.46 mill/kwhr to about 0.29 mill/kwhr. Table 3.10 summarizes the fuel-cycle costs for investor-owned and for publicly owned plants. ®y » 61 Table 3.10. MSBR Fuel-Cycle Costs for Investor-Owned and Publicly Owned Plants Plant factor: 0.8 Cost [mill/kwhr(e)] Investor Public Ownership® OwnershipP Fissile-, fertile-, and carrier- 0.269 0.135 salt inventory Replacement cost of fertile and 0.097 0.097 carrier salts Core- and blanket-processing costs Operation and maintenance 0.075 0.075 Capital costs 0.091 0.053 Bred fuel credit (0.073) (0.073) Net fuel-cycle cost 0.46 0.29 ®Based on 12%/yr capital charges for processing plant and inventory charges of 10%/yr. bBased on 7%Vyr capital charges for processing plant and inventory charges of 5%/yr. 3.5.4 Critique of Nuclear Performance Calculations The performance characteristics given above show that the MSBR has a high specific power [about 1.2 Mw(e)/kg fissile] and a relatively low breeding gain (about 0.05 net fuel bred per unit of fuel burned). Un- certainty in the specific power is due to uncertainties in the fuel in- ventory requirements external to the reactor core (related to the fuel heat exchanger design and flow-distribution systems), as well as to in- accuracies in the critical-mass calculations. _it is estimated that about a 10% uncertainty exists in the fuel volume réquirements.external to the core of the MSBR because of uncertainties in heat transfer, fluid trans- - port, and flow distribution requirements. Relative to critical mass, experience with the MSRE indicates that the calculational methods and applicable neutron cross sections employed are reliable (the calculated 62 MSRE critical mass was within 1% of the experimental value). Also, the methods and cross sections employed are similar to those used by other groups who have had good success in calculating the reactivity of criti- cal assemblies. As a result, the uncertainty in the critical concentra- tion is estimated to be less than 5%. Thus the uncertainty in the specific power appears to be less than 15%. In addition, because of the use of fluid fuel, compensating changes in fissile and fertile material concentrations can be made if the calculated quantities are in error. Finally, since the specific power is high, a small change in its value cannot change the fuel-cycle cost appreciably. Thus uncertainties in speéific power do not appear to significantly affect MSBR performance. With a 10w‘breeding gain, however, uncertainties in nuclear con- stants, fuel-processing losses, and/or physical properties of the fis- sion prodficts can have a significant influence on the fuel doubling time through their influence on the net breeding ratio. A detailed appraisal of the MSBR huclear-performance uncertainties due to the above factors is given in Ref. 21, and the results are summarized below. The importent nuclides in the MSBR are C, Li, Be, F, U, Th, Pa, and fission products. Changes in the neutron-absorption cross-section values of these nuclides can influence the breeding ratio, with some nuclides having more importance than others. The cross-section values are not known in an absolute sense, but they can be inferred from the precision of the various measurements available. On‘this basis, a range of values was assigned to each nuclide that represents a "best judgment" of the values within which the true value will fall. The neutron balance given in Table 3.8 shows the relative absorp- tions in the various materials based on the studies performed. Of the nuclides indicated, only two or three have cross-section uncertainties that could individually affect the breeding ratio by as much as 0.0l. By far the most important nuclide is 233U, and its most important charac- teristic is the value of eta averaged over the reactor neutron spectrum. The 2200-m/sec value and the variation of eta with energy are not known accurately enough to establish eta in MSBR spectrums to much better than about 1% (the 2200-m/sec value used for n?3 was 2.292). The associated uncertainty in breeding ratio is about #0.02 to 0.03, of which the major 8 i » i ¢ £ 63 fraction is due to the uncertainty in the effective thermal value (the uncertainties associated with the 2200-m/sec value and the variation with energy in different energy regions are independent of each other). One of the most abundant materials in the MSBR is fluorine; although its neutron-absorption cross section is low, its high concentration makes it an important material relative to neutron absorptions. For fluorine, the high-energy absorption cross sections are estimated to be uncertain by about #30%. Also, the high-energy neutron reactions in beryllium are uncertain by about *10 to 15%. Uncertainty in the gross cross section for fission products (other than xenon and samarium, whose cross sections are so high that fission yield is the important quantity) is estimated to be about *30% for resonance-energy neutrons, and about *10% for thermal neutrons. Uncertainties in other nuclide cross sections are estimated to be about +10% or less. Based on these uncertainties in cross-section values, the uncertainty in breeding ratio is about #0.02 to 0.03 due to *33U, *0.004 due to *3°U, +0.002 due to protactinium, #0.006 due to fluorine, +0.002 due to 14, +0.002 due to beryllium, and *0.004 for gross fission products. Breaking down these summed uncertainties into their independent uncertainties and taking the square root of the sum of the squares of the independent un- certainties gives a mean uncertainty of about #0.024 in breeding ratio. This result illustrates that the uncertainty in breeding ratio can have a significant effect on the MSBR fuel-yield rate; changing the breeding ratio by *#0.024 would change the fuel-yield rate by about *50%. ‘In addition, the above analysis illustrates the relative importance of the thermal value of 1°? in the MSER. The cross sections actually used in the MSBR studies did not always correspond to values présently considered to be the most probable. For example, the high-energy neutron-absorption cross sections used for fluo- rine are higher than present estimates; also, the graphite absorption cross section (a 2200-m/sec value of 4 millibarns was used) did not allow for burnout of trace impurities. Incorporating such changes would im- prove the breeding ratio by about 0.005. In addition, the assumed be- havior of fission products did not always correspond to present estimates of their behavior in MSBR systems. In particular, it appears most probable 64 that molybdenum, technetium, and other members of group 3 in Table 3.5 will form intermetallic compounds with other fission products rather than remain in solution as fluorides. Under such circumstances the elements will most likely circulate as colloidal~-like metal suspensions (this is indicated by MSRE experience with iron and chromium). In this event the group 3 elements would be removed in fuel-recycle processing, so the effect of the assumed behavior in the MSBR studies was correct. There is a slight possibility that the group 3 fission products will form metal carbides and remain indefinitely in the MSBR core. Such action by a few percent of the group 3 nuclides could lead to a decrease in breeding ratio of about 0.02 or more. As shown in Table 3.5, it was assumed that the group 2 fission prod- ucts would plate out on metal surfaces; at present it appears most likely that these noble metals will remain in colloidal suspension and be removed during fuel-recycle processing. The change in breeding ratio dque to the above change leads to a decrease in breeding ratio of only 0.0CL. It is important that xenon be removed from the MSBR core in order to maintain breeder operation. Experience in the MSRE indicates that the gas removal assumed in Table 3.5 is realistic. Relative to group 5 fission products, it appears that at least cad- mivum and tin will behave like the group 2 fission products and therefore be removed in the fuel-recycle processing. The MSBR calculations assumed that these fission products would be removed through salt discard alone. Changing the behavior of fhis group to that indicated above would increase the breeding ratio by no more than 0.003. Although not discussed previously, it was assumed that 237Np would be removed‘during fuel reprocessing. It appears that this removal can be accomplished by proper operation of the absorber beds. If not removed, the accumulation of 237Np in the fuel stream would decrease the breeding ratio by about 0.01. The fuel-processing losses were assumed to be 0.1% per pass through the fluoride volatility process, and this loss is consistent with experi- mental results. Doubling the losses would decrease the breeding ratio by about 0.006. » " )\ N 65 The nuclear calculations were made with the assumption that all nuclides in the reactor were at their equilibrium concentrations. When starting with 235U as the initial fuel, there will be a period of opera- tion during which nonequilibrium conditions apply. To check the adequacy of the assumption used, the operating time required to approach equilib- rium concentrations with 43°U startup was examined. It was found that 233 and 235U were within 95% of their equilibrium concentrations in less than two years, 234U’was within 95% after eight years, while *2°U was within 80% after ten years. Since 236U buildup is detrimental, startup with 435U fueling will lower the breeding ratio. However, the net effect of 3%y startup is equivalent to increasing the MSBR specific fissile in- ventory by 10 to 15% and considering the equilibrium breeding ratio to apply. This is due to the higher critical mass with 23°U fueling and its decrease with time as the bred fuel is recycled; this keeps the effective fuel "production" rate close to that associated with equilibrium condi- tions, after the first year of MSBR operation. In summary, although there are sufficiently large uncertainties in neutron-cross-section values and in the behavior of fission products to significantly influence nuclear performance, the net nuclear performance presented in Section 3.5.3 appears consistent with present information based on equilibrium fueling of the reactor. Initial fueling with 235U, rather than having equilibrium fueling conditions, will tend to be equiva- lent to a slightly higher specific fissile inventory and a fuel production corresponding to equilibrium conditions. 3.6 0Off-Gas System Xenon and krypton are stripped from the fuel salt in the reactor cir- culating system by sparging with an inert gas, such as helium. Since a xenon-removal cycle time of about 1 min is required to maintain the xenon poisoning at a satisfactorily low level, an in-line sparging system is provided. The sparging gas is introduced at the bottom of the primary heat exchangers to provide some circulation of the salt in event of pump failure. This gas and the fission-product gases are withdrawn in a full- flow gas separator in the pipe between the heat exchanger and the reactor. 66 The flowsheet for the off-gas system is shown in Fig. 3.19. As mentioned above, xenon, krypton, and other fission-product gases are sparged from the fuel-salt circulating system; these gases are removed from the loop in a full-flow centrifugal separator located in the dis- charge of each heat exchanger, with each loop unit discharging about 50 gpm of salt and about 4 scfm of gas.* A jet pump is used to aspirate the fuel-salt-gas stream from the separator; the pump discharges into the salt storage volume above the pump bowl and circulates the helium carrier gas. After passing through the salt storage volume, the carrier gas enters a 1000-ft> decay tank, which is cooled by evaporation of water (similar cooling is used in the MSRE drain tanks®?). The gases then pass through water-cooled charcoal beds, where xenon is retained for 48 hr, and reenter the fuel system at the bottom of the primary heat exchanger. In addition to removing the l35Xe, this system of circulation effectively transfers a large fraction of thé other gaseous fission products to areas where the decay heat can be removed more readily. About 0.1 scfm of the gas stream leaving the charcoal beds (or 0.4 scfm total for the four fuel-salt circulating loops) passes through other charcoal beds and then through a molecular sieve (operated at liquid nitrogen temperature) to remove 99% or more of the 85Ky and other gaseous products. The effluent helium can be recycled into the system or passed through filters, diluted, and discharged into an off-gas stack. The molecular sieves can be regenerated and the radioactive gases driven off can be sent to storage tanks. A helium system also provides cover gas for blanket-salt pump bowls, drain tanks, fuel-handling and -processing systems, etc. The cover gas discharged from these systems passes through charcoal adsorbers and ab- solute filters prior to dilution with air and disposal through the off- gas stack. *¥The full-flow gas separators have been studied only in laboratory- size equipment but are considered to be within the range of present tech- nology. The MSBR loop installation requires 15 small separators arranged in the annulus between the 18- and 24-in. concentric pipes. These small separators would be capable of removing essentially all bubbles larger than 0.0l in. in diameter. " Cw. OUTLET FUEL PUMP & ¢ CW INLET DECAY TAN® 48 hri HOLDUP XENON WATER COOLED CHARCOAL ADSORBERS Fig. 3.19. Off-Gas System Flowsheet. ORNL DWG 66-7135 STORAGE FOR RANE GASES L9 68 3.7 Heat Exchangers The system heat exchangers consist of thé primary heat exchangers used to transfer heat from the fuel and blanket salts to the coolant salt and the boiler-superheaters and reheaters that transfer heat from the coolant salt to the supercritical fluid in the.steamrpower system. Also included, although more closely associated with the steam system than the salt systems, are the reheat steam preheaters. 3.7.1 Fuel-Salt Heat Exchangers Four shell-and-tube two-pass vertical heat exchangers transfer heat from the fuel salt in the tubes to the coolant salt circulated through the shell. The conceptual design is shown in Fig. 3.20, and the perti- nent data are listed in Table 3.11. Each exchanger has a capacity of about 528 Mw(th) and is about 5.5 ft in diameter and 18.5 ft high, including the bowl of the circulating pump, which is an integral part of the heat exchanger shell. Shell, tube,‘ and tube sheets are fabricated of Hastelloy N. The reactor fuel salt enters the heat exchanger from the 18-in.- diam inner passage of the concentric pipes connecting the reactor and exchanger. In the heat exchanger, the fuel flows downward through the annular, outer rows of tubes at a velocity of 11.3 fps. In each unit there are 4167 of these 0.375-in.-0D tubes on a 0.75-in. pitch. Upon reaching the bottom head the salt reverses direction and moves upward at about 13 fps through a center bank of 0.375-in.-diam tubes. There are 3624 of these tubes on a 0.625-in. pitch. Thus each fuel-salt primary heat exchanger has 7791 tubes and about 9665 ft? of effective surface area. The coolant salt enters the heat exchanger at the top and flows dowmward, countercurrent to the flow of fuel salt. It initially flows through the center section of the exchanger, and on reaching the bottom of the shell it turns upward to flow through the tubes in the annular section and leave the exchanger through an annular collecting ring at the top. i 3 69 | - ORNL DWG 65-12379 i _ g / FUEL SALT DUMP TANK ; . FUEL SALT PUMP | T . : : s > FUEL SALT FROM NY/ % , /// AN REACTOR — ALl (¥ 1%5‘,‘ COOLANT SALT L. : FROM STEAM - HEAT EXCHANGER | - FUEL SALT TO ——/‘ REACTOR COOLANT SALT ) TO BLANKET [ - HEAT EXCHANGER DOUGHNUT murruaZ T T DISK BAFFLE | —————————— LONGITUDINAL BAFFLE LONGITUDINAL R 7—7%55 BAFFLE i - :::?‘“TlE RODS & SPACERS CORE TUBE | ANNULAR SHEET —\ : 1 /"TUBE SHEET EXPANSION BELLOWS SHROUD COOLANT SALT DRAIN FUEL DRAIN |- Fig. 3.20. Fuel-Salt Heat Exchanger. | ; ] i { » | 70 Table 3.11. Fuel-Salt Heat Exchanger Design Data Type Shell-and~tube two-pass vertical exchanger with . disk and doughnut baffles Number required 4 Rate of heat transfer, each, Mwr 528 Btu/hr 1.80 x 10° Shell-side conditions Cold fluid Coolant salt Entrance temperature, °F 850 Exit temperature, °F 1111 Entrance pressure, psi 80 Exit pressure, psi 29 Pressure drop across exchanger, psi 51 Mass flow rate, 1b/hr 1.68 x 107 Tube~side conditions Hot fluid Fuel salt Entrance temperature, °F 1300 Exit temperature, °F 1000 Entrance pressure, psi 9% Exit pressure, psi 10 (pump suction) Pressure drop across exchanger, psi 86 Mass flow rate, 1b/hr 1.08 x 107 Mass velocity, l'b/hr-ft2 Center section 5.95 x 10° Annular section 5.18 x 10° Velocity, fps Center section 13.0 Annular section 11.3 Tube material Hastelloy N Tube 0D, in. 0.375 Tube thickness, in. 0.035 Tube length, tube sheet to tube sheet, ft Center section 13.7 Annular section 11.7 Shell material Shell thickness, in. Hastelloy N 0.5 o i 71 Table 3.11 (continued) Shell ID, in. Center section Annular section Tube sheet material Tube sheet thickness, in. Top annular section Bottom annular section Top and bottom center section Number of tubes Center section Annular section Pitch of tubes, in. Center section Annular section Total heat transfer area per exchanger, ft2 Center section Annular section Total Basis for area calculation Type of baffle Number of baffles 40.2 66.5 Hastelloy N =W O~ 0 oo 3624 4167 0.625 0.750 4875 4790 9665 Tube outside diameter Disk and doughnut Center section 5 Annular section 2 Baffle spacing, in. Center section 27.4 Annular section 21 Disk 0D, in. Center section 30.6 ~ Annular section 55.8 Doughhut 1D, * Center section - 25.0 Annular section _ 51.0 Overall heat transfer coefficient, U, - 1110 Btu/hr-ft2 72 Table 3.11 (continued) Maximum stress intensity,® psi Tube _ Calculated Py = 413; (Pp + Q) = 12,000 Allowable Pnp = Sy = 4600; (Py + Q) = 38y = 13,800 Shell ' Calculated Py = 6160; (Pp + Q) = 21,600 Allowsble Py = = 12,000; (Pp + Q) = 35Sy = 36,000 Maximum tube sheet stress, psi Calculated \ | 10,750 Allowable 10,750 ®The symbols are those of Section 3 of the ASME Boiler and Pressure Vessel Code, with Py = primary membrane stress intensity, Q = secondary stress intensity, Sy = allowable stress intensity. The general configuration and arrangement of the exchanger'were largely dictated by the design requirement that the fuel-salt circulating system have a minimum fuel inventory consistent with practical design considerations. Associated factors were permissible stress values and the ability to remove afterheat and drain the core. The heat exchanger calculations were concerned primarily with determining the lengths and number of tubes, the tube pitch, the number of baffles, the baffle spac- ing, etc., which would best suit the specified conditions. A computer program was developed for this optimization work. The program and the details of the calculations for all the MSBR heat exchangers are reported elsewhere.?3 | To distribute coolant-salt flow on the shell side of the exchanger, disk and doughnut baffles are used in the center section. In the annular region there are two baffles, one extending inward from the exterior shell and one extending outward from the barrier that surrounds the core sec- tion. These baffles improve the shell-side heat transfer coefficient; h 73 however, no baffles are used at the top of the annular section, because the hottest fuel fluid enters here and an improved heat transfer coeffi- cient would result in an excessive temperature drop across the tube wall. Also, a baffle is located near each tube sheet to partially insulate it and thereby reduce the temperature drop across the sheet. Fuel or cool- ant salt can be drained from the bottom of the primary exchangers through the concentric drain lines indicated din Fig. 3.20. The stresses that tend to be deveiqped in the heat exchanger due to the temperature differences between the shell and the upflow and downflow tubes are relieved in the design concept by a bellows expansion joint at the lower tube sheet. The stresses in the present design are given in Table 3.11.% 3.7.2 Blanket-Salt Heat Exchangers The four shell-and-tube vertical heat exchangers used to transfer heat from the blanket salt to the coolant salt are very similar to the fuel-salt exchangers, but they only have a capacity of 27.8 Mw(th) each. They are illustrated in Fig. 3.21. Pertinent design data are given in Table 3.12. The coolant salt passes through the fuel-salt heat exchangers and then through the blanket exchangers, in series, entering the latter at about 1111°F and leaving at 1125°F. Since the flow rate is relatively high and the temperature change is small, the exchangers are designed for a single shell-side pass of the coolant salt. The blanket salt in the 0.375-in.-0D tubes makes two passes, however, moving downward at about 10.5 fps in the outer annular section and upward through the inner bank to the pump suction. - *Other exchanger designs were also studied that utilized bent tubes rather than the bellows to absorb the differential expansion; these ex- changers had the pump discharging fuel from the reactor into the heat exchanger so that the point of highest pressure in the system was the exchanger rather than the reactor. The results of these studies are presented in Section 4.4. 1In general, it is believed that the present exchanger design can be improved to minimize engineering development problems but that the estimated capital costs of heat exchangers based on the present design are representative of developed heat exchanger costs. 74 ORNL DWG 65-123 80 _ BLANKET SALY FROM REACTOR BLANKET SALT TO REACTOR DISK BAFFLE ODOUGHNUT BAFFLE il I TO BLANKET DRAIN TANKS Fig. 3.21. Blanket-Salt Heat Exchanger. BLANKET SALT PUMP COOLANT SALT TO STEAM _HEAT EXCHANGER COOLANT SALT FROM PRIMARY EXCHANGER < » 75 Table 3.12. Blanket-Salt Heat Exchanger Data Number required ~Rate of heat transfer per unit, Mw Btu/hr Shell-side conditions | Cold fluid Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Exit pressure, psi Pressure drop across exchanger, psi Mass flow rate, lb/hr Tube-side conditions Hot fluid Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Exit pressure, psi Pressure drop across exchanger, psi Mass flow rate, lb/hr Mass velocity, lb/hr-ft? Velocity, fps Tube material Tube 0D, in. Tube thickness, in. Tube length, tube sheet to tube sheet, £t Shell material Shell thickness, in. Shell ID, in. | Tube sheet material Tube sheet thickness, in. Number of tubes Pitch of tubes, in. Total heat transfer area, ft° Basis for area claculation Shell-and-tube one-shell- pass two-tube-pass exchan- ger, with disk and doughnut baffles 4 27.8 9.47 X 107 Coolant salt 1111 1125 27 9 18 1.68 X 107 Blanket salt 1250 1150 100 10 90 4.3 X 108 10.5 X 10° 10.5 Hastelloy N 0.375 0.035 8.25 Hastelloy N 0.25 36.5 Hastelloy N 1.0 1641 (~820 per pass) 0.81 1330 Outside diameter 76 Table 3.12 (continued) Type of baffle : ’ Disk and doughnut Number of baffles 3 Baffle spacing, in. . 24.8 Disk 0D, in. 26.5 Doughnut ID, in. 23 Overall heat transfer coefficient, U, 1020 Btu/hr«ft2 Maximum stress intensity,® psi Calculated Py = 410; (P, + Q) = 7840 Allowable Pp = Syp = 6500; (Py + Q) = - 35Sy = 19,500 Shell Calculated ~ Pn = 1660; (Pp + Q) = 11,140 Allowable Pp = Sm = 12,000; (Pp + Q) = 38m = 36,000 Maximum tube sheet stress, psi Calculated 2220 Allowable 5900 at 1200°F ®The symbols are those of Section 3 of the ASME Boiler and Pressure Vessel Code, with Pn = primary membrane stress intensity, Q@ = secondary stress intensity, Sm = allowable stress intensity. Straight tubes with two tube sheets are used rather than U-tubes in order to permit drainage of the blanket salt. Disk and doughnut baffles are used to improve the shell-side heat transfer coefficient and to pro- vide the necessary tube support. Baffles on the shell side of the tube sheets reduce the temperature difference across the sheets to keep thermal stresses within tolerable limits. Calculations show that the relatively low pressures and small temperature differences produce stresses that are well within the allowable range.?23 im # 77 3.7.3 Boiler-Superheaters Sixteen vertical U-tube U-shell heat exchangers-are used to transfer the heat from the 1125°F coolant salt to the 700°F feedwater to generate steam at 1000°F and about 3600 psia. Four of these exchangers are in each of the coolant-salt circulating circuits and are supplied by a variable-speed coolant-salt pump (adjustment of the pump speed permits control of the outlet steam temperature). Each exchanger has a capacity of about 121 Mw(th) and has a U-shaped cylindrical shell about 18 in. in diameter; each vertical leg stands about 34 ft high, including the spheri- cal head. The tubes and shell are fabricated of Hastelloy N. The unit is shown in Fig. 3.22. Pertinent design data are given in Table 3.13. Because of marked changes in the physical properties of water as the temperature increases above the critical point (at supercritical pres- sures), heat transfer calculations for this particular exchanger were made on the basis of a detailed spatial analysis with a computer program.23 The calculations established the optimum number of tubes, tube length, nunber of baffles, and baffle spacing, in terms of specified design cri- teria. The results indicated that the optimum design was an exchanger with a long, slim shell and relatively wide baffle spacing. The spacing was greatest in the central portion of the exchanger where the temperature difference between the bulk fluids is high. The 3600-psi fluid pressure on the inside of the tubes dictates that the heads and tube sheets be carefully designed. The relatively small diameter of 18 in. selected for the shell and the spherical heads on the ends of the exchangers allows the stresses to be kept within permissible limits. A baffle on the shell side of each tube sheet provides a stag- nant salt layer that helps-to reduce gtresses in the sheet due to tempera- ture gradients. L e | The coolant salt can be cofipletely drained from the shell. The water can be partially removed from the tubes by gas pressurization, or by flushing, but completeldrainability was not considered a mandatory design requirement. 78 | ORNL DWG 65-12383 SUPER CRITICAL SUPER CRITICAL FLUD OUTLET _ FLUI T —qFLUID WLE COOLANT SALT _ COOLANT SALT INLET OUTLET | TIE ROD & SPACER AFFLES Fig. 3.22. Boiler-Superheater. i M o b e el 88 b A1 L 50w L AL e koL s 79 Table 3.13. Boiler-Superheater Design Data Type Number required Rate of heat transfer, Mw Btu/hr Shell-side conditions Hot fluid Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Exit pressure, psi Pressure drop across exchanger, psi Mass flow rate, lb/hr Tube~side conditions Cold fluid Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Exit pressure, psi Pressure drop across exchanger, psi Mass flow rate, lb/hr Mass velocity, 1b/hr.ft? Tube material Tube OD, in. Tube thickness, in. Tube length, tube sheet to tube sheet, ft Shell material | Shell thickness, in. Shell ID, in. Tube sheet material Tube sheet thickness,.in. Number of-tubes Pitchrof tubes, in. Total heat transfer area, £t2 Basis foraarea calculation Type of baffle Number of baffles ' U-tube U-shell exchanger with crossflow baffles 16 121 4,13 X 108 Coolant salt 1125 850 150 92 58 3.66 X 106 Supercritical fluid 700 1000 3770 3600 166 6.33 X 10° 2.78 X 106 Hastelloy N 0.50 0.077 63.8 Hastelloy N 0.375 18.2 4Hastelloy N 475 349 0.875 2915 Outside surface Crossflow 9 80 Table 3.13 (continued) Baffle spacing - Variable Overall heat transfer coefficient, U, 1030 Btu/hr - £t2 | Maximum stress intensity,® psi Tube Calculated P, = 6750; (P, + Q) = 40,700 ’ ’ 48,000 Shell : Calculated Py = 3780; (P, + Q) = 8540 Allowable : Pp =-10,500; ?Bm + Q)allow = 31,500 Maximum tube sheet stress, psi Calculated . <16,600 Allowable - - 16,600 ®The symbols are those of Section 3 of the ASME Boiler ahd Pressure Vessel Code, with Py = primary membrane stress intensity, Q = secondary stress intensity, Sm = allowable stress intensity. 3.7.4 Steam Reheaters Eight shell-and-tube heat exchangers transfer heat from the coolant salt to the high-pressure-turbine exhaust steam (~57O psia) and raise its temperature to 1000°F. The steam enters the exchanger at about 650°F, having been heated from the 552°F exhaust temperature in a preheater described below. There are two reheaters to each coolant-salt circu- lating loop, each pair being supplied by a variable-speed coolant-salt pump in an arrangement that permits control of the outlet steam tempera- ture. The general arrangement of the reheaters is shown in Fig. 3.23, and design data are givén in Table 3.14. | Each of the eight units has a capacity of about 36 Mw(th); the cool- ant salt enters a unit at 1125°F and leaves at 850°F. Straight vertical shells about 28 in. in diameter and 24 ft long are used.¥ Both shell and *The straight shell occupies less cell volume than a U-tube U-shell design and requires slightly less coolant-salt inventory. ra 81 ORNL DWG 66-7118 STEAM OUTLET TUBE SHEET TIE ROD & SPACER TUBULAR SHELL gl!&g' TUBES DISC BAFFLE DOUGHNUT BAFFLE SALT OUTLET INSULATION BAFF ORAIN LINE STEAM INLET Fig. 3.23. ©Steam Reheater. 82 Table 3.14. Steam Reheater Design Data Type Number required Rate of heat transfer per unit, Mw- : Btu/hr Shell-side conditions Hot fluid A Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Exit pressure, psi Pressure drop across exchanger, psi Mass flow rate, 1lb/hr 1 Mass velocity, lb/hr- P12 Tube-side conditions Cold fluid Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Exit pressure, psi Pressure drop across exchanger, psi Mass flow rate, lb/hr Mass velocity, 1b/hr- 12 Velocity, fps Tube material Tube OD, in. Tube thickness, in. Tube length, tube sheet to tube sheet, ft Shell material Shell thickness, in. Shell ID, in. Tube sheet material Tube sheet thickness, in. Number of tubes Pitch of tubes, in. Total heat transfer area, ft? Basis for area calculation Straight tube and shell ex- changer with disk and dough- nut baffles 8 36.2 1.24 X 108 - Coolant salt 1125 850 106 90 16 1.1 X 106 1.44 X 10° Steam 650 1000 570 557 13 6.3 X 10° 4.0 X 10° 147 Hastelloy N 0.75 0.035 22.9 Hastelloy N 0.5 28 Hastelloy N 4.75 628 1.0 2830 Outside of tubes 83 Table 3.14 (continued) Type of baffle Disk and doughnut Number of baffles 14 and 15 Baffle spacing, in. 8.75 Disk OD, in. 23.2 Doughnut ID, in. 16.0 Overall heat transfer coefficient, U, 275 Btu/hr -ft? Maximum stress intensity,® psi Tube Calculated P, = 5240; (P, + Q) = 15,100 Allowable Pp = 14,500; (Py + Q) = | 43,500 Shell Calculated Py = 4350; (Pp + Q) = 14,800 Allowable P, = 10,600; (P, + Q) = 31,800 Maximum tube sheet stress, psi Calculated 9,600 Allowable : 9,600 e symbols are those of Section 3 of the ASME.Boiler and Pressure Vessel Code, with Pp = primary membrane stress intensity, Q = secondary stress intensity, Sm = allowable stress intensity. tubes are fabricated of Hastelloy N. Disk- and:dbughnut—type baffles support the tubes at closé intervals to prevent.excessive vibration. Baffles on the shell sides of the tube sheets provide a stagnant layer of coolant salt to reduce thermal stresses in the sheet. A special drain pipe at the bottom provides for drainage of the coolant salt. Analyses of the stresses'ihdicated that the values were within per- missible limits. 3.7.5 Reheat~Steam Preheatérs Throttle steam at 3500 psia and 1000°F is used to heat the high- pressure turbine exhaust from about 552 to 650°F before it enters the 84 reheaters. (Heat transfer studies indicate that no freezing of the cool- ant salt takes place in the reheaters if steam enters at 650 rather than 700°F, due to the low value of the steam-side heat transfer coefficient.) Use of this preheater permitted the adoption of the TVA Bull Run Sfieam Plant operating conditions without significant changes affecting costs or performance. This factor was given priority over designing for maxi- mum thermodynamic efficiency and minimum cost, * since the difference would be swall and have little effect on the findings of this study. The design concept for the reheat-steam preheaters is shown in Fig. 3.24, and the design data are listed in Table 3.15. ZEight preheaters *A thermodynamically more efficient arrangement would be to exhaust the steam from the high-pressure turbine at 650°F rather than 552°F, i which would also have the advantage of eliminating the preheating equip- ! ment (estimated cost, $275,000). . ORNL DWG 65-12382 SUPER CRITICAL SUPER CRITICAL FLUID INLET FLUID OUTLET ~/: TUBE SHEET BAFFLE ” STEAM QUTLET | BY+PASS RING STEAM INLET = ’ J———TIE ROD & BPACER ? TUBES /nv-mss RING Fig. 3.24. Reheat-Steam Preheater. e 85 Table 3.15. Reheat~Steam Preheater Design Data Type Number required Rate of heat transfer, Mw Btu/hr Shell~side conditions Cold fluid Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Exit pressure, psi Pressure drop across exchanger, psi Mass flow rate, lb/hr Mass velocity, 1b/hr.ft? Tube-side conditions Hot fluid Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Exit pressure, psi Pressure drop across exchanger, psi Mass flow rate, 1b/hr Mass velocity, 1b/hr-ft? Velocity, fps Tube material Tube 0D, in. Tube thickness, in. Tube length, tube sheet to tube sheet, ft Shell material Shell thickness, in. Shell ID, in. Tube sheet material Tube sheet thickness, in. Number of tubes Pitch of tubes, in. Total heat transfer area, ft° Basis for area calculation One-tube-pass one-shell-pass U-tube U-shell exchanger with no baffles 8 12.3 4,21 X 107 Steam 552 650 595.4 520.0 Supercritical water 1000 869 3600 3544 56 3.68 X 107 1.87 X 108 93.5 Croloy 0.375 0.065 13.2 Croloy 7/16 20.2 Croloy 6.5 603 0.75 781 Tube outside diameter 86 Table 3.15 (continued) Overall heat transfer coefficient, U, 162 Btu/hr-f£t? Maximum stress intensity,2 psi Tube . Calculated Pp = 10,500; (Pp + Q) = 15,900 Allowable Pp = Sp = 10,500 at 940°F; TPm + Q) = 38y = 31,500 Shell Calculated P, = 14,400; (P, + Q) = 33,100 Allowable P, = Sp = 15,000 at 650°F; ey + Q) = 38, = 45,000 Maximum tube sheet stress, psi Calculated 7800 Allowable 7800 at 1000°F SThe symbols are those of Section 3 of the ASME Boiler and Pressure Vessel Code, with Py = primary membrane stress intensity, Q = secondary stress intensity, Sm = allowable stress intensity. are used. This number was determined almost entirely by the selection of reasonable dimensions for the units. The preheaters are part of the steam power system and no biological shielding is required. They are located in the portion of the plant assigned to the feedwater heaters. Fach preheater has a capacity of about 12.3 Mw(th). Each vertical leg of the U-shell is about 21 in. in diameter and the overall height is about 15 ft, including the spherical heads. The tubes, tube sheets, and heads contain the 3500-psia throttle steam and are designed for this_high pressure and temperature. Selection of the U-shell rather than a divided cylindrical shell permits use of small head diameters and reduces the required tube-sheet and head thicknesses. Stress analyses indicate that the stresses are within the allowable limits. Both tubes and shell are fabricated of Croloy. The flow in the preheaters is countercurrent and no baffles are needed in the shell. The U-tube construction accommodates the thermal expansion that occurs. The relatively high steam film resistance to heat — o 87 transfer on the shell side reduces the temperature gradient across the tube wall to permissible levels. 3.8 BSalt-Circulating Pumps Fach of the four separate salt-circulating circuits contains a fuel- salt circulating pump, a blanket-salt pump, a coolant-salt pump for the boiler superheaters, and a coolant-salt pump for the reheaters. The de- sign data for these pumps are listed in Table 3.16. The pump designs utilize the technology developed over the past 15 years, with present pump capacities being extrapolated a factor of 10 for MSBR use. Table 3.16. Salt Pump Dimensions and Performance Requirements Reheat Superheat Si:iim g;z:g;# Coolant Coolant System System Number of pumps 4 4 4 4 Temperature, 1,000 1,150 1,125 1,125 Flow, gpm (each) 11,000 2,200 2,200 14, 000 Head, ft 140 100 110 150 Speed, rpm 1,170 1,750 1,750 1,170 Impeller diameter, in. 24 13 13 24 Pump tank diameter, in. 36 60 Suction diameter, in. 18 8 8 18 Discharge diameter, in. 6 14 Nominal motor power, hp 1,250 500 200 1,250 Motor length, in. 92 72 37 92 Motor diameter, in. 64 40 29 64 A1l pumps are of centrifugal type, with a vertical shaft supported at its lower end in a hydrodynamic journal bearing lubricated by molten salt. The fuel- and blanket-salt pump bowls have diffuser vanes and are an integral part of the pfimary heat exchanger vessels. The equipment arrangements are illustrated in Figs. 3.20 and 3.21. Figure 3.25 shows the fuel- and blanket-salt pumps and Fig. 3.26 shows details of the ‘coolant-salt pump. 88 ORNL-DWG 66-69T71 ELECTRIC MOTOR N DIMENSIONS ARE LISTED N TABLE 3.15 $— { %g” 7 IR v GAS puaee-—-,..f ~ / | /) 1 TEN SALT . JOURNAL BEARING |/ | / ] DIFFUSER TR= = %EE'E{;&L&GQ&E%ORT 2 A /1 /1 / ; / Y — 9 SUCTION g Fig. 3.25. Tuel- and Blanket-Salt Pump. A continuous purge of inert gas flows through each pump during opera- tion. Thus purge gas enters the labyrinth annulus near the upper end of the pump shaft. The labyrinth seals the motor cavity from the gas space in the pump tank. The purge gas flow splits into two paths; one portion flows upward in the annulus to keep lubricating vapors from entering the pump tank, and the other flows downward to prevent the migration of radio- active gases into the mofior cavity. The pumps éonstitute a part‘of the primary containment of the reactor fluid. As such, they would be constructed in accordance with the applicable i 89 ORNL-DWG 66-6972 ELECTRIC MOTCR PURGE PUMP TANK SALT GAS QUT JOURNAL BEARING SELF ALIGNING GAS VOLUME BEARING SUPPORT DISCHARGE DIMENSIONS ARE LISTED IN TABLE 3.15 SUCTION Fig. 3.26. Coolant-Salt Pump. portions of the ASME codes, with proper allowances made for the thermal strain fatigue that would accompany reactor startups, power cycles, and radiation heating. The long shafts on the fuel- and blanket-salt pumps permit the drive motors to be shielded from the reactor radiation and temperature. The electric drive motors are located outside the biological shielding, with the hermetiélcans around these motors serving as part of the reactor containment vessel. A squirrel-cage induction motor is used, with the ball bearings lubricated with radiation-resistant grease capable of withstanding 3 X 10° rad. The electrical insulation also uses special 90 materials having a radiation tolerance up to 10° rad. Motor heat is re- moved by circulating a coolant through coils inside the hermetically sealed motor wvessel. 3.9 Steam-Power System The thermal power of the MSBR is 2225 Mw(th), which provides a full- load net electrical output of 1000 Mw(e) plus about 35 Mw(e) of power for auxiliary equipment. Throttle steam conditions are 3500 psia and 1000°F/1000°F; these conditions are representative of modern steam-power - plant practice and correspond to those employed in the recently completed TVA Bull Run Steam-Electric Plant. The steam-power system flowsheet is shown in Fig. 3.27, and the de- sign and performance data are summarized in Table 3.17. Energy balances were made in determining the thermodynamic performance of the system based on a 700°F inlet feedwater tem;pera.ture.24 The flow rates and steam prop- erties at various points in the system are shown on Fig. 3.27. The net efficiency of the plant is about 45%. The MSBR system has a conventional 1035-Mw(e) gross output cross- compounded 3600/1800—rpm four-flow turbine-generator unit with 3500-psia 1000°F steam to the throttle and reheat to 1000°F after the high-pressure turbine exhaust. The exhaust pressure at rated conditions is 1l.5-in. Hg abs. Eight stages of feedwater heating are used, with extraction steam taken from the high- and low-pressure turbines and also from three points on the turbines used to drive the boiler feedwater pumps. The feedwater leaves heater 4 at about 357°F and 200 psia; it is raised to 3800 psia and 366°F by two turbine-driven centrifugal pumps. The pumps have six stages, run at 5000 rpm, and deliver 8100 gpm against a head of 9380 ft. The drive turbines, which have eight stages, are sup- plied with throttle steam at 1069 psia and 700°F, and they exhaust at about 77 psia. There is one drive turbine per pump, and no standby pump- ing capacity is provided. The MSBR steam-power system differs from the TVA Bull Run plant in having higher feedwater and reheat-steam temperatures. The temperature of - O the feedwater entering the boiler-superheaters was governed by the estimated ik - " " ORNL DWG 66-7125 91 o . . e et s S NT 3.2 e 1054.9 Wwe 29.4 N 2EES Mt sTe% TAN Dk 1000 hwe i QROSS EFF OF CYGLE NET EFF. OF CYCLE REACTOR HEAT WRUT NET HEAT RATE OF PREDNUNL-BO0ETER FUMPS OUTAUT 8 PUNPE SUMMARY OF PERFORMANCE AT RATED LOAD NET OUTRUT BLEC. POWER FOR AUX. Lo et ——— du e m e, — - —————— - ————— e "nll.l_ . ;o { ; | ) 1 “ | J-E.lb_lllti. -.ll-l . oo — e — 1 ! — r A | L | _ u_fluurlll.“ J.r_q._lmV“l_.+p _ w21 F R P ) ! X T T L P i ._m _m v %l _w R oI 1 5 —M —w HII_ llllllll ID- | T 1T _ b L —sem P — e — = - e (I “ b - ———————n | . | mmw | | w [ Ly | & l 2 g ww £ 1 - SB00P- 368,17 MIBH Steam System Flowsheet for 700°F Feedwater. Fig. 3.27. 92 Table 3.17. MSBR Steam-Power System Design and Performance Data with 700°F Feedwater General performance _ Reactor heat input, Mw 2225 Net electrical output, Mw 1000 Gross electrical generation, Mw 1034.9 Station auxiliary load, Mw 25.7 Boiler-feedwater pressure-booster pump load, Mw 9.2 Boiler-feedwater pump steam-turbine power output, — 29.3 Mw (mechanical) Flow to turbine throttle, lb/hr 7.15 X 106 Flow from superheater. 1b/hr 10.1 X 10° Gross efficiency, % (1034.9 + 29.3)/2225 47.8 Gross heat rate, Btu/kwhr 7136 Net efficiency, % 4,9 Net heat rate, Btu/kwhr - 7601 Boiler-superheaters Number of units 16 Total duty, Mw(th) 1932 Total steam capacity, 1b/hr 10.1 X 106 Temperature of inlet feedwater, °F 700 Enthalpy of inlet feedwater, Btu/lb 769 Pressure of inlet feedwater, psia 3770 Temperature of outlet steam, °F 1000 Pressure of outlet steam, psia ~3600 Enthalpy of outlet steam, Btu/lb 1424 Temperature of inlet coolant salt, °F 1125 Temperature of outlet coolant salt, °F 850 Average specific heat of coolant salt Btu/1b*°F 0.41 Total coolant-salt flow, lb/hr 58.5 X 106 cfs 130 gpm 58,300 Coolant-salt pressure drop, inlet to outlet, psi ~60 Steam reheaters Number of units | 8 Total duty, Mw(th) 294 Total steam capacity, 1b/hr 5.13 X 108 Temperature of inlet steam, °F 650 Pressure of inlet steam, psia ~570 Enthalpy of inlet steam, Btu/lb 1324 Temperature of outlet steam, °F 1000 Pressure of outlet steam, psia 557 Enthalpy of outlet steam, Btu/lb 1518 Temperature of inlet coolant salt, °F | 1125 Temperature of outlet coolant salt, °F 850 Average specific heat of coolant salt, Btu/lb °F 0.41 h 93 Table 3.17 (continued) Steam reheaters (continued) Total coolant salt flow, lb/hr cfs gpm Coolant-salt pressure drop, inlet to outlet, psi Reheat-steam preheaters Number of units Total duty, Mw(th) Total heated steam capacity, lb/hr Temperature of heated steam, °F Inlet Outlet Pressure of heated steam, psia Inlet Outlet Enthalpy of heated steam, Btu/lb Inlet Outlet Total heating steam, lb/hr Temperature of heating steam, °F Inlet Outlet Pressure of heating steam, psia Inlet Outlet Boiler-feedwater pumps Number of units Centrifugal pump Number of stages ' Feedwater flow rate, total, 1b/hr Required capacity, gpm Head, approximate, ft Speed, rpm Water inlet temperature, °F Water inlet enthalpy, Btu/1b Water inlet specific volume, ft3/1b Steam-turbine drive Power required at rated flow, Mw (each) Power, nominal hp (each) | Throttle steam conditions, psia/°F Throttle flow, 1b/hr (each) Exhaust pressure, approximate, psia Number of stages Number of extraction points 8.88 x 10° 19.7 8860 ~117 8 100 5,13 X 10° 552 650 595 590 1257 1324 2.92 X 108 1000 869 3600 3544 2 6 7.15 X 10° 8,060 9,380 5,000 358 330 ~0,0181 14.7 ~ 20,000 1070/700 414,000 77 8 3 94 Table 3.17 (continued) Boiler-feedwater pressure-booster pumps Number of units . 2 Centrifugal pump Feedwater flow rate, total, 1lb/hr 10.1 X 10° Required capacity, gpm (each) 9,500 Head, approximate, ft 1413 Water inlet temperature, °F 695°F Water inlet pressure, psia ~3,500 Water inlet specific volume, ft2/1b ~0.0302 Water outlet temperature, °F ~700 Electric-motor drive , Power required at rated flow, Mw(e) (each) 4.6 Power, nominal hp (each) 6,150 liquidus temperature of the coolant salt; it was decided that the coolant salt should not be permitted to freeze, so the feedwater must enter the boilers at 700°F or higher. In the reheaters, however, the heat-transfer resistance of the steam film is high, so the steam can enter at 650°F. The feedwater leaves the conventional eight stages of regenerative feedwater heating at about 551°F, the same as in the TVA Bull Run steam- power cycle. The steam leaves the high-pressure turbine at about 552°F and is heated to 650°F in a shell-and-tube type exchanger (described in Sect. 3.7.5), with supercritical fluid at 3515 psia and 1000°F. The high-pressure heating steam leaves the heat exchanger near 866°F and 3500 psia and is directly mixed in a "mixing tee" with the 550°F feedwater to raise its temperature to about 695°F. The mixture is then boosted to boiler-superheater pressure by motor-driven pumps, and the pumping effort raises the pumped water temperature to the requisite 700°F. The density of the supercritical fluid pumped by the booster pumps is about 34 1b/ft3, and very little compressive work [~9.2 Mw(e)] is involved in raising the fluid pressure. The pumps employed are similar tbrthose used for forced- convection flow in supercritical-pressure steam generators. Each of the two booster pumps has a rating of about 20,000 gpm and 6200 hp. 5 95 3.10 Other Design Considerations 3.10.1 Piping and Pipe Stresses The stresses in the salt and steam piping were studied briefly to determine whether the reactor and turbine plant layouts contained grossly impractical arrangements. The calculations were made with the MEC-21/7094 25 In these estimates, it was assumed that the centers of the re- code. actor and the turbines were fixed and the rest of the system was allowed to move in accordance with the thermal-expansion forces. Stresses were examined at about 150 points with particular emphasis on the locations of suspected high stresses. The sizes of the main piping in the steam-power system are shown in Table 3.18. The assumed vélocities, materials, and other conditions are also given. The maximum calculated stress in piping fabricated of Hastel- loy N was found to be about 10,000 psi, which is about a factor of 3 less than that allowable based on ASME Code requirements. The Croloy steam piping has a maximum calculated stress of 2200 psi, which is well within the allowable value. One case was calculated in which the coolant-salt pumps were re- strained in the direction transverse to the main coolant-salt pipe rum. The maximum stress in the Hastelloy N piping in this case was about 22,000 psi; the maximum stress in the Croloy steam piping was essentially the same as before. BSince the wvertical deflections at the pump location are apparently small, it appears that use of vertical and transverse re- straints will not cause thermal-expansion effects to overstress the piping. 3.10.2 Maintenance The MSBR equipment was designed and arranged so that inspection, maintenance, and replacement of all major equipment would be practical. Most of the maintenance would be done by use df remotely operated tools through openings in roof plugs. The feasibility of such methods has been demonstrated in the MSRE, and information is available relative to the special tools required. Table 3.18. MSER Steam-Power Piping and Operating Conditions Steam Line Cold Rehesat Cold Reheat Hot Reheat Feedwater Line Heating Steam Heating Steam S8izes and Conditions Leaving Boiler- Line to Line to Line Leaving to Boiler- Line to Line from Superheater Preheater Reheater Reheater Superheater Preheater Preheater Number of pipes 8 2 8 8 8 2 2 Nominal pipe OD, in. 14 35 18 16 12 12 12 Wall thickness, in. 3 0.69 0.5 0.5 1.3 2 2 Pipe material A335, Gr P-22 A155, Gr KC-70 Al55, Gr KC-70 A335, Gr P-22 Al06, Gr C A335, Gr P-22 A335, Gr P-22 Operating temperature, °F 1000 552 650 1000 700 1000 866 Allowsble stress at operating 7,800 15,750 15,750 7,800 16,600 7,800 7,800 temperature, psi Flow rate, 1b/hr 10.1 x 10° 5.1 X 106 5.1 X 10% 5.1 X 108 10.1 X 10® 2.9 x 10° 2.9 X 108 Pressure, psia 3600 600 570 540 3800 3600 3500 Specific volume, ft3/1b 0.20 0.88 1.07 1.57 0.029 0.20 0.16 Total volume flow, cfm 33.4 X 10° 78.9 X 10° 90.0 x 10° 132 x 10° 4.9 X 10% 9.6 X 10° 7.9 %X 10% Caleulated velocity, fpm 11.9 X 10° 6.0 X 10° 5.7 x 10° 13.5 x 10° 1,12 x 10° 11.5 x 10° 9.5 x 10° Assumed velocity, fpm 10 to 12 X 10° 5.8 to 7.4 X 10° 5.8 to 7.4 X 10® 15.4 X 10° 15.4 X 10° 1.1 X 10° 1.1 x 102 Total flow area, in.? 481 177 1756 1235 657 138 114 26 o« 97 3.10.3 Containment The primary circulating systems containing the fuel and blanket salts are constructed of HaStélloy N and designed for about 150 psi and 1200 to 1300°F. These systems — consisting of the reactor, heat exchangers, pumps, and connecting salt piping — are all housed in the reactor cell. This cell volume is contained by a reinforced concrete structure lined with steel plate; beneath the roof plugs are seal pans with welded joints. This containment assures a cell leak rate below 1% of the total cell volume per 24 hr. The reactor cell design pressure is about 45 psig. The cells adjoining the reactor cell contain the boiler-superheaters, reheaters, and coolant-salt circulating pumps. These cells, which are also designed for a pressure of about 45 psig, are of reinforced concrete and are sealed in the same manner as the reactor cell. Pressure-suppres- sion systems are provided for both the coolant and reactor cells. These systems are separate and independent and contain underground water tanks for condensing steam. The amount of water present in the reactor cell proper will be small, probably consisting mainly of the water circulated through the shielding and equipment-support cooling coils. The coolant-salt cells, one for each of the separate coolant circulating circuits, are not intercon- nected, and one cell could not credibly receive more than one-fourth of the total coolant salt. However, it is conceivable that all the approxi- mately 1,000,000 1b of steam and water inventory in the steam-power sys- tem might flow into a single coolant cell. The pressure-suppression system is designed to limit the cell pressure to 45 psi in such an acci- dent. , The reactor and coolant-salt cells and the fuel-processing cell are located in a building with a controlled ventilation system. The usual adsorption and filtration equipment are provided. 3.11 Plant Construction Costs The methods and assumptions used in estimating the MSBR cost conform to those used in the advanced-converter reactor studies;® particular 28 reference was made to the Sodium Graphite Reactor, since its circulating systems were similar to those of the MSBR. The construction cost estimates of the reference MSBR design are listed in Table 3.19 in conformance with the AEC Cost Guide.'! The di- rect construction costs totalled sbout $80.7 million, and to this must be added the indirect costs for engineering, contingencies, etc. These indirect costs, listed in Table 3.20, correspond to about 41% of the direct construction costs and givé total MSBR plant construction costs of about $114 million. The‘exisfence of an‘established molten-salt re- actor industry was assumed in estimating costs covering materials, fab- rication, inspection, tranéportation, installation, and testing. Additional information concerning the cost estimates for the vari- ous accounts are given below. In obtaining the turbine plant costs, the estimates used were influenced by the actual costs experienced in the TVA Bull Run Steam Plant (completed about March 1966). Land and Tand Rights (Acet. 20). An investment of $360,000 was as- sumed for land. This is the same cost that has been allowed in other re- actor studies. As is customary, the land was treated as a nondepreciating capital cost and was subject to a lower fixed charge rate, as indicated in Table 3.19. Structures and Improvements (Acct. 21). The preliminary character of the MSBR study did not warrant extensive optimization of the plant layout. The turbine-room floor dimensions of the TVA Bull Run plant were incorporated in the MSBR drawings. The reactor plant portion of the building was considered in two parts. The portion partially below grade and containing the more mas- sive structures was estimated at $1.30 per cubic foot of building volume. The upper high-bay portion was costed at $0.8O/ft3. These costs do not include the containment, shielding, and overhead cranes, all of which are included in separate accounts. The total estimated direct construction cost of $9.3 million for buildings and structures appears to be typical of 1000-Mw(e) nuclear power stations. Reactor Equipment (Acct. 221). The MSBR reactor vessel is about 14 ft in inside diameter and 19 ft high with torospherical heads; the 99 Table 3.19. Cost Estimate for MSBR Power Station Account No. Item Amount, (in thousands of dollars) 20 21 22 aIncluded in indirect costs and in total plant cost. Land and Land Rights® Structures and Improvements 211 Ground improvements ol 2 3 o4 5 .6 7 Reactor buildingb Turbine building, auxiliary building, and feedwater heater space Offices, shops, and laboratories Waste disposal building Stack Warehouse Miscellaneous Total Account 212 Total Account 21 Reactor Plant Equipment 221 Reactor equipment 222 223 225 226 227 228 d 2 .3 o 5 .6 Reactor vessel and internals Control rods Shielding and containment Heating-cooling systems and vapor-suppression system Moderator and reflector Reactor plant crane Total Account 221 Heat transfer systems 1 2 .3 ) Reactor coolant system .11 Fuel-salt system .12 Blanket-salt system Intermediate coolant system Power system ) .31 Steam generators (boiler-superheaters) .32 Reheaters Coolant supply and treatment Total Account 222 Nuclear fuel handling and storage (drain tanks) Radioactive waste treatment and disposal (off-gas system) Instrumentation and controls Feedwater supply and treatment o1 .2 '3 ‘4 l5 Makeup supply and feedwater purification Feedwater heaters Feedwater pumps and drives Reheat-steam preheaters Pressure-booster pumps Total Account 227 Steam, condensate, and feedwater piping capital expense in estimating fixed charges. ‘iiJ Phoes not include containment cost; see account 221.3. " 866 4,181 2,832 1,160 150 76 30 8,469 9,335 1,610 250 2,113 1,200 1,089 265 6,527 5,054 1,947 6,530 3,323 300 18,832 1,700 450 4,500 470 1,299 1,600 275 407 4,051 Land is classified as a nondepreciating 100 Table 3.19 (continued) Account It Amount No. : : o : . . (in thousands of dollars) 229 Other reactor plant equipment (remote maintenance) .1 Cranes and hoists 500 .2 Special tools : 1,500 .3 Decontamination facilities 1,000 .4 Replacement equipment ‘ 2,000 Total Aceount 229 5,000 Total Account 22 ' - 45,129 23 Turbine-Generator Units 231 Turbine-generator units 19,174 232 Circulating-water system ' : 1,243 233 Condensers and auxiliaries 1,690 234 Central lube-oil system : 80 235 Turbine plant instrumentation ' 25 236 Turbine plant piping _ 220 237 Auxiliary equipment for generator 66 238 Other turbine plant equipment ‘ : 25 Total Account 23 22,523 24 Accessory Electrical Equipment 241 Switchgear, main and station service . 550 242 Switchboards 128 243 sStation service transformers 169 244 Auxiliary generator 50 245 Distributed items 2,000 Total Account 24 2,897 25 Miscellaneous 800 Total Direct Construction Cost 80,684 Privately Owned Plant Total indirect costs (see Table 3.20) 33,728 Total plant cost® 114,412 Less nondepreciating capital Land 360 Coolant-salt inventory 354 Total Depreciating Capital 113,698 Publicly Owned Plant Total indirect costs (see Table 3.19) | 30,011 Total plant cost® 110,695 Less nondepreciating capital Land 360 Coolant-salt inventory 354 Total Depreciating Capital 109,981 cInclu.des land and coolant-salt inventory costs. h 101 Table 3.20. Distribution of Indirect Costs® Percentage of Ingigzct Acc;gtiited Ttem Acc;gziited (in thousands (in thousands of dollars) of dollars) Direct construction cost 80,684 General and administrative 6 4,841 85,525 Miscellaneous construction 1 855 86,380 Architect-engineer fees 5 4,319 90,699 Nuclear-engineering fees 2 1,814 92,513 Startup costsP 0.7 646 93,159 Land 360 93,519 Coolant~salt inventory 354 93, 873 Contingency 10 9,387 | 103, 260 Interest —-private financing 10.8 13,152 114,412 Total indirect costs 33,728 (private financing) Interest — public financing 7.2 7,435 110,695 Total indirect costs 30,011 (public financing) a'Indirec‘l; costs follow those used in the advanced converter reactor studies.? bStartup costs are based on 35% of first year's nonfuel operating and maintenance costs. vessel walls are about 1.5 in. thick and the heads about 2.25 in. Based on fabrication experience with similar vessels and materials, $8.00/1p was used to cover the installed cost of the vessel, supports, etc. There are 534 Hastelloy N tubes 1.5 in. in diameter and 18 in. long, and a like number of tubes 3 in. in diameter and 18 in. long. These were estimated to have an installed cost of $6.00/1b. The cost of brazing the graphite tubes to these Hastelloy N tubes was estimated at roughly $100/braze, or about $107,000. An additional $393,000 was 102 allowed for special inspections, assembly of the graphite, etc., to bring the total cost of the vessel to about $1.6 million. The control rods for the MSBR do not employ expensive drive or scram mechanisms and were not studied in detail for this preliminary report. An allowance of $250,000 was made for the rods. Réihfbrced concrete for'shielding and containment was estimated to cost $80/yd3, in place. The 0.25- to 0.5-in. steel liner plate was esti- mated to cost $l.50/1b, in place. The thermal insulation was considered to have an installed cost of $6.OO/ft2. An allowance of $100,000 was made for water cooling of the structures. The MSBR reactor cell requires heating, cooling, and a vapor-suppres- sion‘system to limit the pressure in an emergency condition, but con- ceptual studies of these systems were not undertaken. An allowance of $1.2 million was made for these items. The graphite used as the MSBR moderator must be of high density and high quality and be closely inspected. About 76% of the core volume is graphite. It is assumed to cost $10/1b based on information from a manu- facturer and the apparent feasibility of extruding the required shapes. The reflector graphite was assumed to be 6 in. thick. The cost of this graphite was estimated at $5/1b. Heat Transfer Systems (Acct. 222). The costs of the shell-and-tube heat exchangers were determined by breaking down each component into weights of shells, tubes, etc., and using typical costs for materials and fabrication to arrive at a total estimated cost per square foot of surface. These values checked well with costs of similar reactor plant heat transfer equipment for both actual equipment and for estimates used in other studies. The cost of Hastelloy N piping carrying molten salts was estimated at $10/1b. | | The costs of salt-circulating pumps were estimated by extrapolating experience with existing molten-salt pumps and using costs for liquid- metal pumps.2® The costs were increased 5% to include supports and by 10% to include installation and testing. The quantity of coolant salt required for one filling of the circu- lating system is about 2833_ft3. At an estimated cost of $1.00/1b, the h 103 coolant-salt inventory cost is about $354,000. This is & nondepreciating type of capital expense and was treated in the same manner as the land cost. Nuclear Fuel Handling and Storage (Acct. 223). No conceptual design work was done on the MSBR fuel- and blanket-salt drain tanks. An allow~ ance of $1.7 million was made for the eight tanks. Feedwater Supply and Treatment (Acct. 227). The estimated costs for the makeup supply, feedwater purification, and feedwater pumps and drives are largely based on values used in other 1000-Mw(e) reactor plant studies and on the TVA Bull Run Steam Plant data. A value of $44/1b was used for the eight Croloy reheat-steam pre- heaters. The two high-pressure low-head 20,000-gpm pressure-booster pumps in the feedwater supply to the boiler-superheater were estimated at $4/gpm capacity. The motor costs were based on unit costs of $lO/hp. The variable-speed drives were also based on unit costs of $10/hp. Steam, Condensate, and Feedwater Piping (Acct. 228). The cost of condensate and feedwater piping for the 915-Mw(e) TVA Bull Run Plant is reported to be $3.62 million. On this basis, the piping for the 1000- Mw(e) MSBR was estimated to be $4.07 million. Other Reactor Plant Equipment (Acct. 229). Maintenance of the MSBR will probably require remotely controlled cranes and hoists and the use of special tooling and remote-brazing and -welding equipment. Decontami- nation and hot storage facilities are needed. No conceptual designs were made for this equipment. The costs listed correspond to judgments, with estimates tending to be high due to lack of design data and of mainte- nance experience. | The MSBR maintenance procedures will involve replacement and subse- quent repair rather than in-place repair of items such as salt pumps and primary heat exchangers. This will entail an inventory of replacement equipment, an expense that could be interpreted as part of the initial capital investment rather than as an operating expense. An allowance of $2 million was made for this replacement equipment. Turbine-Generator Units (Acct. 231). The turbine-generator founda- tions were estimated at $370,000, a more or less standard allowance for 1000-Mw(e) station studies.? Erection costs were taken to be $700,000, 104 again a standard value. The cost of a cross-compounded four-flow turbine- generator unit with 43-in. last-stage blades was based on General Electric Company pricing data, with a 78% discount factor applied to the book wvalue. The excitation equipment was assumed to be of the brushless type, with no provisions for standby excitation. Circulating-Water System (Acect. 232). The estimated cost of about $1.24 million for the circulating-water equipment was taken from the SGR 9,10 cost estimate. The TVA Bull Run cost data were not applicable because the circulating-water installations include provisions for future plant expansion. Condensers and Auxiliaries (Acét. 233). The total cost of the four- section horizontal single-pass 320,000-ft? units for the 915-Mw(e) TVA Bull Run Plant was $1.3 million. This was extrapolated to $1.5 million for the MSBR. The $l90,000 allowance for the MSBR condensate pump was also extrapolated from the TVA data. Central Lube-0il System (Acct. 234). An allowance of $80,000 was made for this account on the basis of TVA cost information. Turbine Plant Instrumentation (Acct. 235). This account covers tur- bine plant control boards and instruments not included with the steam piping (Acct. 228) and instrumentation (Acct. 226). An allowance of $25,000 was made on the basis of the SGR estimate.®’*® (The TVA Bull Run data were not available in a form such that this account could be ex- tracted conveniently.) Turbine Plant Piping (Acct. 236). The TVA Bull Run Plant reported cost of $160,000 was extrapolated to the 1000-Mw(e) plant size; in addi- tion, $12,000 was added for the preheater, booster-pump, etc., to make a total of $220,000 for this account. ' Auxiliary Equipment for Generator (Acct. 237). Although the esti- mated cost of the turbine-generator unit is presumed to include the auxiliary equipment, the preliminary nature of the estimate led to in- clusion of $66,000 for miscellaneous equipment end uncertainties. Other Turbine Plant Equipment (Acct. 238). This miscellaneous ac- count is of little significance in the total cost, and other reactor plant studies have not always included it. On the basis of TVA experi- ence, however, $25,000 has been included in the MSBR estimate. o 105 Accessory Electrical Equipment (Accts. 241—245). This account covers the cost of hundreds of electrical items, such as motor starters, etec., scattered throughout the plant, and amounts to a significant por- tion of the total plant investment. The estimate of about $2.35 million for the total of accounts 242 through 245 is the same as that used in the SGR study.g’lo Account 241, which covers both main and station ser- vice switchgear, was reduced below the SGR estimate because the MSBR has smaller pump motors, utilizes turbine-driven boiler-feedwater pumps, and does not require large motor-driven pumps for emergency cooling of the type needed in the SGR. However, the total of about $2.9 million for Account 24 is only slightly less than the total of $3.0 million used for the SGR. Indirect Costs. The indirect costs, which amount to about 41% of the total direct construction cost, have a very important bearing on the total capital cost and the final production expense. The indirect costs for the MSBR follow those used in the advanced-converter study9 and are listed in Table 3.20. The percentages used appear to be more representa- tive of present practice than those suggested in the AEC cost evaluation handbook.t! Each percentage expense is applied to the accumulated cost total preceding the particular item. The land and coolant-salt inventory costs are included in the in- direct costs so that the contingency and interest costs reflect these expenses. However, the land and coolant-salt costs are deducted from the total plant cost to obtain the depreciating capital outlay. 3.12 Power-Production Cost Power costs are made up of capital charges, operating and mainte- nance costs, and fuel-cycle costs. In computing capital charges, an important quantity is the fixed charge rate. For an investor-owned MSBR plaht, a fixed charge rate of 12%/yr was applied to depreciating capital, while 10%/yr was'applied to nondepreciating capital. These fixed charge rates are the same as those used in the advanced-converter reactor studies;? the distribution of the charge rate for depreciating capital is given in Table 3.21. 106 Table 3.21. Fixed Charge Rate Used for Investor-Owned Power Plants Tten : (%/yr) Return on money invested® 6 Thirty-year depreciationP 1 Interim replacements® | 0. Federal income taxesd 1l Other taxes® 2 Tnsurance other than liability® 0 Total 12.0 &Return was based on one-third equity capital financing, with a return of 9% after taxes, and two-thirds debt capital drawing 4.5% interest. bThe sinking-fund method was used in deter- mining the depreciation allowance (plant life of 30 years assumed). °In accordance with FPC practice, a 0.35% allowance was made for replacement of equipment having an anticipated life shorter than 30 years. dFedera.l income taxes were based on "sum-of- the-year digits" method of computing tax defer- rals. The sinking-fund method was used to nor- malize this to a constant return per year of 1.8%. ®Ihe FPC recommended value of 2.4% was used for "other taxes." A conventional allowance of 0.20% was made for property damage insurance. Third-party liability insurance is listed as an operating cost. : For publicly owned plants, the fixed charge rate employed was 7%/ yr for depreciating capital; the distribution of this chargé rate is given in Table 3.22. For nondepreciating capital the charge rate was 5%/yr . The operation and maintenance charges are given in Table 3.23 and are consistent with those used for the advanced-converter studies;® however, the staff payroll costs were increased by 35%, since preliminary infbrmatidn regarding the proposed revision to Section 530 of the AEC &;J Cost Guide!! indicates that such an increase is required to be consistent 1k 9) 107 Table 3.22. Fixed Charge Rate Used for Publicly Owned Power Plants Ttem Rate (%/yr) Return on money invested 4.00 Thirty-year depreciation 1.75 - Interim replacement 0.35 Local taxes plus insurance 0.90 Total 7.00 Table 3.23. Operation and Maintenance Costs for a 1000-Mw(e) MSBR® Ttem Annual Cost Operating Total payroll, 70 employees with 20% $ 900,000 for fringe benefits and 20% for general and administrative expense Private insurance 260, 000 Federal insurance, at $30/Mw(th) 67,000 Maintenance Repair and maintenance materials 1,065,000 Makeup coolant saltb (2% replacement 7,000 per year) Contract services 72,000 Total operating cost $2,371,000 Unit cost, mill/kwhr(e), 0.8 load 0.34 factor a . . . The operating and maintenance costs associated with the fuel-recycle processing plant are included in the fuel-cycle costs. bMakeup carrier salt for the réactor salt cir- cuits is included under fuel-cycle costs. e o b v e e © 108 with present-day salaries. The operation and maintenance costs associated with the fuel-processing plant could also be included here but, instead, are included under fuel-cycle cost so that the latter can be more di- 'rectly compared with the fuel-cycle costs of reactor plants employing off-plant fuel fabrication and processing. | Combining the capitai costs, operation and maintenance costs, and the fuel-cycle costs gave the power-production costs summarized in Table 3.24 for investor-owned and publicly owned utilities. As shown, the power-production cost would be about 2.75 mills/kwhr(e) in an investor- ovned plant and about 1.73 mills/kwhr(e) in a publicly owned plant. In a utility system complex, the incremental cost between zero-power and full-power operation influences the load factor of an individual plant. This incremental cost for the MSBR is shown in Table 3.25, along with other costs that are independent of power level. As shown, the Table 3.24. Power-Production Cost in 1000-Mw(e) MSBR Load factor: 0.8 Capital Cost Annual Cost (in thousands (27te) (in thousands [migzz?kSEi?e)J of dollars) yr of dollars) Private~-Ownership Financing Fixed charges Depreciating capital 113,700 12 13,644 1.947 Nondepreciating capital 714 10 71 0.010 (1and plus coolant-salt inventory) Operation and maintenance costs 2,371 0.338 Fuel-cycle cost 0.459 Total estimated production cost 2.75 Public Financing Fixed charges Depreciating capital 110,000 7 7,700 1.099 Nondepreciating capital 714 5 36 0.005 (1and plus coolant-salt - inventory) Operation and maintenance costs ' 2,371 0.338 Fuel-cycle cost 0.287 Total estimated production cost 1.73 120 9 109 Tabie 3.25. MSBR Power Cost Breakdown into Fixed and Incremental Costs Financing Ttem Private® PublicP Annual fixed charges, $/kwyr 16.2 9.04 Fixed operating costs,C mill/kwhr(e) 0.39 0.39 Total fixed power cost,® mills/kwhr(e) 2.70 1.68 Tncremental power cost,® mill/kwhr(e) 0.05 0.05 Total power-production cost, mills/kwhr(e) 2.75 1.73 a12%/yr fixed charges on reactor plant, including process- ing plant; lO%/yr inventory charges for nondepreciating items. b7%/yr fixed charges on reactor plus processing plant; 5%/yr inventory charges for nondepreciating items. Not opti- mized for changed conditions. ®Includes 0.06 mill/kwhr(e) for fixed operating cost of the processing plant. dBased on 0.8 load factor. eIncrem_ental cost in going from zero- to full-power opera- tion (0.8 load factor); includes incremental fuel-cycle cost and incremental operating costs. incremental cost between operation at zero power and at full power is only 0.05 mill/kwhr(e) and would provide a high incentive for operating with a high plant factor. Since the reactor has "on-line" refueling, there is no basic reason why the plant has to be shut down except for maintenance; operation with a 0.9 load factor would decrease MSBR power costs to 2.49 mills/kwhr(e) for investor-owned plants and to 1.59 mills/kwhr(e) for publicly owned plants, 110 4. ALTERNATIVE CONDITIONS FOR MSBR DESIGH. As a part of this study, various alternative conditions were con- -sidered for the initial MSBR design in order to improve the plant and to measure the incentive for achieving such conditions. One of the more impoftant conditions is the ability to economically remove protactinium directly from the blanket stream of the reactor. Another desirable con- dition is that of introducing feedwater into the boiler~-superheaters at 580°F rather than 700°F. The ability to maintain a high plaht factor at all times is also of importance. These items, as well as others, are discussed below. 4,1 Protactinum Removal from Blanket Stream Even.though fluoride volatility processing appears to be a satisfac- tory process for removal of uranium, the ability to remove 233Pa directly and economically from the blanket region of an MSBR would significantly improve the performance of the reactor. One possible process involves oxide precipitation of protactinium. Several laboratory experiments27’28 have demonstrated that protactinium can be readily precipitated from a molten fluoride mixture by addition of thorium oxide and that the precipi tate can be returned to solution by treatment with HF. Experimental re- sults also indicate that treatment of protactinium-containing salt with Zr0, leads to oxide precipitation of the protactinium and that after beta decay of the protactinium, the resulting UOp; will react with ZrF, to give UF,. More recent experimental results have indicated another method for removing protactinium directly from the blanket fluid. This involves treating the molten blanket salt with a stream of bismuth containing dis- solved thorium metal. The thorium reduces the protactinium (and also any uranium) to metal, which can then be accumulated on a stainless-steel- wool filter. The deposited metal can be hydrofluorinated and/or fluori- nated to return the protactinium (and any uranium) to the fuel-recycle - process as the fluoride. Thus there is experimental evidence that simple processes are available for direct removal of protactinium from the blanket 111 stream of an MSBR. Practicable application'of such processes would de- crease absorptions of neutrons by protactinium to a negligibly low level and also remove economic restrictions as to the permissible average neu- tron flux in the circulating-blanket volume (related to thorium inventory needs ). The mechanical design of the MSBR with protactinium removal would be essentially the same as that given previously, and the primary change would be in the nuclear design and fuel-cycle performance. The resulting reactor plant is termed the MSBR(Pa) and refers to the initial MSBR design modified for protactinium removal from the blanket stream. Either the oxide-precipitation process or the liquid-metal extraction process appears feasible as a method of removing'protactinium from the blanket stream. It was estimated thafi for either process the blanket- processing costs would be equivalent to those associated with uranium re- covery by fluoride volatility processing plus an additional capital in- vestment for equipment. This additional investment varies with the blanket-processing rate associated with protactinium recovery and is esti- mated to be about $1.65 million at a blanket-salt processing rate of 1000 ft3 per day; for other processing rates the capital investment is esti- mated to vary in accordance with the throughput rate raised to the 0.45 power., The same design methods used for the MSBR were employed in obtaining the MSBR(Pa) design conditions, except that the blanket-processing method and costs were altered in accordance with the above discussion. The re- sulting MSBR(Pa) design conditions are given in Table 4.1, The results of the MSBR(Pa) nuclear performance calculations are summarized in Table 4.2, while Table 4.3 gives the neutron balance for the associated design conditions. These results can be compared with those in Tables 3.7 and 3.8 to give the relative nuclear performance of the MSBR(Pa) versus the MSBR. The essential differences are the decreased neutron absorptions by protactinium and the lower thorium inventory for the MSBR(Pa) design conditions. In obtaining the reactor design conditions, the optimization pro- cedure considered both fuel yield and fuel-cycle cost as criteria of 112 Table 4.1. MSBR(Pa) Design Conditions Power, Mw Thermal Electrical Thermal efficiency Plant load factor Dimensions, ft Core Height Diameter Blanket thickness Radial Axial Reflector thickness Volume fractions Core Fuel salt Fertile salt Moderator Blanket Fertile salt Salt volumes, ft> Fuel . Core Blanket Plena Heat exchanger and piping Processing Total Fertile Core Blanket Heat exchanger and piping Processing Total Salt compositions, mole % Fuel LiF BeF» UF, (fissile) 2225 1000 0.45 0.80 166 26 147 345 33 717 72 1121 100 24 1317 63.6 36.2 0.22 113 Table 4.1 (continued) Salt compositions, mole % (continued) Fertile LiF 71.0 UF, (fissile) 0.0005 Core atom ratios Thorium to uranium , 41.7 Carbon to uranium 5800 Fissile inventory, kg 681 Fertile inventory, 1000 kg 101 Processing Fuel stream Fertile stream Equivalent cycle time, days Uranium removal process 42 55 Protactinium removal None 0.55 process _ Equivalent rate, ft3 per day Uranium removal process 16.3 23.5 Protactinium removal None 2350 process Unit processing cost, $/ft’ 190 652 aEquivalent unit processing cost based on recovery of ura- nium by the flouride volatility process and protactinium concen- tration in accordance with protactinium removal rate, which gives the same processing cost as that associated with direct protac- tinium removal from fertile stream. performance. Although most.emphasis was given to obtaining a low fuel- cycle cost, a fractional weight was given to maximum fuel yield, so the design conditions do not correspond to minimum fuel-cycle costs. This is illustrated in Fig. 4.1, which shows the minimum cost as a function of fuel yield. The design conditions for thé MSBR(Pa) and also the MSER correspond to the designated points in Fig. 4.l. The MSBR(Pa) fuel-cycle costs are listed in Table 4.4. Comparison with results in Table 3.9 shows that direct protactinium removal from the blanket stream reduces fuel-cycle costs by about 0.1 mill/kwhr(e) 114 Table 4.2. Nuclear Performance for MSBR(Pa) Design Conditions Fuel yield, % per annum 7.95 Breeding ratio 1.0713 Fissile losses in processing, atoms per fis- 0.0051 sile absorption Neutron production per fissile absorption (ne) 2.227 Specific inventory, kg/Mw(e) 0.681 Specific power, Mw(th)/kg 3.26 Power density, core average, kw/liter , Gross 80 In fuel salt 473 Neutron flux, core average, neutrons/cmz-sec Thermal 7.2 X 10*% Fast 12.1 X 104 Fast, over 100 kev 3.1 X 104 Thermal flux factor in core, peak-to-mean ratio _ Radial 222 Axial - 1.37 Fraction of fissions in fuel stream 0.996 Fraction of fissions in thermal-neutron group 0.815 Mean n of 233y 2.221 Mean 1 of 23°U 1.958 and the thorium inventory requiréments by nearly a factor of 3. Table 4.5 summarizes fuel-cycle costs for privately and publicly financed MSBR(Pa) plants, while Table 4.6 gives estimated power-production costs. Table 4.7 gives MSBR(Pa) fixed and incremental power costs similar to those given in Table 3.24 for the MSBR. As shown, it is more economical to operate the plant at full power than to let the plant idle at zero power; operation at 0.9 load factor rather than 0.8 would lead to power- production costs of 2.35 and 1.46 mills/kwhr(e) for private and public financing, respectively. @ 115 Table 4.3. Neutron Balance for MSBR(Pa) Design Conditions Neutrons per Fissile Absorption Material Absorbed Total Absorbed by Fission Produced 232 0.9970 0.0025 0.0058 233pg, 0.0003 233y 0.9247 0.8213 2.0541 234y 0.0819 0.0003 0.0008 235y 0.0753 0.0607 0.1474 2367 0.0084 0.0001 0.0001 237y 0.0009 238y 0.0005 Carrier salt (except 6Li) 0.0647 0.0186 611 0.0025 Graphite 0.0323 135%e 0.0050 149gm 0.0068 151lsm 0.0017 Other fission products 0.0185 Delayed neutrons lost® 0.0049 Leakagéb 0.0012 Total 2.2268 0.8849 2.2268 a‘Delayed neutrons emitted outside core. bLeakage, including neutrons absorbed in reflector. Table 4.4. TFuel-Cycle Cost for MSBR(Pa) Design Conditions Cost (mill/kwhr) Fuel Stream Fertile Stream Total Grand Total Fissile inventory® 0.1125 0.0208 0.1333 Fertile inventory 0.0000 0.0179 0.0179 Salt inventory 0.0147 0.0226 0.0373 Total inventory | 0.188 Fertile replacement 0.0000 0.0041 0.0041 Salt replacement 0.0636 0.0035 0.0671 Total replacement _ 0.071 Processing 0.1295 0.0637 0.1932 Total processing ' 0.193 Production credit 0.105 Net fuel-cycle cost 0.35 aIncluding 233pg, 233y, and 235y. 116 : ORNL-DWG 66-7456 0.6 / MSBR DESIGN POINT \ FUEL CYCLE COST [mitts/kwhe}] 04 / MSBR(Pa) DES'DW/ 0.3 g 02 2 3 4 5 6 7 8 9 FUEL YIELD (%/yr) Fig. 4.1. Variation of Fuel-Cycle Cost with Fuel Yield in MSBR and MSBR(Pa) Concepts. Table 4.5. MSBR(Pa) Fuel-Cycle Costs for Investor-Owned and Publicly Owned Plants Load factor: 0.8 Cost [mill/kwhr(e)] Item Privated PublicP Ownership Ownership Fissile-, fertile-, and carrier-salt inventory 0.188 0.094 Replacement cost of fertile and carrier salts 0.071 , 0.071 Core- and blanket-processing costs Operation and maintenance 0.069 0.069 Capital costs 0.124 0.073 Bred fuel credit (0.105) (0.105) Net fuel-cycle cost 0.35 0.20 ®Based on lZ%/yr capital charges for processing plant and inventory charges of 10%/yr. ’ bBased on ‘7%/y"r capital charges for processing plant and inventory charges of 5%/yr. h 117 Table 4.6. Power-Production Costs of 1000-Mw(e) MSBR(Pa) Load factor: 0.8 Power Cost [mills/kwhr(e)] Item Private Public Financing Financing Fixed charges Depreciating capital 1.947 1.099 Nondepreciating capital (land plus 0.010 0.005 coolant-salt inventory) | Operation and maintenance costs 0.338 0.338 Fuel-cycle cost 0.348 0.202 Power-production cost 2 .64 1.64 Table 4.7. MSBR Power Cost Breakdown into Fixed and Incremental Costs Private® PublicP Item . . . . Financing Financing Annual fixed charges, $/kwyr 15.9 8.90 Fixed operating costs,® mill/kwhr(e) 0.38 0.38 Total fixed power cost, d mllls/kwhr(e) - 2.65 1.65 Incremental power cost,® mlll/kwhr(e) —0.01 —0.01 Total power-production cost,‘mllls/kwhr(e) ' 2.64 - 1.64 12%Vyr fixed charges on reactor plant, 1nclud1ng Processing "‘plant lO%Vyr inventory charge for nondepreciating items. 7%Vyr fixed charges on reactor plus processing plant; 5%/yr inventory charges for. nondeprec1at1ng items. Not optimized for changed condltlons. “This 1ncludes 0. 055 mlll/kwhr e) for fixed operatlng cost of the processing plant. dBased on 0.8 load factor. ®Incremental cost in going from zero to full-power operation (0.8 load factor); this includes incremental fuel-cycle cost and incremental operating costs. 118 For comparison, a summary of the power cost and fuel-utilization characteristics of the MSBR(Pa) and the MSBR is given in Table 4.8. - Table 4.8. Comparison of Power-Production Cost and Fuel-Utilization Characteristics of the MSBR(Pa) and the MSBR MSBR(Pa ) MSBR Specific fissile inventory, kg/Mw(e) 0.68 0.77 Specific fertile inventory, kg/Mw(e) 105 268 Breeding ratio 1.07 - 1.05 Fuel-yield rate, %/yr 7.95 : 4 .86 Fuel doubling time,2 years 12.6 | 20.6 Private Public Private Public Financing TFinancing Financing Financing Capital charges, milils/kwhr(e) 1.95 1.10 1.95 1.10 Operating and maintenance cost, 0.34 0.34 0.34 0.34 mill /kwhr(e) Fuel-cycle cost,b mill/kwhr (e) 0.35 0.20 0.46 0.29 Power-production cost, mills/kwhr(e) 2.64 1.64 2.75 1.73 % Inverse of the fuel-yield rate. bCosts of on-site integrated processing plant are included in this value. 4.2 Alternative Feedwater Temperature Cycle The 700°F feedwater temperature and the 650°F temperature of the "cold" steam to the reheater in the initial design were dictated by the 700°F liquidus temperature of the coolant salt. It would be an obvious advantége if it were not necessary to divert almost 30% of the throttle steam for heating of the feedwater and reheat steam, since this diversion leads to a loss of available energy. An even more significanthsaving could be achieved if the 9.2 Mw(e) of power required to drive the feed- water pressure-booster pumps could be eliminated; also, removal of the - reheat~-steam preheaters and the booster pumps would reduce capital in- vestment requirements. Thus, savings can be achieved by lowering the (1] 119 temperature of the steam-cycle fluid entering the boilers and reheaters. To determine the incentive for developing a coolant salt having a low liquidus temperature, the MSBR steam-power cycle was studied with condi- tions of 580°F feedwater temperature and 550°F reheat steam. In order to differentiate and compare cases, use of 700°F feedwater and 650°F re- heat steam is designated case A, while case B represents the alternative conditions. The cycle arrangement for the case B conditions is shown in Fig. 4.2. In this cycle the 552°F steam from the high-pressure turbine exhaust is introduced into the reheaters without preheating. The feedwater is heated from 550 to 580°F by the addition of one more stage .of feedwater heat- ing; steam extracted from the high-pressure turbine is used. The con- densate from this new heater is cascaded back through the feedwater heaters to the deaerating heater in the usual manner. The heat balances and the analysis of the steam cycle with case B conditions were performed in the same manner as for case A conditions.?% Table 4.9 compares the design data for the two cases. Thé elimination of the feedwater pressure-~booster pumps required in case A saves about 9.2 Mw(e) of auxiliary power, which, together with the improvement in the cycle thermal efficiency due to the additional stage of feedwater regeneration, makes about 9.7 Mw(e) additional power avail- able from the casé B cycle. The overall net thermal efficiency is thus improved from the 44.9% dfitained from case A to 45.4% in case B. The cost estimates for fihe MSBR steam station are given in detail in Section 3.11 for caseJAQ‘ To qémplete the discussion of case A versus case B conditions, the cost eSfifiates forfthe affected items of equipment were compared; the reéfiltégare summarized in T&bfié 4.10. As shown, the case B arrangement requirés about $465‘OOO less capitél_expenditure,,pri- marlly due to removal of the pressure-booster pumps . ¥ The net effect of changing from case A to case B condltlons, assuming that inexpensive coolant salt is available for both cases, is to increase *¥In this cost study it was assuimed that the 580°F llquldus-temperature coolant salt has the same cost (about $1.00/1b) as the MSBR coolant salt. 120 g % 3 : E = Jf g |1 4 t ; 2 s & EE 1 3 ! & - NI L = _ | Hi-iszszai: 0 MMT " “ M—m = m e 3N ES R =1 ' m o = ¥ fl & wm qlll “1a8 u q' _— | z “ z m O — " i “ ‘M 1 m m « - - . - | g 5242 | | 1 T, 8 ggf mmmn_ _ | e U L omogi% gl b ! ! bogEEEigLige ! : Il d 51 HicEfp iz " .l m mu_l |||||||| - ”m.p_.wuu-nrumnn i 3 3! g 41 " i 1 & | ‘ { — i, =1 SENERATOR ‘A" anoss TMT B304 Mve 5,008, 9500 - 12 WEMEATER 00 Mwt T T T AL — TSR 4 f) o] I|||_+"|.|4 1 | ) b, _ . P T le (T T g I b PRI TH | oo LTI B % _ ”|+J@" EITTO | L i _ a1 TR L 2 — o I D wat-—riH s g o T H .mulnlnlllln_ I m —-ll I-_mutfl_ll..l. “ —u |||||| J m w. 4 — - ) [ L =g ——1 | M 2 i I ] ” 1 - o | “ "“ .u._r ———sma—— " { .P“I “ ¢ e e - n P AT ' Wm _ I ! i i1 1 ! o - - | Do bty T B : | _ g e | H : & 1 e | . _ 3 o BT e s BT @ 3 B T | | b m e e TR ¥ F oo b e : i I . _ [ ~ imaa o | g (I R | _ | “w b — - o § N I, I “ _ o _ | o _ M._ _ | 4 & .... | T | L 5 = : . || _ _ _ By _ _ | i b | o ‘3 w 5§ — o m 3% = || L _ 6% 4 “ — i _M | I - mm 1 " - o “ ) [ _ 3§ - H — g _ s N | — 3 REM 0 LEGEND waTER —— steam e/ Linsiad obe. "w or ——— COOLANT SALT rlllllll-lé)\P ‘ @ SovDer T Alternate Steam System Flowsheet for 580°F Feedwater. Fig. 4.2. 121 ‘Efi) Table 4.9. MSBR Steam-Power Steam Design and Performance Data for Case A and Case B Conditions Case A — MSBR Steam Cgse B — MSBR Alternative Cycle with 700°F Steam Cycle with Feedwater 580°F Feedwater General performance Reactor heat input, Mw 2225 2225 Net electrical output, Mw 1000 1009.7 Gross electrical generation, Mw 1034.9 1035.4 Station auxiliary load, Mw(e) 25.7 25.7 Boiler-feedwater pressure-booster pump load, Mw(e) 9.2 None : Boiler-feedwater pump steam-turbine power output, Mw 29.3 30.6 j Flow to turbine throttle, 1b/hr 7.152 X 106 7.460 X 10° % Flow from superheater, 1b/hr 10.068 x 10°¢ 7.460 X 10° Gross efficiency 47.83 ' 47.91 Gross heat rate, Btu/kwhr 7136 7124 Net efficiency, % 449 45.4 Net heat rate, Btu/kwhr 7601 7518 Boiler-superheaters Number of units 16 16 Total duty, Mw{th) 1931.5 1837.0 Total steam capacity, 1b/hr 10.068 X 10° 7.460 X 10° Temperature of inlet feedwater, °F 700 580 Enthalpy of inlet feedwater, Btu/lb 769 .2 583.6 Pressure of inlet feedwater, psia ~3800 ~3800 Temperature of exit steam, °F 1003 1003 Pressure of exit steam, psia ~3600 ~3600 Enthalpy of exit steam, Btu/lb 1424.0 1424.0 Temperature of inlet coolant salt, °F 1125 1125 Temperature of exit coolant salt, °F 850 850 Average specific heat of coolant salt, Btu/lb-°F 0.41 0.41 Total coolant-salt flow 1b/hr 58,468 X 106 55,608 X 106 cfs 129,93 123,57 gpm 58,316 55,463 Steam reheaters Number of units 8 8 Total duty, Mw(th) 293.5 388.0 Total steam capacity, lb/hr 5.134 X 106 5.056 X 106 Temperature of inlet steam, °F 650 551.7 Pressure of inlet steam, psia ~570 ~600 Enthalpy of inlet steam, Btu/lb 1323.5 1256.7 Temperature of exit steam, °F 1000 1000 Pressure of exit steam, psia ~540 ~540 Enthalpy of exit steam, Btu/lb 1518.5 1518.5 Temperature of inlet coolant salt, °F 1125 1125 Temperature of exit coolant salt, °F 850 850 Average specific heat of coolant salt, Btu/lb*°F .41 C.41 Total coolant-salt flow : 1b/hr 8.884 X 108 11.744 X 108 { efs 19.742 26.008 : gpm : : 8861 11,714 ; Coolent salt pressure drop, inlet to outlet, psi ~60 ~60 : Reheat-steam preheater Number of units 8 None Total duty, Mw(th) 100.45 Total heated steam capacity, lb/hr : 5.134 X 106 Inlet temperature of heated steam, °F 551.7 Exit temperature of heated steam, °F 650 Inlet pressure of heated steam, psia ~580 : Exit pressure of heated steam, psia ~570 Inlet enthalpy of heated steam, Btu/lb 1256.7 Exit enthalpy of heated steam, Btu/lb’ 1 1323.,5 Total heating steam, 1b/hr - 12.915 X 10° Inlet temperature of heating steam, °F 1000 ; , Exit temperature of heating steam, °F 866 ; \iiJ Inlet pressure of heating steam, psia 3515 Exit pressure of heating steam, psia ! ! | ! 122 Teble 4.9 (continued) Case A — MSER Steam Case B — MSER Alternative Cycle with 700°F Steam Cycle with - Feedwater 580°F Feedwater Boiler-feedwater pumps Number of units 2 -2 Centrifugal pumps Number of stages 6 6 . Feedwater flow rate, 1b/hr total 7152 X 10% 7460 X 10° Required capacity, gpm - 8060 8408 Head, ft ~9380 ~9380 Speed, rpm oo 5000 5000 Water inlet temperature, °F 357.5 357.5 Water inlet enthalpy, Btu/lb 329.5 329.5 Water inlet specific volume, £t°/1b ~0.01808 ~0.01808 Steam~turbine drive . Power required at rated flow, Mw (each) 14.66 15.30 Power, nominal hp (each) 20,000 20,000 Throttle steam conditions, psia/°F 1070/700 1070/700 Throttle flow, lb/hr (each) 413,610 431,400 Exhaust pressure, psia ~7 ~77 Number of stages 8 8 Number of extraction points 3 3 Boiler-feedwater pressure-booster pump ‘Number of units 2 None Centrifugal pump Feedwater flow rate, 1b/hr total 10.067 X 108 Required capacity, gpm (each) 9500 Head, ft ~1413 Water,inlet temperature, °F 695 Water inlet pressure, psia ~3500 Water inlet specific volume, £t3/1b ~0.03020 Water outlet temperature, °F ~700 Electric-motor drive Power required st rated flow, Mw (each) 4.587 Power, nominal hp (each) 6150 the thermal efficiency from 44.9 to 45.4% and to reduce construction costs by about $465,000. The lower construction cost reduces power costs by about 0.008 mill/kwhr(e), while the increased efficiency lowers power cost by about 0.026 mill/kwhr(e) (private financing), to give a total saving of about 0.034 mill/kwhr(e) [0.021 mill/kwhr(e) for public financ- ing]. This saving in a 1000-Mw(e) plant (0.8 load factor) corresponds to about $238,000 per year. The present worth (6% discount factor) of this saving over a 25-year period is about $1.5 million. For several MSBR power plants, the saving would be proportionally greater. Thus, there is an economic incentive for developing a coolant salt with a low liquidus temperature so long as its inventory cost ddes not outweigh the potential saving. If the inventory cost of the coolant salt for case B were about $2.4 million more than that for case A, the potential saving would be W 123 Table 4.10. Cost Comparison of 700°F and 580°F Feedwater Cycles for MSBR2 Number oose A — 700°F Case B — 580°F of Feedwater Feedwater Units Feedwater pressure-booster pumps 2 $ 400,000 None Reheat-steam preheaters 8 180,000 None Special mixing tee 5,000 None Feedwater heater No. OP None $ 150,000 Charge for extra extraction noz- None 45,000 zle on turbine for heater No. O Boiler-superheaters 16 6,000,000C 5,900,000 Reheaters 8 2,720,000€ 2,880,000 _ $9,305,000 $8,975,000 Cost differential Direct construction cost $ 330,000 Total construction costf $ 465,000 %Table shows only those costs different in the two cycle arrange- ments and is not a complete listing of the turbine plant costs. bThe high-pressure feedwater heater added in case B was designated "No. 0" in order not to disturb the heater numbers used in case A. “Estimated on basis of $130/ft2. dpstimated on basis of $140/ft2. ®Estimated on basis of $125/ft2. fIndirect costs were assumed to be 41% of the direct costs. cancelled by the increased coolant-salt inventory cost (for a privately owned plant). 4e3 Modular-Type Plant An important factor in low poWer costs is the ability of the power plant to maintain a high plant-availability factor. Thus design features 124 that can improve this factor are desirable if these features do not them- selves introduce compensating disadvantages. A feature of the MSBR plant design is the use of four heat exchanger circuits in conjunction with one reactor vessel in such a manner that if one pump in the fuel circuit stops, the reactor is effectively shut down. If, on the other hand, it were practicable to have four separate reactor circuits, with each connected to one of the four heat exchanger circuits, stoppage of a fuel pump would shut down only one~quarter of the station capacity, leaving 75% available for power production. In ordér to deter- mine the practicality of using a number of reactors in a single 1000-Mw(e) station, the design features of a modular-type MSBR plant, termed MMSER, were investigated. | The MMSBR design concept considers four separate and identical fe- actors, along with their separate salt circuits. The only connections of the four reactors are through the fuel-recyéle plant. The designs of the heat exchangers, the coolant-salt circuits, and the steam-power cycle remgin essentially as for the MSBR. Each reactor module generates ther- mal power equivalent to that required for producing 250 Mw(e) net. The flow diagram given previously for the MSBR (Fig. 3.7) also is essentiglly valid for the MMSBR. Salt flow rates and capacities of the various components remain as in the MSBR design. Figures 4.3 and 4.4 give plan and elevation views of the four dis- tinct reactor cells, along with their adjacent steam-generating cells. Any reactor module can be shut down and serviced while the other three remain operating. The reactor core consists of 210 graphite fuel cells operating in parallel within the reactor tank. The design of the graphite tubes sepa- rating the fuel and blanket salts is similar to that used in the MSBR. The reactor core region is cylindrical-with a diameter of about 6.3 ft and a height of about 7.9 ft. The reactor vessel is agpproximately 12 ft in diameter and about 14 ft high. Except for the use of four reactor vessels instead of one, all design features of the MMSBR are similar to those of the MSBR. The design conditions associated with one reactor module are summarized in Table 4.11. (;J STE TERS GENERATORS COOLANT PUMPS PRIMARY EXCHANGER AND AN FUEL PUMP BLANKET HEAT THERMAL EXCHANGER SHIELD AND PUMP o INSULA REACTOR Fig. 4.3. Modular Plant Reactor Cell — Plan AA. Y ORNL DWG 66-7115 CONTROL ROD DRIVE FUEL AND BLANKET PUMP DRIVE MOTORS T SALT PUMPS STEAM GENERATORS REACTOR REHEATERS BLANKET HEAT EXCHANGER = 60-0" -3 Eé =3 STEAM PIPING PRIMARY HEAT = EXCHANGER =3 Fig. 4.4. Modular Plant Reactor Cell — Elevation BB. 127 Table 4.11. MMSBR Design Conditions for One Module Power generation Thermal Electrical Thermal efficiency Plant factor Dimensions, ft Core Height Diameter Blanket thickness Radial Axial Reflector thickness Reactor volumes, ft° Core Blanket Salt V‘olumes‘,rft3 Fuel Core Blanket Plena Piping Heat exchanger and pump Processing Total - Fertile Core Blanket , . Heat exchanger and piping Processing - Total Salt compositions, mole % Fuel ' o TIiF BeF» UF, (fissile) Fertile S TLiF BeF> ThF, Average power density in core fuel salt, kw/liter 556 250 45 0.80 245 1.000 41.5 22 25 82 7.5 185 12 1000 25 24 1061 63.6 36.2 0.22 71 473 128 The nuclear and fuel-cycle performance of a four-module plant gen- erating 1000 Mw(e) was studied both for protactinium removal from the blanket stream and for the case of no direct protactinium removal. The same methods and bases as those for the MSBR studies were employed. Analo- gous to previous terminology, these cases are termed MMSBR(Pa) and MMSER. The results obtained are summarized in Table 4.12. Comparison with the results obtained for the MSBR(Pa) and the MSBR indicates that the nuclear and fuel-cycle performance of a modular-type plant compares favorably with that of a single-reactor-type plant; the modular plant tends to have slightly higher breeding ratio, fissile inventory, and fuel-cycle cost. Table 4.12. Nominal Nuclear and Fuel-Cycle Performance o of 1000-Mw(e) Modular Plants Investor-owned plant: 0.8 load factor MMSBR (Pa.) MMSBR Fuel yield, % per year 7.3 5.0 Breeding ratio 1.073 1.053 Specific fissile inventory, kg/Mw(e) 0.76 0.80 Specific fertile inventory, kg/Mw(e) 125 310 Fuel-cycle cost, mill/kwhr(e) 0.38 0.48 Doubling time, yr® 13.7 20 % Inverse of fractional fuel yield per year. Capital cost estimates were also made for the modular plant. The primary difference between the MMSBR and MSBR-type plants is the use of four reactor vessels and cells in the modular plant rather than the one in the MSBR. However, the reactor vessels in the modular plant are smaller, and their combined cost is not much more than that of the single 1arge vessel. Also, the modular plant permits better placement of cells and a reduction in building volume. The resultant capital cost estimate for the modular plant was essentially the same as that obtained for the single-reactor plant. Using a cost estimate of $112/kw(e) for a privately owned plant, along with the MSBR estimate for operation and maintenance .costs, and the fuel-cycle costs from Table 4.12 gives the power-generation 129 costs summarized in Table 4.13. These costs are virtually the same as those for the MSBR-type plants (see Table 4.8) and thus indicate the desirability of a modular-type plant if the plant availability factor is improved by its use. Table 4.13. Power-Production Costs for Modular-Type Molten-Salt Breeder Reactors Investor-owned plant: 0.8 plant factor Cost [mills/kwhr(e)] MMSBR(Pa,) MMSBR Fixed charges 1.93 1.93 Operation and maintenance costs 0.34 0.34 Fuel-cycle costs? 0.38 0.48 Total power-production costs 2.65 275 aCapital charges of processing plant are included in fuel-cycle costs. 4.4 Additional Design Changes In reactor design studies it often occurs that certain features of the detailed design undergo changes as more understanding is obtained of the overall problems and as new wajs are discovered to solve a given de- sign problem. Such changes'have*taken place during the MSBR design studies; of these, the most important are those associated with the pri- mary heat exchanger designs and the pressures that exist in the various circulating-salt systems. The revised design conditions are discussed below. An objectional feature of the MSBR heat exchanger design shown in Fig. 3.20 is the use of expansion bellows at the bottom of the exchanger. These bellows permit tubes in the central portion of the exchanger to change in length relative to those in the annular region due to thermal 130 conditions. Since such bellows may be impractical to use under reactor operating conditions, a new design was developed that eliminated them. Figure 4.5 shows the revised heat exchanger design. The expansion bellows were eliminated, and changes in the tube lengths due to thermal conditions are accommodated by the use of sine-wave type of construction, which permits each tube to adjust to thermal changes. In addition, the coolant salt now enters the heat exchanger through an annular volute at the top and passes downward through a baffled outer annular region. The coolant then passes upward through a baffled inner annular region and exits through a central pipe. | In Fig. 4.5, the flow of fuel salt through the pump is reversed from that shown in Fig. 3.20 in order to reduce the pressure in the graphite fuel tubes. TFuel salt enters the heat exchanger ifi the inner annular region, passes downward through the tubes; and then flows upward through the tubes in the outer annular region before entering the reactor. The blanket-salt heat exchanger was also revised to give g design similar to that of Fig. 4.5. The general features of these exchangers and their placement in the reactor cell are shown in Fig. 4.6 (for com- parison with the initial MSBR design see Fig. 3.8). The blanket-salt pump was also altered so that blanket salt leaving the reactor now enters the suction side of the pump. From the viewpoint of reactor safety, it is important that the blanket salt be at a higher pressure than the fuel salt.?® Under such circum- stances, rupture of a fuel tube would result in leakage of fertile salt into the fuel and a reduction in reactivity. In order to achieve this condition with a minimum operating pressure in the reactor vessel, the fluid flow was reversed from that in the initial MSBR design, with fluid leaving the reactor entering the suction side of the pumps. The result- ing flow diagram is shown in Fig. 4.7 (for comparison with initial design see Fig. 3.7). | In addition, it is desirable that any leakage between the reactor fluid and coolant-salt systems be from the coolant system into the fuel or blanket system. In order to achieve these conditions, the MSBR op- erating pressures were revised to those shown in Table 4.14. o 131 ORNL DWG 66-7136 FUEL LEVEL (DUMP) FUEL SALT DUMP TANK o.g" FUEL LEVEL (OPERATING) ! ALT PUMP ; FUEL SALT TO FUEL SALT PUM " & FROM REACTOR D | COOLANT SALT FROM | STEAM HEAT EXCHANGERS | I —COOLANT PaSS il SEPARATING BAFFLE I4l_4u | | 1§ ; i i FLOW ARRANGEMENT | ; i 1R FUEL SALT-IN TUBE SIDE | it b | (LR COOLANT SALT-IN SHELL SIDE 1l 10" 1[I} | ' 6'0'bo - . (it “’1 K =: i \I! ‘ 1‘1 . U | | ~I ]I 7 A i AT, AHIA A N S | | e TO FUEL i ¥ DRAIN TANKS ™ COOLANT SALT TO BLANKET HEAT EXCHANGER Fig. 4.5. Revised Fuel Heat Exchanger for MSBER. i ! | i i i 5 i 132 ORNL DWG 66-7109 FUEL PUMP . BLANKET PUMP MOTOR SgyTEOL ROD MOTOR IVE - CONSTANT — & SUPPORT |~P=T" HANGERS FUEL DUMP | o+ 4 ; TANK WITH mw [ oy W FTTwhRTTTT - Lt COOLING COILS FOR AFTER HEAT REMOVAL , FT. DIA. . ¢ CORE ~ TOR "+ VESSEL , SALT ' %L DIST. v - +. PLENUMS ) . * . v .‘.' “ 'l ’ PRIMARY — SRR HEAT | * LA EXCH. * «' s ‘4 "‘ _ " . ' - ; . ‘. 'l:: L Y LS. ¥ . B ..n v - . “ & < PR v CESE P TE) 5 '_::‘ ,.:"‘ » ¢ @ 3 . * & ah LR : . ,~.."' -' - . v I o. *. - REACTOR CELL HEATERS " e o '.¢ ° . e . e ae " -, - ]. . . ‘A.'. 0.. S . h. [P 704.[\‘ & ! ’ o : * L s A v " .. \En L] n “: ot e Fig. 4.6. MSBR Cell Elevation Showing Primary Heat Exchangers and Their Placement. Juizser REACTOR . VESSEL HO5°F BLANKET SALT DRAIN TANKS LEGEND BLANKET e o s COOLANT —--— STEAM -———— H,0 - o _10% b/ " PecemmnPsia MevreoemnnBlU./ 1D, N . Fraeze Valve FUEL SALT DRAIN TANKS Fig. 4.7. Revised MSBR Flow Diagram. NET EFFICIENCY 449 % - . ORNL DWG 66-7022 5.134 # _ f - | 10.067 # 7152 # 1 1518.50-540p-1000° | 550p - 1000°F F —— T - : i i il424h-35|5r|000' i > 1 ] i . ' GEN. 3600p- 1000°F | , ; HP 201t Yue ) | surgiNg [P TURBINE [ 5272 Mwe | | | Gross BOILER | - . oo S - SUPERHEATERS | 2seis| 3 | 850°F \ J\ )L 3 {" {'" 551.7° ! ! COOLANT SALT GEN. PUMPS | TURBINE [ TuRBINE [] 5077 Mwe vl Gross 130 f1 Ysde : Q"uzs—? | L | wor | REHEAT STEAM ; bl ; | ! PREHEATERS CONDENSER 8 FEEDWATER i 3800 p_foo o 1307.8h SYSTEMS ] | |m692h 3500p-866°F 5000 550.9° = mash 1| ZOE| geanl 57°"'°5°'Fj3\ usp-eesF 1 | 766.4h BOSSTER MIXING TEE PUMPS COOLANT SALT PERFORMANCE DRAIN TANKS NET OUTPUT 1,000 Mwe GROSS GENERATION 1034.9 Wwe BF BOOSTER PUMPS 9.2 Mwe STATION AUXILIARIES 25.7 Mwe REACTOR HEAT INPUT 2225 Mwt NET HEAT RATE 7,601 Biu/kwh eeT e e i a2 s g £ o e o e AR e 0 Bt T B T e 134 Table 4.14. Pressures in Various Parts of Revised MSBR Salt Circuits Flow diagram given in Fig. 4.7 Nominal Location Pressure | (psig) Fuel-salt system Core entrance 50 Core exit 25 Pump suction 10 Pump outlet , 150 Heat exchanger outlet 60 Blanket-salt system Blanket entrance | 66 Blanket exit 65 Pump suction 64 Pump outlet 155 Heat exchanger outlet 67 Coolant-salt system Pump suction before boiler-superheaters 130 Pump outlet before boiler-superheaters 280 Inlet to fuel heat exchangers 220 Outlet from fuel heat exchangers 160 Outlet-inlet to blanket heat exchangers 142 Pump suction before reheaters 130 Pump outlet before reheaters 240 Reheater outlet 220 As given in Table 4.14, the minimum pressure difference between the core and blanket regions is about 15 psi plus the static head differential or a minimum total difference of about 30 psi. If it is desirable to increase this pressure differential, the blanket-salt pump could be changed so that it discharges into the reactor blanket'region, giving a minimum differential pressure between the core and blanket fluids of about 120 psi. Whether this change is necessary or-whethér it would increase the reactor vessel design pressure is dependent upon the safety criteria that need to be satisfied. A design pressure of 150 psia was used in determining the thickness of the MSBR reactor vessel. 135 5. ALTERNATIVE MOLTEN-SALT REACTOR DESIGNS A number of possible molten-salt reactor designs were considered, and some of these are discussed below. Generally, the alternative de- signs were studied only in concept and not in detail, so the results are more qualitative than those given previously. Also, the technology re- quired for these alternative designs is relatively undeveloped, although there are experimental data which support the feasibility of each con- cept. An exception is the molten-salt converter reactor (designated MSCR), which was studied in detail by Alexander et al.20 and whose appli- cation essentially requires only scaleup of MSRE and associated fuel- processing technology. However, the MSCR is not a breeder, although it approaches a bréak—éven breeder system. It is included to place molten- salt breeders and converters in perspective relative to nuclear perfor- mance, fuel-cycle cost, and power-production cost. The terminology employed for each design concept will be discussed first, along with a summary of the associated design conditions and fuel- cycle performance. Additional information for each concept is given in individual sections below. 1In all cases, a 1000-Mw(e) power plant is considered. The designations MSBR(Pa) and MSBR have the same meanings as before and represent the reference breeder reactor design with and without di- rect protactinium removal from the blanket stream, respectively. The MMSBR(Pa) designation also has the same meaning as before and represents the modular version of the MSBR(Pa). These concepts were presented above and are included here for-completeneés.- The MSBR(Pa-Pb) designation refers to the MSBR(Pa) modified by use of direct-contact cooling of the molten-salt fuel by molten lead. Lead is immiscible with molten salt and can be used as a heat exchange medium within the reactor vessel to significantly lower the fissile inventory external to the reactor. The lead alsc serves as-a heat transport medium between the reactor and the steam generators. The SSCB(Pa) designation refers to a Single-Stream-Core Breeder with direct protactinium removal from the fuel stream. This is essentially a single-region reactor having fissile and fertile material in the fuel s et e 1o A s e e e e A L R s 136 stream, with protactinium removal from this stream; in addition, the core region is enclosed within a thin metal membrane and is surrounded by a blanket of thoriumrcontaining salt. Nearly all the breeding takes place in the large core, and the blanket "catches" only the relatively small fraction of neutrons that "leak" from the core (this concept is also referred to as the one-and-one-half region reactor). | The MOSEL(Pa-Pb) designation refers to a MOlten-Salt Epithermal breeder haVing an intermediate-to-fast-energy spectrum, with direct pro- tactinium removal from the fuel stream and direct-contact cooling of the fuel region by molten lead. No graphite is present in the core of this " reactor. The MSCR refers to a Molten-Salt Converter Reactor which has the fertile and fissile material in a single stream. No blanket region is employed, although a graphite reflector surrounds the large core. The design conditions and fuel-cycle performance for the above- mentioned reactor concepts are summarized in Table 5.1; in all cases the methods, analysis procedures, and economic conditions employed were analogous to those used in obtaining the reference MSBR design conditions. In general, fuel recycling was based on fluoride volatility and vacuum distillation processing; direct protactinium removal from thé reactor system was also considered in specified cases. 5.1 MSBR(Pa-Pb) Concept The MSBR(Pa-Pb) concept is essentially identical to the MSBR(Pa) concept, except that heat is removed from the fuel salt by direct contact with circulating molten lead. The lead is pumped in a circuit external to the reactor and transports the reactor energy to the steam-generating equipment; the circulating-lead circuit takes the place of the coolant- salt circuit used in the MSBR design. A conceptual arrangement for this reactor is shown in Fig. 5.1. The lead is discharged through many jet pumps located under the reactor core; the aspirating action of the jet pumps causes circulation of fuel salt through the fuel tubes of the reactor. To effect this action, each inner fuel tube terminates below the core in a venturi head; lead, flowing " ¥ Ry ¥ Table 5.1. Summary of Design Conditions and Fuel-Cycle Performance for Reactor Designs Studied Design Conditions Reactor Designation® MSBR(Pa.) MSBR MMSBR(Pa) MSBR(Pa-Fb) SSCB(Pa) MOSEL(Pa-Fb) MSCR Dimensions, ft Core Height 12.5 12.5 7.9P 12.5 16.0 3.0¢ 20.8 Diameter 10.0 10.0 6.3b 10.0 9.8 6.5¢ 16.6 Blanket thickness : Radial 1.5 1.5 2.0 1.5 1.2 3.0 Axjal 2.0 2.0 2.0 2.0 0.0 Volume fractions, core Fuel 0.169 0.169 0.17 0.169 0.193 0.5 0.105 Fertile 0.073 0.074 0.05 0.076 0.0 0.0 0.0 Moderator 0.758 0.757 0.78 0.755 0.807 0.0 0.895 Salt volumes, ft3 Fuel Core 166 166 166 166 230 63.5 476 External 551 547 574 110 600 0.7 654 Total 717 713 740 276 830 64.2 1130 Fertile, total 1317 3383 1570 1324 983 758 0.0 %See text for explanation of reactor designatiocns. The core dimensions for this case refer to one module of a four-module station. ®For this case, the core had annular geometry; the fuel annulus inside diameter was 3 ft, and the outside diameter was 6.5 ft. LeT Table 5.1 (continued) Reactor Designation® Design Conditions MSBR(P=a) MSER MMSBR(Pa) MSBER(Pa-Pb) SSCB(Pa) MOSEL(Pa-Pb) MSCR Fuel-salt composition, mole % LiF 63.6 63.6 63.6 63.6 71.0 71.0 70.0 ThF4 0-0 O-O ODO 0-0 8068 24-0 16-55 UF, (fissile) 0.22 0.23 0.21 0.23 0.23 5.0 0.45 Core atom ratios Th/U 41.7 39.7 28.4 41.5 37.7 4.76 36.7 c/u 5800 5440 5980 5520 6280 0.0 6525 Power density, core average, Xw/liter Gross 80 80 80 80 66 618 17 In fuel salt 473 473 473 473 341 1236 165 Neutron fluxé core average, neutrons/cm® «sec Thermal 7.2 x 10 6.7 x 104 7.3 x 10*% 6.8 x 10** 6.1 x 10 0.0 x 10'* 1.9 x 10*4 Fast 12.1 x 104 12,1 x 10*% 11.7 x 10*% 12.1 x 10*% 10.0 x 10** 72.2 x 1014 2.7 x 104 Fast, over 100 kev 3.1 x 10 3.1 x 10*% 3.0 x 10*% 3.1 x 10** 2.6 x 10*% 23.3 x 10** 0.7 x 10*4 Weutron production per fissile 2.227 2.221 2.229 2.226 2.226 2.280 2,201 absorption (ne) Nuclear and fuel-cycle performance Fuel yield, % per year 7.95 4.86 7.31 17.3 6.63 10.3 Breeding ratio 1.07 1.05 1.07 1.08 1.06 1.14 0.96 Fuel-cycle cost, mill/kwhr 0.35 0.46 0.38 0.25 0.374 0.13 0.57 Specific fissile inventory, 0.68 0.77 0.76 0.34 0.684 0.99 1.63 kg/Mw(e) dUse of direct-contact lead cooling would lower the fuel-cycle cost to about 0.32 mill/kwhr(e) and the specific fissile inventory to about 0.41 kg/Mw(e). 8eT - 139 CONTROL ROD ORNI~-DWG 66-6677 REFLECT 10-0"DIA. X 10™-0" LEAD LEVEL (OPERA FUEL LEVEL FUEL DUMP TANK WITH AFTER HEAT BLANKET TO HEAT EXCHANGER GAS FUEL LEVEL (DUMP) LEVEL DUMP) LEAD IN | —— - e e S e g - Fig. 5.1. Two-Region Circulating-Lead Reactor — Elevation. upward to this point, discharges horizontally out of the venturi tube and in the process draws fuel salt into the venturi to cause intimate mixing’of the salt and lead. This mixing generates large areas for heat transfer between the salt and lead and resulis in efficient heat exchange between the two media. After passing through the venturi, the lead and salt separate by gravity due to density difference, with the lead flowing dovnward to the lead outlet lines. 140 The separated fuel salt floats on the lead and forms a 4-in.-deep layer. The core fuel tubes are submerged in this salt layer, and open- ings into their annular regions provide flow passages through which fuel flows into the core wvolume. There are no mechanical pumps in the reactor cell. The only heat exchange within the reactor cell is that provided by the direct-contact lead-and~fuel jet pumps. The only iiquid lines leaving the reactor cell are the lead lines and the fuel-processing line, which communicates with the fuel layer at the bottom of the reactor. The blanket probably would be cooled with lead also; however, since the blanket volume is not criti- cal, the blanket salt could be cooled by pumping it through a tube-and- shell exchanger as in the MSBR. Use of lead cooling requires niobium cladding of metal parts of the system. However, this requirement does not appear to introduce a signifi- cant economic penalty. At the same time the primary heat exchangers are eliminated, with their attendant costs and dperating requirements. The significant advantage produced by direct-contact cooling is the reduction in fissile~-fuel holdup external to the core proper. As shown in Table 5.1, the MSBR(Pa-Pb) concept has a very high fuel-yield rate of about 17%VYT, corresponding to a fuel doubling time of 5.8 years. 5.2 8SCB{Pa) Reactor Concept 5.2.1 SSCB(Pa) Reactor Concept with Intermediate Coolant In the single~stream-core breeder reactor, or one-and-one-half re- gion reactor, the fuel salt contains fertile as well as fissile material. Within the core proper there is no separation of fluids, so graphite tubes of the type needed in the MSBR are not required. A thin metallic mem- brane of Hastelloy N, niocbium, or similar structural material about 0.12 in. thick surrounds the core and separates the core region from the blanket region. | The reactor core is cylindrical and is about 14 ft high and about 10 £t in diameter. The core structure is an assembly of graphite blocks with passages for flow of fuel salt. An annular, cylindrical graphite barrier divides the core into two regions so that the fluid makes two 141 passes through the core. Leakage between the regions is permissible, and therefore the barrier can be constructed by simply interlocking the graphite sections. The core structure is built on a tube-sheet-like support plate, which also serves as a flow distributor for the incoming fuel salt and a collector for the discharge stream. Below this plate are the plenum chambers for fuel distribution. These plenums consist of a central cir- cular region and an annular region, which are separated by a curtain- like barrier. The center plenum directs the fuel to the central region of the reactor, while the annular plenum receives fuel salt as it leaves the annular region of the reactor core. Some bypass of fuel salt be- tween the reactor inlet and outlet plenum chambers is permissible. The energy generated in the fuel salt is transferred to an inter- mediate coolant as in the MSBR concept. The steam-power cycle is also the same as for the MSER. The blanket region contains ThF,; in a carrier salt. Neutrons dif- fusing from the core regioh are absorbed by thorium in the blanket to produce about 5% of the bred 233y, Cooling of the blanket stream is done in a manner similar to that used in the MSBR concept. Direct protactinium removal from the fuel stream is an important feature of this concept. The ability to do this practically in the pres- ence of relatively hlgh uranlum.concentratlons has not been demonstrated conclusively; however, the ox1de—pre01p1tatlon process shows promise of being applicable to protactinium removal from molten salts containing both thorium_and uranium. 5 2. 2 SSCBLP&) with D:Lrect Conta.ct Cooling The performance of the SSCB(Pa) can be 1mproved if molten lead is found to be practlcal as a direct~contact coolant for molten salts con- rtalnlng thorium and uranlum This concept which is termed SSCB(Pa-Pb), is shown in Flg. 5.2, which also 1llustrates features of the SSCB(Pa) conqept As in the MSBR(Pa Pb) concept, the lead coolant not only ab- sqrbe fihermal_energy from the fuel salt, but also supplies the motive power for circulating the fuel salt through the core. 142 ORNL-DWG 66-6680 CONTROL ROD DRIVE ;:’/T—FUEL LEVEL | ——LEAD LEVEL _—LEAD RETURN BLANKET RETURN BLANKET TO ! HEAT EXCHANGER | - —ILEAD TO PUMP AND ¢ |HEAT EXCHANGER Fig. 5.2. One-and-One~Half Region Circulating-Lead Reactor — Ele- vation. As illustrated in Fig. 5.2, the reactor core is mounted above a pool of lead. TFuel salt, which is floating on the lead, flows through the suction pipes into the inlet plenum below the central region of the core and then—through the core in a two-pass arrangement. From the outlet plenum the fuel is channeled radially out and down- ward to peripheral lead-activated ejectors. These ejectors discharge the mixture of lead and fuel salt into the lead pool. Dfiring this con- tact the cooler lead extracts heat from the fuel salt. In the pool, the ~less dense fuel salt rises to the top and is returned to the core. The heated lead is piped away from the pool to a pump and is passed through e 143 the steam superheaters and reheaters. Cool lead is returned to the ejectors. The blanket salt may be cooled in a similar fashion, as indicated in Fig. 5.2, or the blanket salt may be passed through a shell-and-tube heat exchanger cooled by lead returning to the fuel loop. Direct-contact lead cooling reduces the external fuel inventory by permitting efficient heat exchange in a system requiring short runs of fuel piping. Although niobium is needed as a structural and/or cladding material in systems containing lead, fewer heat exchangers may be re- quired. As indicated in footnote d of Table 5.1, the SSCB(Pa-Pb) concept gave a fuel-cycle cost of 0;32.mill/kwhr(e) and a specific fissile in- ventory of 0.41 kg/Mw(e). | 5.3 MOSEL(Pa-Fb) Reactor Concept The MOSEL reactor concept has no moderator (in the sense that no material is introduced for moderating purposes) and operates in the intermediate-to-fast energy range (mean fission energy of 10 to 20 kev). The core contains only molten-salt fuel and the lead introduced for cool- ing, while the blanket contains ThF, in a carrier salt. Niobium is used as the structural'or'cladding material where there is the possibility of contact with lead. ’ Figure 5.3 1llustrates the reactor concept; the core is toroidal in shape, having a cross section about 3 £t wide by 4 ft high. The internal diemeter of the torus is 4 ft. The core is in a tank of blanket salt, and except for the lead outiet pool ét the bottom; is nearly surrounded by blanket salt. ’ e Lead is pumped in through a perforated header at the tqp of the toroidal core. The lead falls through the fuel salt and extracts energy from the core. In the process, the falling lead causes circulation of fuel salt w1th1n the core region in a rotatlonal pattern, ‘with salt flow- ing upward on each side of the central region. The central reglon con- tains about 50 vol % lead, and the lead separates from the salt by gravity, with the fuel salt floating on the Tead. "Although a protac- tinium removal scheme was assumed in the nuclear design calculations, 144 ORNL-DWG 66-6673 T0 FutL REPROCESSING 8 oo —mmmm= OFF GAS SYSTEM LEAD COULANT™._ i HEADER N TO BLANKET “ REPROCESSING . ~ COOLING LEAD RETURN ™. \\ ANNULAR CORE _ REGION = ™~ . \\ el __ BLANKET 1 BLANKET REGlONS"—-\\ ‘ ~o ' Ty o / MAKE-UP : \ ! ' ' ' ' = | , v — & v ' f ' : T v v oy ¢ i Y J\\' ' ' v 1 v v t \ I ' ' ' T 1 1 ' 1 ¥ t 1] v l ! ! Y ¥ .o v . ~I7,0" A ' EnS . AV ( v e , - v - FUEL TRAP ] . Al L 1 ¥ — - Y COOLING LEAD L —1 _ TOPUMP ] . 8 HEAT - o B HEAT . SR FUEL MAKE-UP -~ N o\ J ; AN L . FUEL SALT LEAD POUL - ' DRAIN LINE ! u N—————— 3~ - ; LEAD AND BLANKET _ _ * SALT FILL 8 DRAIN Fig. 5.3. MOSEL(Pa-Pb) Lead-Cooled Reactor — Elevation. the reactor performance given in Table 5.1 would change only slightly if fuel recycling was accomplished with only fluoride volatility and vacuum distillation processing. The design shown in Fig. 5.3 is conceptual in nature, and the actual requirements for separation of the salt and lead phases may involve more 145 than simply separation by gravity forces alone. However, mechanical methods of separation are permissible, and preliminary work indicates that they are feasible. Although preliminary, the results obtained for the MOSEL(Pb) concept indicate the potential performance of an inter- mediate-to-fast energy molten-salt reactor and the versatility of molten salts as reactor fuels. These studies also illustrate that MOSEL-type reactors need efficient methods for removing energy from the reactor core without requiring a large fuel inventory external to the core, since the fissile concentra- tion in the carrier salt is high (about 20 times higher than in a thermal reactor). Direct-contact cooling with lead appears to lower the external inventory requirements to a level sufficient for attaining low fuel dou- bling times and low fuel-cycle costs. 5.4 MSCR Concept The molten-salt converter reactor is a single-region single-fluid reactor moderated by graphite, with the fertile material physically mixed with the fissile fuel salt. The graphite is an arrangement of vertical bars, with fuel passages permitting single-pass flow through the core. The reactor concept is described in detail in the report by Alexander et al.?0 The essential differences between the MSCR concept referred to here and that described by Alexander et al. concern the steam-power cycle and the processing scheme. In the previous report, a Loeffler boiler was used in conjunction with a subcritical steam cycle, while here a supercritical steam-power system and once-through boiler-superheaters are considered that are identical to those given for the MSBR. These changes substantially increase the thermal efficiency and lower the unit capital cost of the previous MSCR plant. Also, the previous system did not use vacuum distillation processing, since the discovery of its appli- cation came at a later date. Incorporation of the vacuum distillation process for carrier-salt recovery, as considered here, leads to substan- tial improvements in fuel-cycle performance. The fuel~-cycle cost of the MSCR concept is given in Table 5.1. The capital costs were not studied specifically but should be comparable with those for the MSBR, that is, 146 about $ll4/kw(e). Assuming the operating and maintenance costs to be 0.34 mill/kwhr(e), as for the MSBER, gives power-production costs under 2.9 mills/kwhr(e) based on an investor-owned plant and a 0.8 load factor. &Y " 147 6. EVALUATION Of the reactor designs and concepts considered in this study, the MSBR(Pa) plant appears to have euperior power-production cost and nuclear characteristics, as well as technology requirements that demand only a reasonable amount of developmental effort. The estimated power-production cost of 2.64 mills/kwhr(e) for investor-owned MSBR(Pa) plants with a load factor of 0.8 indicates that their development can lead to large economic savings. Also, the low specific inventory requirements (less than 1 kg of fissile material per megawatt of electricity produced) and the low fuel doubling time of about 12.6 years, which corresponds to a capability for doubling the installed power capacity every 8.7 years; leads to ex- cellent fuel-conservation characteristics. The results obtained for the MSBR design indicate that this plant also has good performance characteristics, although not so good as those for the MSBR(Pa). At the same time, the MSBR plant appears'less demand- ing of its fuel-recycle technology. Molten-salt reactors appear well-suited for modular-type plant con- struction. ©Such construction causes no significant penalty to either the power-production cost or the nuclear performance, and it may permit MSBR's to have very high plant-availability factors. Use of direct-contact cooling of molten salts with lead signifi- cantly improves the potential performance of molten-salt reactors and indicates the versatility of molten salts as reactor fuels. However, in order to attain the technology status required for such concepts, a 51gn1f1cant development program appears necessary. | The molten-salt reactor concept that requires the least amount of development effort is the MSCR, but it is not a breeder system. The equilibrium.breeding ratio and thé power-production cost of the MSCR plant were estimated to be about 0.96 and 2.9 mllls/kwhr(e), resPec— tively, in an investor-owned plant with a load factor of 0.8. Although this represents_excellent performance as an advanced converter, the de- velopment of MSBR(Pa) or MSBR plants appears preferable because of the lower power-production costs and superior nuclear and fuel-conservation characteristics associated with the breeder reactors. 10. 11. 12. 13. 14. 148 References W. R. Grimes, Chemical Research and Development for Molten-Salt Breeder Reactors, ORNL~IM series report to be issued. G. M. Adamson, Materials Development for Molten-Salt Breeder Re- actors, ORNL-TM series report to be issued. W. L. Carter, Fuel and Blanket Processing Development for Molten- Salt Breeder Reactors, ORNL-TIM series report to be issued. D. Scott, Components and Systems Development for Molten-Salt Breeder Reactors, ORNL-TM series report to be issued. J. R. Tallackson, Instrumentation and Controls Development for Molten-Salt Breeder Reactors, ORNL~TM series report to be issued. R. Blumberg, Maintenance Development for Molten-Salt Breeder Re- actors, ORNL-TM series report to be issued. R. W. Swindeman, The Mechanical Properties of INOR-8, USAEC Report ORNL~-2780, Oak Ridge National Laboratory, January 1961. See also Ref. 2. R. C. Robertson, MSRE Design and Operations Report, Part I. 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Hanauer 168. 169. 170. 171. 172. 173-197. 198. 199. 200. 201. 202. 203. 204. 205. 206. 207. 208. 209. 210, 211. 212. 213, 214, 215, 216. 217, 218. 219. 220. 221, 222, 223, 224. 225, 226, 227 . 228, 229, 230. 231, 232, 233, 234. 235, ORNL-3996 UC-80 ~ Reactor Technology TID-4500 P, G. Herndon H. W. Hoffman G. H. Jenks W. H. Jordan Lincoln Jung P. R. Kasten C. R. Kennedy T. W Kerlin H. T. Kerr S. 5. Kirslis J. A. Lane C. E. Larson (K-25) R. B. Lindauer A. P. Litman D. B. Lloyd A. L. Lotts M. I. Lundin H. G. MacPherson R. E. MacPherson C. D. Martin, dJr. F. C. Maienschein W. R. Martin H. E. McCoy H. F. McDuffie D. L. McElroy T. L. McLean J. G. Merkle H. J. Metz A, S. Meyer S. E. Moore W. R. Mixon J. F. Mardock M. L. Myers L. C. Oakes R. C. Olson A. M. Perry T. W. Pickel J. W. Poston J. W. Prados A, 5. Quist J. L. Redford R. C. Robertson D. P. Roux R. Salmon 152 236. Ann W. Savolainen 255. D. R. Vondy 237. C. D. Scott 256. D. D. Walker 238. D. Scott 257. T. N. Washburn 239, J. H. Shaffer 258. G. M. Watson 240, M. J. Skinner 259. J. R. Weir 241. G. M. Slaughter A 260. R. C. Weir 242. A. E. Spaller 26l. A. M. Weinberg 243, I. Spiewak 262, W. J. Werner 244. R. S. Stone 263. J. H. Westsik 245. J. C. Suddath 264. M. E. Whatley 246. J. R. Tallackson 265, J. C. White 247. W. Terry - 266. H. D. Wills 248. R. E. Thoma 267. L. V. Wilson 249. D. E. Tidwell 268. M. L. Winton 250. G. M. Tolson 269—271. Central Research ILibrary 251. D. B. Trauger 272—273. Y-12 Document Reference 252. R. W. Tucker Section 253. J. W. Ullmann 274—~514. Laboratory Records Department 254. W. E. Unger 515, Laboratory Records, RC External Distribution 516. R. A. Charpie, UCC, New York 517. C. B. Deering, AEC, ORO 518. F. C. Di Luzio, Office of Saline Water, Department of Interior 519. M. C. Edlund, University of Michigan 520. E. A. Eschbach, Pacific Northwest Laboratory 521. H. Falkenberry, Tennessee Valley Authority, Chattanooga 522. A. A. Giambusso, AEC, Washington 523. R. E. Hoskins, Tennessee Valley Authority, Chattanooga 52%. Lenton Long, Pacific Northwest Laboratory 525. W. B. McDonald, Pacific Northwest Laboratory 526. M. W. Rosenthal, AEC, Washington 527. S. Sapirie, AEC, ORO 528. M. Shaw, AEC, Washington 529. W. L. Smalley, AEC, ORO 530. J. A. Swartout, UCC, New York 531. A. Taboada, AEC, Washington 532. J. M. Vallance, AEC, Washington 533. M. L. Whitman, AEC, Washington 534-535. General Atomic Library (Attn: S. Jaye, H. B. Stewart) 536. Division of Research and Development, AEC, ORO 537-538. Reactor Division, AEC, ORO 539-860. Given distribution as shown in TID-4500 under Reactor Technology category (75 copies — CFSTI) -