ORNL S o MASTER COPY Distribution Limited to Recipients Indicated EXTERNAL TRANSMITTAL AUTHORIZED ORNL Central Files Number 61-2-46 ) MOLTEN-SALT REACTOR EXPERIMENT PRELIMINARY HAZARDS REPORT NOTICE This document contains information of a preliminary nature and was prepared primarily for internal use at the Oak Ridge National Laboratory. It is subject to revision or correction and therefore does not represent a final report. The information is not to be abstracted, reprinted or otherwise given public dissemination without the approval of the ORNL patent branch, Legal and Information Control Department. OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION - Y Contract No. W-T4O5-eng-26 MOLTEN-SALT REACTOR EXPERIMENT PRELIMINARY HAZARDS REPORT S. E} Beall W. L. Breazeale B. W. Kinyon February 28, 1961 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the | U.S. ATOMIC ENERGY CCMMISSION External Transmittal Authorized ORNL-CF-£1-2-46 7. e e . ’ 1 = ) [ g - . [ SUMMARY ‘The Molten Salt Reactor Experiment (MSRE) is a circulating-fuel, low-pressure, high-temperature reactor. .The major objectives are the demonstration of the safety, dependability, and serviceability-of”sueh a reactdr and the obtalning of additionfil information about graphite and fission-product gases in the environment of an operating molten- salt reactor. . The MSRE fuel is the tetrafluoride of enriched'U235 (UF@) dis- solved in an LiF—Bng carrier, with ZrF4 and ThF, additlons The coolant salt is an LiF-Bng mixture without additives. The core consists of | unclad-graphite pieces held in position By'molybdenum bands. A nickel molybdenum alloy, INQR;B, especially developed'as.a'container material for molten fluorides, is used for all container and structural menmbers in contact with either the fuel salt or the coolant salt. The heat from the reactor is dissipated to the atmosphere through a radiator in - the coolant-salt systemn. The cover gas for ‘both the' fuel- and the eoolan%wsalt systems is hellum The off-gas system is designed to hold up fission gases until the activity level permits discharge to the atmosphere. | The instrumentation and controls are designed to shut down the reactor safely if excessive reactivity occurs. Periodié'Sampling per- mits evaluation of fuel stability and corrosion rates. “ To demonstrate the serviceability of the system, provisions are made for remote and. semidirect maintenance of the equipment in the reactor cell and other regions of high residual activity. Direct maintenance will be. performed in other areas, including the radiator pit. - The possible accidents consldered are reactivity excursions, fuel separatien, loss of flow, controcl rod fallure, and several mechaniéa.]T possibilities for contalnment failure. The maximum credible accident, rupture of tfie fuel- and coolant-salt systems and subsequent spilling into the cell of all of. the salt, would not burst the container. Any escape of fission products from the:container should result in an ex- posure of less than 25 rem to anyone in the reactor building or in the surrounding area. siii- CONTENTS Slmam ........0.‘...........9......‘...............‘.....O....... iii List Of Ta:bles L B B BB B B B BN B BN BN BE BN BN BE BN N BN BL BN BN L BN BN BN BN BN N B IR B B B BN BN BN B AL AR L BN BB B B AL L J vii List Of Figuz‘es ......._..:’.............0..-.......0.90............. l. Introduction .l.0.0l.v.l-'.;.....OQ.......‘.........0...."...... 2, Reactor comlex 0..0...0.....0..0..000..0......0.......‘...0.. 2,1 2,2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 2.10 2.11 General Description ...........Ol....OO....O....'...O... F“lel and- Materia]-s ..0...0.‘...QQU..’.l‘...'."..0......:.000 Reactor vessel .....0..OO....CD..DO-0.0.0...0....0...... Reactor and Coolent-System PUmMPS cecoeecescccsccoscecss Primary Heat Exchanger .,...1............L............. Salt-to-Air Radiator‘;...,..f.......................... Drain and Storage TENKS '« v seeoesesasssossonseseososesss 2.7.1 Fuel Drain TANKS teeeeseescrcssessssessesbocsnes 2.7.2 Flush-Salt Tank ..ccosscescessscceccoscsccssscss 2.7.3 Coolant Drain Tank teeseanoncosesssasacaessnenes 2.T.4 Storage ToNK secesccososessvssessossscssssscsssss Cover-Gas System ..o...........................;....... Freeze VAlves .eccceccscsvcsecscscscsescscsccssssosssses Smpler MeChanism .O...OO0.0;...0.0...OO....Q.....Q...l 'Nonnuclear Heating L3RI BN B BN BN BN NNCRE RN BN NN BN BN N BN NN BN BN BN BN BN NN B N NN BB BN BN Instmentation and Contmls .....GO.O.AD...GG....O..........‘. 3.1 3.2 General ......\0'0.....000’9.0..0900.00’.0.0.0......._Q.O. 3.1.1 controlRequiremnts .'...DO..OG0.;.........‘... (a) Xenon POidsonirlg LI B BB B B B BN B B BN BB BN AR BN B BN BN BN BN NN BB N (.b.)f Power coefficient .0.000...;?....0.0..00..... ' (éj De]-ayed Neutmns ® 0 00 0800 99 S90S ST OCOEN PSSP PSS (d-) Bumup [ B BN BN BN BN RN -BN BN BN BN BE BN BN NN BN BN N BN BN R RN BN R B BN BN NN BN BN BN BN BN BN N 3 .l.2 O-ther Contml Featums .0 .l. -.. 0060008 S0 PP Q0O RPOOSS 3.1.3 Nomnuclear Controls .cceccesessosccsssscssosssne mstments .......9......0..0..00.......lv....QOOQCOOOO. '3.2.1 Reactor-Power Measurement ....cccoeceeessccccsocscs 3.2.2 Fuel INVentory cececescevrecosscrscsccssccsocssssssce -lve viil 1l 3 h‘o 5. T +3.2.3 Nuclear-Instrumenté',,.........,.;.............. 3.2.h Radiation Mbnitofing ...l.......,..,..,......,.. 3.2.5 Pressure Measurements seo0000e00000s0000esssse e I3.2.6 Flow MEasurements a,.;.,,.....a................. 3.2.7 Temperature Measurements .cscosceoevccsccsccssco 3.2.8 ‘Liquid-Level Measurements ..eeoceccecocosecccnsss Reactor msics Data 000.0.00!06000.D0.0.0......0009‘.0;0.0.‘.{.‘ ’Ihe Reactor conmlemnt DO00000D.OOO0.00.....0.0!000000;00.0.. 5.1 Bui‘ldin.g 00.0.00000000.0o.000000..0.....000;00.0..0900000 5!02 Containment_WDOG000DG00.00‘00.0000..00000000'000OOOQ......O.. 5.2,1 Fuel-System Contalner Desigh .ccceccoscseccccoces 522 O-bher Gells 000.0000.0.0.0....COQO.O.OODO.OOG..OGV 5 23 Penetrations .0....0....00.'.00..000000.0.0.00..0 503 Shieldin:g .00000O..0..00...00..00..000000000000000000000 501" Arrmlgement of Equipment'OOOOOO0.0000....GOO..0000.00... 5.5 Mailltenazlce BOOOOOOD....OOOOO0.00..O0.00...0000000{0000.. Construction, Startup, and Operation .eescocceeccoccccccssesse 601 Constmction .0.:..COOOOOGOOOOOOO0.00....OOOOOQObOOODQOQOO'_ 602 Fl\lShwsalt Test o-‘oo_ooooooooo..oooooocoeoeo‘oqoooo‘o‘_ooooooo 6.3. Startup ooooooooonooooonoocoo.ooogoo,ooooo.o"o_ooo;t;’o'oo;oooo' 601I' APProaCh to Po;wer:-...0DOOOOODOOOOOGOOOO‘O‘;00-.0..,00000_..0 6.5 operations Personnell'00000000.0OOO..DO.0400000000000...00. Hazafis Ana]-ysis .OD'OOO000.0..000.0....._.D..OO...‘OUOOGO00..00A.. T.1 Damage to the Primary Container ...... coesoesossacacene T.1l.1 Reactivity ExcursionsS ..cscscccecscescecocssccsscs (a) Star.tupAccid.en‘t».,....,“..,_.....,.o,“..”..... | (b) Graphite Problems .cccesssoosccccsscccsaoees T.1l.,2 Fuel Separation .,.,.....,........o.”..“._,..‘.....‘;,. - T.1.3 Flow Stoppage (a) Fuel-Circulating-Pump Failure .,-,....o..a;.. (b) C@olantPumpFailure cecosscoecsossesosssass () Simultarne@ushmpFailures cosoo0sccacesceos (d) FlO‘W‘ Stoppa»ge in F'Ulel I-O@P soecooecoecosesoo e T.2 7.3 7 L T.1l.4 Control System Fallure cceoseseoscoceossocsccssse 701.5 Drain-Tank Hazards ooooooooooooooocooooo.oooooooo- T.1.6 Other Possibilities for Primary- Container_Dmge 00O OO GO0 0GOS SO OG0 S0 SO0 OSSOSO (a) Freez.e-Vaive DEIAZE «ocococceccecccccecsssons (b) Freeze FLETIZES 0 000 s's0aososcosossossossessss (e) .Exceésive Wall Temperature .scococscocososos (A) Excessive SLIESSES oeococsosscssscssoossssss (e) Corrosion .., T.1.T Detection of Salt Spillage Ru_ptgre of the Secondary Container cesoscccssccoetossoss 70203- MSSile Damge 0000000..000.0000.'0OOOOOOOOOOAGOO»O;. 702.2 -ExceSSive Pressure 6 0OD0D0O S0 OSSO OO0 SDODD SO G®CO0SOOSOOSPOOOS (a) Sal‘b Spi-llage 00000;000'0000..0‘0‘60000000000.00 (-b) Oil"'Line Rupture 00000..00.300lOOOOOOOOOOOQ.O. 702.3 Acts Of Nature .QDOOOO...‘O0600.080.000000’0.000.. (a) Earthquake 00 0D0S$ 000G GO0 OSSO0 SOO000O0CO0O0S S SO0 (b) Flood ocoooooooooocoooooooo'oooo-ooooooooooooo 70201'I' Sabotage oooo.oo.!000.0‘0000000000.000000.00....0. Consequencés of Radloactivity Release from the Secondaw container 0O ® 0 000 O0®®0 & 000G O0C 00O &0 2DGS 0 @ &9 T.3.1 Rupture of the Secondary Container .cccccececocses 7.3.2 Maximm Credible Activity Release .oceocoococeses T-3.3 Beryllium and Fluorine Hazards ceccecccosecsceses {(a) Beryllium ........,...;;,.........,..,.;.... () Fluorine ...coovcecseoccccoscscocceosoocoosass The Site ® 00 0O 00 PSOPO0OOCS OO0 O SO0 O0O00O0C0O0OSS®O0 SO0 S00O0O0S$080L0B00SD0 APBMH *CPROO0DODSSPOOGH GO0 OCHODBOOOTOSSO00O000O0CS0OOS®SDOSOROSSOSOOSO S SO - A B. dhemistry and Corrosion cccccocococcoecoessscosccoscccssscao Hot=Spot ANBLYSLE ooooosecoooooosossoasesoosooonsoossss Specification £or Drain TaNK .eccooseccoccssoscoscoeccssosns Component Development Program in Support of MSRE .cceeese - Calculation to Support Maximum Credible Accident ..oecsoe Graphite conmatibility With Salt 90 ¢ &8 0 0CHSOO0SQO000COSOSSOESS WCES 0 O0 P OO OO TS OO O EPS®O0000O000CO00 S0 SO0 SO SO S 00000000000 O0DOO0ESES -vi- 103 107 117 120 v Table No. 1 2 3 b - 5 “ 6 T 8 9 L \i / 4. £ LISTiOF TABLES o & i f Page Composition and. Properties of Fuel and Coolant forMSRE l.....o..l.ll..l.'.b...............ODOII‘.'. T Composition and Properties of INOR=8 ..e.vveevecsencons 8 'Properties of MSRE- Core Gr&phite cesessecscscessserrees 9 Reactor-Vessel Design Data .........,.;...,;.;......... 13 Design Date for Fuel and Coolant PUIDS .u.ievssssessess 16 Design Data for Primary Heat Exdhangér'......a....;....-' 18 ‘Design Data for Salt-to-Air RadiBtor ...ececeeeeciessss 21 Design Data for Fuel Drain Tank, Coolant Drain Ta-n.k and. FluSh"Salt T&D.k oo--.olnoo-o.._‘_-oooooooonoo-o'o-o 25 Ré&C’bOI‘ PhyBiCS D&ta ....o.ogeoo.i-oo-on-'.-'o.oo.-ooooooo )'I-B ..vj_i.. ¢ Fig. O 00 1 O I & w M M ) O A2 LIST OF FIGURES MSRE Flow Diagram LA R AR R R R R R EEEEENENRNEBENEMENBEERNEESE RSN BN B Artist's Impression of MSRE Arrangement ....csoceceevsoess Reactor-Vessel ASsembly ..ceecescocosscossssssssssssscsss Typical Graphite Stringer Arrangement ........c.eceeveees. Circulation Pump and MOtOT ceeeeeecesocvesvcsccssossscnsss Primary Heat EXChanNgZer ...ecsceesosceccssccassoscsscsasss Salt-to-Air Radiator ....ccecceciecneccncnsncncencnnnnsnes Fuel Drain and F111 Tanks .eecceesecsssscessocossoncascoas Cooling Thimble for Fuel Drain and FAll Tanks ........... Cover-Gas SUPPLY ceccosccssscctssacscsasssossssssssssnsoas Fuel-System Off-Gas Disposal ..ecececececocesoscacncnanns Coolant-System Off-Gas Disposal ..ecevecccrscccsccconanns Radiant Heat Freeze Valve ..ccececoscsccrsssaccsscoscnsnas Arrangement of Ion Chambers ccceeessccsccccscssssosccsssse Plan of Building T503 «eeeeveeoesnnsesnsneosesoneessnnons Shielding and Sealing Membrane for Top of Cell .ecevevess North-South Sectional Elevation of Buillding 7503 eteeeeees Arrangement of Equipment: Plan View ..ccceceecccsccaacss Arrangement of Equipment: Elevation ....ecciv0e... coesas Remote Manipulator and Shielded Control RoOm ...cececeass Afterheat POVET GENETALION «.oeeseseeseossonseoscessonses Map of Cities and Counties Surrounding _ Oak Ridge Area LI B NN B BN B RN BN B R R N R R R R B NN NN R R R R R RN R R R R R IR R ) Plot Plan: Molten-Salt Reactor EXperiment ......eeeeeo.. Estimated Vapor Pressure of Fuel Salt ................,{. Phase Diagram of Coolant Salt ..ceececevisssrsccscssssennes -viii- «) - ACKNOWLEDGMENT The authors are indebted to many members of the Mblten—Sait Reactor Program for their contributions to this report, and to the Program Direétor; R. B. Briggs, for his advice and suggestions. Initial studies of hazard evaluation were made by Messrs. Remo Galvagni, Italy} John W. Holtzclaw, U.S5.A.; Osamn'Kawaguchi, ‘Japan; and Francisco Z. Pines, Spain, students of Oak Ridge‘School of Reactor Technology. Their findings represent a significant contribution to the material of the present report. X - L ok MOLTEN-SALT REACTOR EXPERIMENT PRELIMINARY HAZARDS REPORT - S. E. Beall W. L. Breazeale- B. W. Kinyon 1. INTRODUCTION One of the principal programs of the~0ak Ridge‘National Laboratory is the development of liquid-fueled reactors. Since 1951 the Leboratory. has constructed and operated two experimentallreactors fueled with uranium in aqueous solution and one fueled with molten salt. The first. experiment with each of these concepts demonstrated nuclear ‘feasibility only. IEngineering feasibility, dependability, and other factors were to be determined in later experiments such as the current Homogeneous Reactor Experiment No. 2 and the subject of this report, the proposed Molten~Salt ReactOr Experiment (MSRE). : | The development of molten-salt systems has beenrpursued continuously since 1951, although the ma.jor effort was supported by the aircraft reactor program. Application of the molten-salt reactor to stationary power production has almays been considered desirable for three highly important reasons: , _ , 1. Molten-salt reactors have a great advantage because they have no fuel elements and-consequently none of the problems associated with fuel ' elements. :Burnup is not limited by radiation damage or reactivity loss. There are relatively simple methods for reprocessing fuel and blanket salts, and their reconstitution involves only dissolution of UF4 or ThF4 in a carrier salt with no metallurgical, ceramic, or mechanical processing. 2. Molten-salt reactors can operate at very high temperatureito produce steam at conditions comparable to those for the hest fossil-fuel plants. The use of a fluid fuel, circulating at high‘rate, can be com- bined with large temperature differences in the core and heat-transfer systems to'produce very high power density. High power density and low fuel inventory in the reactor and the processing plants combine to produce ' high specific power. In spite of the high temperature, the operating pressure is <50 psig. -2- 3. The nuclear and physical characteristics of the salt and the use of unclad graphite as a moderator make possible”thé achievement of very good neutron economy. Breeding ofi the t‘horium-U233 cycle with a fuel yield of about 8 per year appears to be attainable. In order to demonstrate that many of these desirable features can presently be embodied in a practical reactor which can be 0perated safely and reliably, and can be serviced without unusual difficulty, the Osk Ridge National Laboratory has proposed recently this molten-salt reactor ‘experiment. An additional important objective of the experiment is to - provide thé first large-écale test of unclad-graphite moderator, fuel salt, and container materials in-lbng-term oPefation"at high temperature and power. ' This is a preliminary report prefiared for review to obtaln approval of the proposed site. \It is based on the present incomplete reactor design and is primarilx‘concérnéd with the hazards of the eiperimfint as it iéflpresently visualized. The hazards studies of the Aircraft Reactor Experimentl and the proposed (but not built) Aircraft Reactor Test® provide a good background for the prdblems presently foreseen and the proposed solutions discussed in this study. Furthermore, experiéncq in cperating three fluld-fuel reactors pyér a perio& of nine years pro- vidés a good hasis for the design criteria and operating practices. Although the general design of the reactor and its facilities has been investigated for several months, detalls are still unsettled and importan£ changes may be made before the designlis completed. | lSu;perscript nurbers refer to similarly numbered items in the 1ist of - references on page 120. 'y 1 -3- 2. REACTOR COMPLEX 2.1 General Description The proposed Molten Salt Reactor Ekperxment (MSRE) is designed for - a heat generation rate of lO Mw by use of principles which will apply to the design of a much larger power reactor A flow diagram for the reactor and coolant systems and an artist 's concept of the facility are presented in Figs. 1 and 2. | | | In the reactor primary system the molten-salt fuel is c1rculated through a cylindrical reactor vessel which contains a graphite core matrix._ Under design operating conditions,‘fuel enters the core at 1175°F and leaves at 1225°F. Then it flows to a 1250-gpm sump~type pump mounted directly above and concentric with the reactor vessel. The pump discharges the fuel through the shell side of a cross-baffled shell-and-tube heat exchanger and back to the reactor inlet. o _ N A coolent salt is pumped through the tubes of the primary heat ex- changer and then through tubes of an air-cooled radiator hy another sump ~ type pump. It flows at a rate of 830 gpm and cycles‘between 1Q259Exand _llQO°F._ The coolant-salt pressure is kept higher than the fuel-salt_ pressure‘to prevent the escape of fuel in the event of a tube failure; | Air is blown over the bare tubes of the radiatorlhy;two’axial blowers_ofi 164,000 cfm total:capacity. ~Electrical heaters on the piping and equipment " of the fuel and coolant systems keep the salt'molten at all times. A liquid-vapor interface is maintained in the reactor fuel systemlin the sump of the pump. Fuel isvbypassed through the sump at a rate of 50 gpm, and the gaseous fission products in the bypass stream are trans- ferred to a helium cover gas. There is a continuous flow of 7 liters/min of helium through the sump to the off-gas system; the helium system is used to pressurize the reactor to 20 psia. In addition to the reactor and coolant systems, the plant is provided with such auxiliaries as drain tanks for fuel and coolant salts, equipment for sampling the fuel in the reactor, alhelium-cOver-gas system, facili- ties for handling radiocactive wastes, and the usual nuclear and process . control instrumentation and plant services. PRIMARY SALT HEAT EXCHANGER 1025 F:‘ UNCLASSIFIED ORNL-LR-DWG 56870 COOLANT PUMP 830gpm LiF — 70% BeF, — 23% ZI'F4 - 5% UF4 - 1‘70 REACTOR VESSEL | i | :1 l ThE, — 1% | | | || FREEZE VALVE' SPARE FiILL AND FLUSH FILL AND DRAIN TANK TANK DRAIN TANK (68 cu ft) (68 cu ft) (68 cu ft) REACTOR CELL 1400°F LiF — 66% BeFE— 34 % AIR 167,000 ctfm 100°F Fig. 1. Flow Diagram of Heat Transfer System. SECONDARY SALT — o | 300°F COOLANT DRAIN TANK (40 cu ft) -1-‘— o) i o w u 7 7, < 4 Iz} =z 35 ORNL -LR-DWG 52876 ‘Fig. 2. Artist's Impression of MSRE Arrangement. ! _ | ; ! - Under normal steady operating conditions the reactor is self- controlling, as a result of the negative temperature coefficients of reactivity of the fuel -and the graphite moderator. The temperature coefficient of -3 x 1072 (Ak/k)/°F of the fuel provides for fast con- trol. The total temperature coefficient is -9 x 10-> (2k/k)/°F, and this coupled with'the'small amount of excess reactivity loafied‘into.the reactor provides the margin of safety against nuclear excursions to excessively high temperatures. . Nuclear control devices are'provided primarily to hold the reactor suberitical below 1000°F during startup, to compensate_for some fission product poisoning and burnup, and to keep the critical temperature below 1300°F during abnormal periods of operation. The reactor power is controlled by regulating the rate of heat removal. The nuclear reaction can be stopped by the control devices-and the system can be shut down completely by draining”the fuel. Fuel addition in large amounts for the complete’loadings will take place in the fuel drain tank. Subsequent additions to compensate for burnup and fission-product poisoning will be made through a sampling and enriching system communicating with the gas space in the pump bowl. The system components are of all-welded construction. CompOnents in the reactor fuel system are connected to the piping by specially de- veloped freeze flanges which utilize frozen salt as a sealant for the high-temperature fluid fuel. .Braaed'connections are planned for'the radicactive auxiliary systems. The use of these joints makes possible remote maintenance of the system following power operation. Except for flanged connections to the primary heat exchanger, the coolant‘system is of all-welded construction and can be maintained directly within a few minutes after shutdown No velves of the ordinary type are used in contact with the fuel or coolant salts. Flow 1s prevented by freezing salt in designated' | sections of pipe. The freeze valves can be thawed in a few minutes and are the best choice for drain valves. 2.2 Fuel and Materials Fuel for the MSRE is a solution of U235Fu ThF), and ZrF), in an 2 L1 F-BeF carrier salt. The composition and properties of the fuel are a , L -T- _ | listed in Tableul..-Li7F_is a salt of. good fluid-flow and heat-transfer properties and low neutron cross section. Low melting points are dbtained'. in mixtures with BeF2 U235Fh is the primary fuel constituent ThFLL is present as a fertile material. The fuel is representative of the. core fuel for a two-region breeder or a one-region U235 burner reactor. Table 1. Composition and Properties of Fuel and Coolant for MSRE.' Fuel Salt Coolant Salt Composition, mole % LiF (99.97% 1i') _ | 70 66 BeF, | 3 3k ZrF), . | 5 ThF), . | 1 U, - - ~L . Physical properties , , l, , Temperature, °F | | 1200 - 1062 . Density, 1b/ft> 15L4.3 120.5 Viscosity, 1b/ft-hr 17.9 . 20.0 Specific heat, Btu/l1b-°F : 0.4 0.57 Thermal conductivity, Btu/hr-ft (° F/ft) 2.75 . . 3.5 Liquidus temperature, °F - 828 + 5 . 84T x5 . Oxygen as 02 or in CO, H20, and other compouhds reacte\with the four- component mixture to precipitate U02; however, if ZrFu.is present in an amount such that Zr/U ~ 3/1, only Zr0, 1s precipitated by reaction with oxygen-containing materials. During handling and while in the reactor, the fuel must be blanketed by an inert cover gas such as heliqm, to pre- vent contamination by gases and vapors containing oxXygen. ~ The coolant.selt'is an LiTF-Ber mixture of composition and properties as shovn in Table 1. The same genefal considerations_that apply to hand- ling of the fuel also apply to the coolant } The principal materlal of construction for the resctor systems is INOR-B, a n1ckel-molybdenum-chromium alloy developed,at‘ORNL Por use with fluoride salts'at'high'temperature; ’The composition and properties of -8~ INOR-8 are shown in Table 2. When the material 1s attacked, chromium is leached from the elloy, resulting in the formation of subsurface voids. Under most circumstences the rate of attack is governed by the rate of diffusion of chromium in the alloy. Measured rates of attack in typical fuel and coolant salts have been less than 1 mpy at temperatures to at least 1300°F. : | I. Table 2. Composition and Properties of INOR-8 Chemicel Composition” II. Physlical Propertiés Element - % | Element Ni, nlino Ba.-l.o_( g 66 - 71) Mn, m&x- MD, mmc. 1500 - ]_.8«;0 Si’ max. CI‘ 6-0 - 8-0 Cu., III.B.X- Fe, max. 5.0 B, max. C " O-Oh‘ - 0008 W, max. Ti + Al, max. 0.50 P, max. S, max. 0.015 Co, mex. Density, g/cc 3 1b/in7 . Melting point, °F Thermel conductivity, Btu/hréft2(°katju, At 1112°F " 1292°F Modulus of elasticity, psi At 1170°F 1290°F - Specific heat (est.), Btu/1b=°F Mean coeff. of thermal. expansion % 0.80 0.50 0.35 0.010 0.50 0.010 0.20 8.79 0.317 2370 - 2430 12.20 13.01 E 26.2 x 102 24,8 x 10 0.095 at 212°F °F in./in./°F_ AP(°F) AL/L (in./in.) 70-1200 *¥-7.81 x 10 1130 8.82 x 10-5 III. Mechanical Properties 1/4 Min. Spec. 2/3 Min. Spec. 4/5 Rup. Str. Stress ' Max. Tensile Yield for for Allow. Temp. Strength Strength 10° hr 0.1 CRU ©Stress (°F) (psi) (psi) (psi) (psi) (psi) 1200 17,100 16,800 © 8,300 7,500 4,950 1250 16,100 16,600 6,200 5,400 3,600 1300 15,000 16,400 4,800 4,100 2,750 1350 13,800 - 16,300 2,050 3,600 3,100 c =0- “Although the salt has moderating propertieés, use of a separate mod- erator has the advantage of reducing the inventory of fuel in the reactor and provides’ for ‘better neutron economy in a. breeder ) Unclad graphite is compatible with salt -at high temperature both in and out of radiatlon and is the preferred moderator. _The‘properties‘of a graphite that satisfies the requirements of the MSRE arerlisted in Table 3. Table 3. Properties of MSRE Core Graphite ' Physical -properties - . Bulk dens1ty, g/cc | | Por051ty ' accessible, % _ inaccessible, % total, % : Thermal conductiv1ty, at amblent temp, unirradlated Btu/hr ft2(°F/ft) parallel with grain normal to grain ‘Temperature coeff1c1ent of linear expansion, Ffl ' i iV ” . parallel with grain normal to grain Matrix coefficlent of permesbility to helium at 70°F, cm®/sec Absorption of salt at 150 psig, vol % Average specific heat at 1200°F, Btu/1b-°F Mechanical strength properties Tensile strength, psi Flexureikstrength,'psi Compressive strength, psi Modulug of elasticity paraliel with grain, psi normal to grain, psi 1.87 - 1.89 n,'? ~ 8.9 ~ 15.9 70 60 -5 1.3 x 10 1.6 x lO-6 1 x 10"lL - 1x107 0.50 Q.42 1500 - 2400 2000 - 3500 8600 1.9 x 102 1.5 x 10 - =10- 2.3 Reactor Vessel The reactor consists of a cylindrical vessel approximately 5 ft in digmeter and 7-1/2 ft high with an inner cylinder which forms the inner wall of the shell-cooling annulus and serves to support the graphite matrix with its positioning and suppbrting members and flow-regulating orifices. Figure 3% is an assembly drawing of the reactor vessel and core. Fluid enters the vessel at the top of the cylindrical section and flows downward in a spiral path along the wall. With the design flow of 1250 gpm in the l-in. annulus, the Reynolds modulus is 22,000. At the estimated heat generation rate of 0.2 w/cm5 in the wall, 23 kw of heat is removed while maintaining the wall temperature at less than 5°F above the bulk fluid temperature. The fuel loses its rotational motion in the lower plenum and then ‘flows upward through the graphite core matrix, which cofistitfites‘about 77 .5% of the core volume. The moderator mafrix is constructed of 2- by 2- by 63-in.‘stringers of graphite which are loosely pinned to re- straining beams at the bottom of the core. Molybdenum bands at the top and center of the assembly restrain the bowing induced by the radial neutron flux gradient. Flow. passages in the matrix are 0.400- by 1.20-in. rectangular channels machined in the faces of the stringers. A typical érrangement of graphite stringers is shown in Fig. L. Fiow through the core is lam- inaf, but because of the good thermal properties of the graphite and fuel, the graphite temperature at the midpoint is only 4O°F above the fuel | mixed-mean temperature at the center of the core. ) Provision is made for remote removal and replacement of five stringers at the center of the core. They will be examined periodically to determine whether the graphite deteriorates with increasing exposure time. An INOR—Bnpiece is installed in the top dished head to displace fuel and to provide a part of the shielding for equipment above the reactor. Design data for the reactor vessel are detailed in Table k. -11- UNCLASSIFIED ORNL-LR- DWG 52034R FUEL OUTLET . GRAPHITE SAMPLE BLOCK (o] 3 w T 1] Wi o« O o a o T FUEL INLET CORE YOKE GRAPHITE-MODERATOR STRINGER REACTOR VESSEL GRAPHITE-MODERATOR STRINGER - REACTOR GORE CAN FUEL PASSAGE CORE-POSITIONING GRAPHITE BEAMS VESSEL DRAIN LINE CORE GRID SUPPORT Fig. 3. Reactor - Vessel Assembly. -]12- UNCLASSIFIED ORNL-LR-DWG 56874 PLAN VIEW TYPICAL MODERATOR STRINGERS SAMPLE PIECE CROSS ~COMMUN! ~ CATING CHANNELS NOTE: NOT TO SCALE Fig. 4. Typical Graphite Stringer Arrangement. -13- Table 4. Reactor-Vessel Design Data - Structural material Core vessel oD ID Wall thickness Design pressure Design temperature Fuel inlet temperature . Fuel outlet temperature Inlet Annulus ID Annuwlus OD Over-all height of core tank -~ Head thickness Inlet pipe Outlet pipe Core container ID oD Wall thickness Height Graphite éore Diameter Core stringer Number of fuel channels Fuel-channel size: Effective core length Fractional fuel wvolume INOR-8 . :59-1/8 1in. 58 in. 9/16 in. 50 psi 1300°F 1175°F 1225°F Constant-area distributor 56 in. 58 in. 94 in. 1 in. 6-in.-OD tubing, 0.205-in. wall 8-in. schedfhd pipe 68-1/2 in. 55-1/4 in. . 2 x 2x 63 in, 1064 1.2 x 0.4 in. ~ 65 in. 0.225 =1L 2.4 Reactor and Coolant-System Pumps The fuel circulation pump is a sump-type centrifugal pump with a vertical shaft and an overhung impellef. It has a 75-hp motor and is capable of circulating 1450 gpm of salt against a head of ~ 50 ft. Figure 5 is a drawing of the pump, and design data are presented in Table 5. | The pump assembly consists of motor and housing, bearing shaft and impeller assembly, and sump tank. The sump tank is welded into the reac- tor systeh Piping and serves as the expansion tank for the fuel and as a place for fhe separation of gaseous fission products. The bearing housing is flanged to the sump tank so that the rotating parts can be removed and replaced. The motor is loosely coupled to the pump shaft, and the motor housing is flanged to the upper end of the bearing housing to permit separate removal of the motor. The pump is equipped with ball bearings which are lubricated and cooled with oil circulated by an external pumping system. The oil is con- fined to the bearing housing by mechanical shaft seals. Helium is circu- lated into a labyrinth between the lower bearing and the sump tank. Part of the gas passes through the lower seal chamber to remove oil vapors which lesk through the seal. The remainder flows downward along the shaft to prevent radioactive gas from reaching the oil chamber. | Massive metal sections are incorporated in the pump assembly as shielding for the lubricant and the motor. The mobtor is enclosed and sealed to prevent the escape of radioactive gas or fluids which might leak through the pump aésembly under unusual conditions. Water cooling coils are attached to the housing to remove heat generated by the motor. ' A similar pump is provided for the coolant system except:that fewer provisions are required for protection agasinst radiation. The pump is driven by a 125-hp motor and is designed to circulate 850 gpm of salt against a head of 100 ft of fluid. The complete design data for the coolant pump are included in Table 5. |2 & -15- ~ UNCLASSIFIED ORNL~LR~-DWG 56876 MOTOR AND DRIVE # COUPLING HOUSING BUFFER GAS FOR GASKET\ e LTI FES 1 T L L TR 1 18 LUBRICATING OIL INLET R PO 1 : ' C [ ", ; 1" T l! ""‘lf. .] o !i! 1113 | j PURGE GAS INLET , : ' ol - OIL LUBRICATED BEARINGS i ‘ - | T : GAS SEAL ih ZHH LUBRICATING OIL OUTLET SHIELDING PURGE GAS OUTLET q | | | f f GAS FILLED EXPANSION SPACE NORMAL OPERATING LEVEL 1 | | TO HEAT . =\ EXCHANGER DIMENSIONS: . ' _ - 36-in. DIA AT PUMP ! *gdoflm BOWL x 88 in. HIGH “ 11l . ) ,|'E "R [ i FROM REACTOR FWg.5.FueICHcMafloh Pump. -16- Table 5. Design Data for Fuel and Coolant Pumps -*Fuel Pump Coolant Pump ' Flow, gpm 1200 - 1450 850 Head, ft 50 * 5 at %&50 gpm 100 -Motor horsepowef, hp 75 ‘j o 125, Pump speed, rpm 150 1750 Intake o 8-in. schéd-4o pipe 6-in.-OD tubing, - ) o 0.205-in. wall Discharge 6-in.-0D tubing, 5-in.-0D tubing, 0.205-in. wall 0.165-in. wall Sump-tank volume,* £ 5.2 ' 4.8 Normal salt volume, £t° 2.61 . 2.4 Expansion volume, ** £1° 1.83 o | 1.6 Sump tank | S Outside diameter, in. 36 o 36 Height, in. 15 b Over-all assembly height, ft 7-3/4 7-1/3 Structural material INOR-8 _ ~ INOR-8 *Not including volute. **Normal to maximum. 2.5 Primary Heat Exchanger The primary heat exchanger_(Fig. 6) contains 165 tubes (1/2-in. OD, 0.045-1in. well, 13 ft long) and is designed to transfer 10 Mw from fuel salt (in the shell) to coolant salt (in the tubes). The design is a con- - ventional cross-baffled shell-and-tube configuration with emphasis placed on ruggedness and reliasbility rather than high heat-transfer performance. Design data are listed in Table 6. | -Space limitations in the reactor cell require a short unit. The U- tube configuration mekes this possible without greatly reducing the effi- ciency of heat transfer as compared to a straight counter-flow unit and eliminates a thermal expansion problem. The tubes are welded and back- brazed to the tube sheet in order to greatly reduce the probability of leakage between the fuel and coolant. SSSSSSSSSSSS RRRRRRRRRRRRRRRR » Yo-in-0OD HEAT EXCHANGER TUBE THERMAL-BARRIER PLATE (25 % CUT) o COOLANT INLET ) 2, X 5%) NN ) 83 N , : Lo 164-in. OD x O0.2-in. WALL x B-ft LONG " ‘|“ ., ) .“‘ . " " i ",'J'JL"‘; gy 4 el \ 'fv i AT X Al ' , * COOLANT-STREAM . ; .COOLANT OQUTLET .| SEPARATING BAFFLE fl FUEL OQUTLET Fig. 6. Primary Heat Exchanger for MSRE. CROSS BAFFLES ‘LT‘ =18~ Teble 6. Design Date for Primary Heat Exchanger Structural material Heat load Shell-side fluid "Tube-side fluid Layout | " Baffle pitch Tube pitch Active heat-transfer length of shell ‘ Over-all length Nozzles Shell side ‘Tube side Shell diameter Shell thickness Tube-sheet thickness Number of tubes Tube size Tube length Heat~transfer surface ares Fuel holdup Design temperature Shell side Tube side - Design pressure Shell side Tube side Terminal temperatures at ‘design point Fuel salt Coolant salt Effective log mean temperature difference INOR-8 10 Mw Fuel salt Coolant salt -25% cut, cross-baffled and U-tubes 12 in. 0.775 in., triasngular 6 £t 8 ft 6-in.-0D tubing, 0.205-in. wall 5-in.-0D tubing, 0.165-in. wall 16 in. ID 1/5 in. 1-1/2 in. 165 1/2-in. OD, 0.045-in. wall 13 ft ‘259 £t ~ 5.5 £t° 2 1300°F 1300°F 50 psi 75 psi Inlet 1225°F; outlet 1175°F Inlet 1025°F; outlet 1100°F 133°F. ..]_9_ 2.6 Salt-to-Air Radistor The thermal energy of the reactor is rejected to the atmosphere by means of & salt-to-air radiator. The radiator contains 120 tubes . (3/4-in. OD, 0.072-in. wall, 33 ft long) and is assembled as shown in Fig. 7. Design data are listed in Table 7. o | .Several features were inqbrporated in the design as protectioh. agelnst freezing of coolant salt in the radiator: 1. Tubes are of large diemeter. | 2. The heat-removal rate per unit area is kept low by using tubes without fins so that most of the temperéture drop is in the air film. 3. The minimum salt temperature is kept 75°F above the freezing point. | ., The headering system is designed to assure even flow distri- bution between the tubes. o 5. 1In the event of flow stoppage, doors on the-radiator,housing close within 30 sec, and heaters are turned on to prevent the salt from freezing. , ‘ The layout of the tube matrix will allow movement of the tubes with minimum restraint during thermal expansion. The tubes are pitched to promote drainage.‘ ‘ The radistor is supported and retained in a structural steel frame which 1s completely enclosed and insulated. Reflective. shields protect structural members from excessive temperatures. The frame also provides ‘guides for the vertically sliding doors which are closed to thermally ~ isolate the radiator, should any situation develop which could cause salt freézing in the radiator tubes. Two doofs are employed, one each upstream and downstream; and they . are raised and lowered at & speed of 7 ft/min during normal operation by a gear-reduced motor driving an overheed line shaft. The doors are sus- pended from roller chains which run over sprockets mounted on the line shaft. The enclosure is capable of sustaining full blower pressure in any position, and may be used to vernier air flow across the rgdiatOr'as a control on the reactor load. (s UNCLASSIFIED ORNL-LR-DWG 52037 R4 5-in. SECONDARY-SALT INLET DIMENSIONS: ‘9-ft WIDE 6-ft HIGH 45-in. DEEP \3/4-in.—OD X 0.072-in-WALL TUBING; 10 ROWS, 12 TUBES PER ROW 8-in. HEADER 5-in. SECONDARY - SALT OUTLET 2%-in. TUBE MANIFOLD AIR FLOW Fig. 7. Salt-to Air Radiator. -0z- ~21- Fmergency closure is effected by de-energizing a magnetic clutch between the motor and the line shaft or, alternatively, by de-energlzing magnets used to attach the roller chain to the door. Shock—absorbingm_ means are provided. Emergency closure 1s not contingent on door posi- tion. Table 7. Design Data for Salt-to-Air Radiator Structural material | INOR-8 Duty : 10 Mw Temperéture differentials | _ | | _ | ‘Salt Inlet 1100°F; outlet 1025°F Mr | . Inlet 100°F; outlet 300°F Air. flow | - 164,000 cfm at 15 in. B,0 Salt flow | 830 gpm at ‘avg. temperature Effective mean AT . 920°F Over-all coefficient of heat transfer 53 Btu/ft?-hr-éF Heat-transfer surface area | | 685 £2 Design temberature | 1300°F Deéign bressure 15 pSi Tube diemeter | ~ 0.750 in. Well thickmess | © 0.072 1z. Tube‘matfix | 12 tubes per row; 10 rows deep Tube spacing o | 1-1/2 in., triasngular Suvheaders | | 2-1/2 in., IPS Main headers 8 in., IPS Air-side AP , 11.6 in. H,0 Selt-side AP : - 6.5 psi 2.7 Drain and Storage Tanks . Five tanks are provided for safe storage of salt mixturés when7 they are not in use in the reactor and coolant systems. They are: two fuel drain tenks, a flush-salt tank, a fuel- and flush-salt storage tank, and a coolant drain tank. -22a 2.7.1 Fuel Drain Tanks The fuel drein tanks have the important function of subcritical storage of the fuel and must have means for femoving decay heat and for meintaining selts molten when the internal heat generation rate is low. Two tanks of the design shown in Fig. 8 are provided, each of 67 £t capacity. Each tank can hold an entire fuel charge; so one is for normal usé and the othér is a épare. The low moderating power of the salt makes criticality impossible even with nearly double the planned U235 loading.- After long-term operation at 10 Mw, sudden draining of the fuel requires that it be cooled at a rate of 100 kw to prevent excessive fuel temperatures. Evaporative cooling was chosen over gas or other means on the basié_of simplicity and independence of utilities. Heat is removed by 4O bayonet cooling tubes (Fig. 9) inserted in thimbles in the tank. . Water is fed through fihe center tube, and steam is generated in the surrounding annulué. Heat is transferred from the thimble to the cooling- tube through a 5as-filled annulus by radiation and conduction. Normally the steam is condensed by ajwatér?cooled condenser, but it can be.ex- _ Hausted to the stack in the event of failure of the coolant supply. A 300-gal feed-water reserve can provide cooling for 6 hr. The drain tanks are insulated and are provided with electrical heaters. They have dip—tube fill and drain lines ard gas connections for maintaining a helium blanket for venting and for pressurizing to transfer the salt. Design data for the drain tanks are presented in Table 8. - 2 07 ° 2 ° FluSh"sa:.Lt T&Dk A salt of composition similar to the fuel salt but without thorium or uranium is used to flush the reactor system to remove possible chemical contaminants and to aid in preheating before the fuel is charged. During shutdown it can be used to remove residual fuel after the primary system” | has been drained but before the piping is opened for maintenance. The flush salt is stored in a tank (see Table 8) similar to the fuel drain tank, but w1thout coollng tdbes. P-4 @ iR 8 O & o > 5 =2 - Y €T g5 Z = 5 ® £ s O = 4 <+ T m i} - o= | O = | 0 oW T wo M- _._N._ = .NI (] ™~ _H.. o z W = Ay o =4 — (@) oo X w [t w a < - <1 a — o | wQ o = = - =2 q o wi = = 3 ) L T [7)] -~ c o T L he! ' c o O q f= _ O | LCAC,~IC,;“Oa OO, o =N " o R s A 2oy /./l - = .zflflfiwrww,%mp.wmww,#554444;.444444%%%4%%%%&%%444,414415441;55;.4/.1.41.4431.44%44%%/.4%##/.54%4..flml.%awfi..//., N /xf/////”. m ol " F A N R A, At ///fi///////////////l\ /////// ”fi . ” - QP. ,@//////w /fi o 4 D 9 < l,,\ . -« LS ; - & N Y R R RN R e nuw N o — Q _ : ,\ L —Plv w \/ . J o G-r G pM_ o J 2= zc @ _y 1w o oI Ogq no < OO - w o a2 e’ ©a o Qa - @ z8 2 == L 03 L& 0% o o, ga Wo = e - w > 1 oo Y p=d = > Z =1 Jac w = m' ad B wd a3 Y © o mu._ 20 0 o Iy Jw @ W [ > W O Rl qZ o L= W > @ L ] ) ok~ UNCLASSIFIED ORNL—-LR—-DWG 52038R¢ CQOLING-WATER INLET\\ /STEAM OUTLET N N N \ N N N \ \ : v N TANK HEAD ¥ y o INTERMEDIATE WELDING NIPPLE OUTER CONTAINMENT TUBE AIR SPACE 1 | STEAM RETURN TUBE \) ¢ i|fllit! w— SALT CONTAINING TUBE SPACER FINS flSTEAM ANNULUS | SPACER FINS ;. IN AIR ANNULUS WATER FEED TUBE — J{\¥'F AlIR SPACE Fig. 9. Cooling Thimble for Primary-Salt Drain Tanks. s Table 8. Désign Data for Fuel Drein Tank, Coolent Drain Tank, and Flush-Salt Tank Fuel drain tank Height Diameter Volume Total Fuel (normal) Gas blanket (normel) Wall thickness Vessel Dished head Design temperature Design pressure Coolifig method Cooling rate Coolant thimbles Number Diameter Coolant drain tank Height - Diameter Volume Total. Coolant salt Gas blanket Wall thickness Vessel | Dished head. Design temperature Design pressure Cooling method INOR-8 81-1/2 in. (without coolant headers) 48 in. | 67.6 tt° RN R % ~8 £t 1/2 in. 3/4 in. 1300°F 50 psi Boiling water in double-walled thimbles 100 kw 40 2 in. OD INOR-8 76 in. 36 in. 39.5 £t° ~3l4 £t ~6 17 3/8 in. . 5/8 infb 1300°F 50 psi None 26 - Table 8. (Continued) Flush-salt tank _ , INOR-8 Height 76-1/4 in. Diameter _ 48 in. Volume N | Total ~ 67.1 £t° Flush salt ~59 £t Gas blanket ~8 Pt Wall thickness Vessel 1/2 in. Dished head - 3/4 in. Design temperature 1300°F Design pressure 50 psi- ' Cooling method = None 2.7.3 Coolant Drain Tank A tank (see Table 8) of 40-£t° capacity and similar to the drain tanks but without cooling tubes is provided for the coolant salt. 2.7T.4 Storage Tank Occasionally it will be necessary to remove the fuel charge from the reactor‘for reprocessing and to add a fresh charge. A separate storage tank of 67--ft5 capacity is provided for storing used fuel while it is being removed in small batches to a reprocessing plant and for accumula- ting small batches of new fuel until it can be charged to onme of the drain tanks. ) | The storage tank is like the fuel drain tanks except that it has no cooling tubes and therefore cannot accept salt from the reactor until | the aftefheat has decayed for about two weeks. The tank is equipped with lines for transferring salt to the fuel drain and flush tanks and to equipment provided for loading and unloaaing carriers. 2.8 Cover-Gas System Because the fuel salt is sensitive to oxygen, it must be protected by an oxygen- and moisture-free cover gas at all times. The principal . -‘»27_ functions of the cover-gas system. are to supply an inert gas for blan- keting. the salt and for the pressure transfer of salt between components, to pnovide a means‘for disposing of radioactive gas, and to provide a higher gas pressure in the coolant system than in the fuel system. A simplified flow diagram of the system is presented in Fig. 10. The cover gas is helium supplied in cylinders at 2400 psig. It is purified by passage through filters, dryers, and oxygen traps (possibly titanium, zirconium, or uranium chips at high temperature). Purified gas is then sent to two distribution systems. The total flow 1s about 10,500 liters/day (STP), and it is distributed at about 50 psig. The largest flow of gas is directed to treated-helium storage tanks and then to the primary distribution system, which supplies nurge for the fuel circulation pump, the freeze-flange“buffer zones, &and the fuel drain tanks, where it is in direct contact with the fuel salt. Gas which -passes through the fuel pump is circulated through a series of pipes where it is held and cooled for at least 2 hr to dissipate heat from the decay of short-lived fission products. Then it passes through a .charcoal bed where xenon and Krypton respectively are retained for at least 72 and 8'days, and through.a filter and blower to the off-gas stack. There it is mixed with a flow of 20,000 cfm of air which provides dilution of 1: 15 000. The charcoal bed is a series of nipes packed with activated carbon It and a spare bed are mounted vertically in a sealed, water- filled secondary container; either or both beds may be used. Fuel-salt transfers require more rapid venting of gas, but the heat load is low. A third charcoal bed is provided for venting those gases before they are sent to the stack. Fig. 11 is the off-gas disposa.l flow- sheet. | ' Although not included in the initial installation, prov131on will be made for recirculation of gases from the outlet of the carbon bed through a purification system and into treated-gas storage tanks. : The'cover—gas distribution for the coolant system (also shown on Fig. 10) supplies a smell flow of helium to the coolant system, the sampler-enricher system, and to the fuelwpump motor. That equipment mustvbe'supplied with nonradiocactive gas and will not normally be contam- inated by gaseous fission,products. Gas from the eoolant;system is 260 psig FILTER DRYER FRESH HELIUM STORAGE TANKS 20-day SUPPLY 2400psig MAXIMUM PRESSURE OXYGEN % < REMOVAL . FRESH HELIUM SURGE TANKS 4313, 260 psig OPERATING PRESSURE. Fig. 10. Cover-Gas Supply IEEE N/ RRIEERIRIARE UNCLASSIFIED ORNL-LR-DWG 56872 FUEL-PUMP SWEEP-GAS UPPER INLET FUEL-PUMP SWEEP-GAS LOWER INLET FUEL-SALT STORAGE (2) FLUSH-SALT STORAGE INTERMEDIATE TRANSFER TANK SUPPLY HEADER FOR RADIOACTIVE HELIUM (50 psig) ' ' SUPPLY HEADER FOR NON-RADIOACTIVE HELIUM (50 psig) COOLANT-PUMP SWEEP GAS COOLANT-PUMP FLANGE BUFFER | COOLANT-PUMP MAIN FLANGE GASKET COOLANT-PUMP MOTOR BUFFER COOLANT-PUMP MOTOR FLANGE GASKET FUEL-PUMP FLANGE BUFFER FUEL-PUMP MAIN FLANGE GASKET FUEL-PUMP MOTOR BUFFER FUEL-PUMP MOTOR FLANGE GASKET FUEL-PUMP LIQUID-LEVEL BUBBLERS FUEL-PUMP LUBE OIL SYSTEM FREEZE-FLANGE BUFFER SAMPLER-ENRICHER | . COOLANT-SYSTEM SALT STORAGE FUEL TRANSFER CARBOY - 93- cw | | WA ™ l t HELIUM | SUPPLY 0 N VOLUME CHARCOAL BEDS (2) —_— l o i o HOLDUP 4.2 liters/min — 72 DAYS Xe HOLDUP L1 ) ' : FUEL i iq MAX — °F PUMP : 1Q psig MAX —-1200 ‘. | H— Jd - REACTOR CELL UNCLASSIFIED ORNL-LR—-DWG 56877 ACTIVITY MONITOR 5cfm MAX PRESSURE 2-in. Hg VACUUM EQUALIZING LINE —mm 7503 BUILDING VENTILATION HDR ¢ < O psig—85°F Y 27,400 cfm ) FLUSH SALT STORAGE TANK VENT }%D FUEL STORAGE TANK VENTS X ACTIVITY ___? ' : MONITOR ‘ 46'__ ACTIVITY : MONITOR | — 1 CHARCOAL BED 1 cfm - INTERMITTENT FLOW NEGLIGIBLE HEAT LOAD ~——t ) VENT - HEADER\ “SNFUEL . SSFUEL TRANSPORT INTERMEDIATE CARBOY VENT ~—SAMPLER ENRICHER VENT TRANSFER TANK VENT | §‘< Fig.11. Fuel-System Off-Gas Disposal. FILTER STACK - 68- -30- vented directly through filters to the off-gas stack as indicated on Fig. 12, Monitors will stop the flow to the stack on indication of high activity. 2.9 TFreeze Valves The-molten salt in both the.fuel,and coolant circuits will be sealed off from the respective drain tanks by meané of freeze valves in the drain lines. These valves (Fig. 13) are simply short, flattened sections of pipe which are cooled to freeze the salt in that section. Calrod heaters surround each valve so that the salt can be thawed quickly, when necessary,.to drain the system. The”salt can also be thawed slowly without the heaters by stopping the cooling. The valves are mounted with traps on both sides so that salt cannct be blown out of the line. The £111 and discharge lines fof the fuel and the flush-salt tanks are manifolded together and are connected to an outer stofage tank. There is & freéze valve in each line. 2.10 Sampler Mechanism Small quentities of fuel can be added or removed by means of the sampler mechanism which is connected to the fuel-pump bowl. A special container located outside the containmentlvessel above the pump elevation " encloses the working parts. A cable assembly with a reel is used to lower a small bucket into the salt pool in the pump bowl. Fuel samples can be removed for analysis, and new fuel can be added in small (less than 120 g) quantities to compensate for burnup. Since this operation purposely breaches the secondary and primary containers, it is extremely important that substitute protection be provided. This is accomplished by building a special solder-seal dis- connect and two isolated compartments into the sampler, with protection against both being open at the same time. Furthermore, the sampler enclosure 1s ventilated to the charcoal-filter system. With these mechanical devices and with special attention being given to operating procedures, it is belleved that the sampling operation can be handled with complete safety. HELIUM SUPPLY ACTIVITY MONITOR J 1000 liters/day —A\/ W\ {0 psig MAX r-- I UNCLASSIFIED ORNL-LR-DWG 56873 111 I COOLANT PUMP — XD B £ COOLANT SYSTEM SALT STORAGE TANK Fig.12 Coolant-System Off-Gas Disposol‘. F|LTER FAN STACK -Tg- AlR BLAST COOLING JETS ( RADIANT HEATING COIL Fig.13. Radiant Heat Freeze Valve. UNCLASSIFIED ORNL—-LR-DWG 56875 -33- 2.11 DNonnuclear Heating External heating of salt-bearing components of the reactor system is necessary: | 1. to prevent freezing of the salts, 5. +to raise the reactor temperature to & subcritical value for'experimental convenience, "3, to heat the salts for reactor startup. Replaceable'eleCtric heaters were chosen as the safest, most reliable means of supplying the large (~300 kw) high-temperature requirements. Diesel-driven generators and two separate TVA substations make a complete failure of the heating system extremely unlikely. In the 1000 to 1509°F range nearly all the heat is transferred by radiation. This makes it | unnecessary that the heating elements be in contact with the vessel walls and results in a safer system from the standpoint of overheating and . arcing damage. | ‘ | Different kinds of heaters are applied to different parts of the reactor. The core vessel, drain tanks, flush tank, and storage tank are equipped fiith hairpin-shaped resistance heaters which fit into wells sur- rounding the vessels. The piping and the heat exchanger are covered with clamshell-type resistance heater assemblles, which are designed for easy removal. Reflective insulation is incorporated in the design. The coolant system and the radiator are heated by Calrod-type radiant elements with ordinary insulation, because these areas can be maintained directly. -34- 3. INSTRUMENTATION AND CONTROLS .'3.1 General The MSRE is & safe reactor because: 1. It has a good negative temperature coefficient. 2, Only a small amount of reactivity is fequifed to compensate for xenon‘poisofiing, the negétive power coefficient, and " burnup of fuel between fuel additions. | 3. The stability of the fuel increaées with increasing temperature. 4. There are no known mesns by which the amount of uranium in the core can be increased rapidlye- | Normally the reactor will operate at 1175 to 1225°F. However, it can be cooled to 900°F and heated to 1300°F with almost complete freedom. The fuel salt beginszto freeze at 830°F, and freezing should be complete at T90°F. The serious consequence of ffeezing is that salt expands on melting so that remelting must be done with extreme care or small pipes will rfipture and salt will_be spilled into the containment cells. In desligning the reactor, the maximum stress allowed 1s two-thirds G6f that which will produce a minimm creep rate of 1% in 105 hr at 1300°F. The same sustained stress will produce 1% creep strain in 2000 t6 4000 hr and rupture in 5000 to'i0,000 hr at_1500°F; At 1700°F the time to 1% strain is 40 to 100 hr and the time to rupture is 200 to 600 hr. The equipment and piping will be designed so that the stresses, neglecting relaxation, will not change much with dhanging temperature in an isothermal system. Heating to 1500°F for a thousand hours or to 1700°F for a few hours should have little effect on the life of the equipment. | | Large temperature gradients are the main cause of excessive streés on heating and cooling the reactor. For this reason normal heating and .cooling will be done so that température'differences in the system are less than sbout 100°F. To ensure this, the normal rates of change of temperature will be kept below 1°F per minute for total changes greater than sbout 100°F. If a nucleaf excursion causes the reactor core tem- perature to rise 300 to 500°F very rapidly and to remain at the higher ~35 e temperature, the thermal stresses induced in the piping and equipment as they heat somewhat less rapidly to about the same temperature will -cauge some ylelding and distortion.. The fuel pump probably will have to be replaced. Several such severe;thermai shocks would be required to breach thevsystem and permit the fuel to leak. 3.1.1 ControereQuirements The MSRE has temperature coefficients asfifoIlows: 1. Tuel COEPFiCient .....ovieeseseses =3 x 1070 (2k/k)/°F 2. Graphite moderator coefficient ... -6 x 10> (2x/k)/°F The fuel coefficient prevails during very rapid trensients. The iso- thermal temperature coefficient is the total or -9 x 10~ (2&k/k)/°F. Excess reactivity over that required for the reactor to be criti-: cal while cleen and noncirculating at the design operating temperature (1200°F)'must be provided in order to maintain the temperature wvhile the reactor is operating steadily at full power. Thiswreactivitylmust be sufficient'to compensate‘for xenon poisoning; loss of delayed neutrons in the circulating system, end some burnup of fuel. (a) Xenon Poisoning. Xenon will be removed continuously from the 4% of the fuel flow which circulates through the pump bowl. Some xenon will c1rculate with the fuel, and this will permit apprec1able , amounts of gas to diffuse into the voids in the graphite. Also there will be some permeation of fuel 1nto the graphite. The exact amount . has not been determined for the MSRE graphite, but enough work has been done with 31milar graphites (see Appendix F) to form a good basis for assuming that only 0.5% of the graphite volume will be occupied by fuel. ZXenon produced in this fuel will contribute to the poisoning., The steady-state xenon poisoning at 10 Mw is estimated to be 1. 3% in Mk/k. The peak poison level after 10-Mw.operation is 4 to 5% There is no need to compensate for the peak poisoning because the stripping operatlon continues to remove xenon from the fuel when the reactor pover is reduced. (b) Power Coefficient. Heating in the graphite as the power is. raised causes the reactor to have a power coefficient of reactivity estimated to be =0.02% (Ak/k)/Mw or -0.2% total at 10 Mw. -36- (¢) Delayed Neutrons. Decay of precursors in the piping and heat exéhangersiresults in loss of delayed neutrons from the reactor core. The reactivity changes by -0.1T% in Ak/k when fuel circulation is started and increases again when circulation 1s stopped. (@) Burn p. The removal of fuel and increase in fission-product poisoning by burnup changes the reactivity by about -0.002% in Ak/k per Mw-day. Fuel will be added to the circulating gystem at intervals of 10 days or less. The total change in reactivity between additions will be -0.2% in Ak/k or less. | The total excess reactivity over the émount required for steady operation at 10 Mw and 1200°F is the sum of the sbove or 1.9% in Ak/k. In the absénce of nuclear control devices, the critical temperature would eventually rise from 1200 to about 1400°F in the event of pump stoppage. Also, the system would have to be heated to 1400°F to keep the reactor subcritical during the recharging of fuel and system start- up operations following a shutdown for maintenance. This temperature is sufficiently above the design temperature that-nuclear control devices are necessary during startup and to eliminate large temperature changes that would accompany power changes during normsl operation. The control should have a worth of sbout 4.6% in Ak/k so that the reactor can be held subcritical down to 900°F. This will make it unnecessary to heat the system above "1000°F during normal loading perations and will provide some margin for varying the temperature while at full power. Fuel will not be kept in the reactor at lower temperature or while the containment cell is open; so no additional shutdown margin is necessany ' Although the reactor requires a nuclear control system, the excess reactivity is so small that its insertion at any possible rate will not cause the fuel to escape the piping. Therefore the reactor does not require and is not pioVided with an infallible, fast-acting safety system; ‘ : | | In the present reactor design, spaces are provided in the graphite assembly for four l-in.-diam control thimbles near the center of the core. Calculations indicate that as much as twice the reQuired control ..37;_ can be incorporated in.these thimbles if desired. Botfi;liquid and solid control devices are being studied, and the most satisfactory will be | adopted for this_reactor. . The control devices are hot.required or desired to operate rapldly. _However,-it,will be . important to\havé‘accurate ) knowledge of the position of the poison at all times to provide a con- .tinu6us,indiéation of changes in-reactivitj, Design details and. results of full-scale mockup tests of the comtrol system will be included in the Final Hazards Reporfi. v | 3.1.2 Other Control Features As discussed above, the primary function of the nuclear control system is to maintain the critical temperature of the reactor below the system design temperature of 1300°F. Poison will be inserted if the circulating pumps stop (see Sec. 7.1.3), if the mean temperature rises too fast, if the power should rise more then 50% above normsl, or if the escape of rédioactivity is detected. Simulafor studies have indicated - that the self:comtrol features of the MSRE will not hold the power steady at low levels. Automatic operation of the control system will be -provided to hold ‘the power constant at any desired level up to 10 Mw.. Additional protection against the reactor temperature exceéaing 1300°F for long periods is provided by the drain system. The freeze valve in the reactor drain line can be thawed and the fuel can be. drained in sbout 15 min. | 3.1.3 Nomnuclear Controls Although not associéted with nuclear safety, it is important to keep the fuel and coolant salts from'freezing in the reactor piping. As has been described, electric heaters with an extra reliable power supply are provided on all circulating lines and on all vessels con- taining salt. The heaters are always energized, except those on the coolant salt-to-air radiator. - _ A special protective circuit prevents the coolant salt from freezing in the radiator. Three actions are involved: shutting off the air blowers, applying full.power to the radiator heaters, and closing the\radiator doors to give thermal isolation. If the salt -38- temperature at.the radiator exit drops from the design point of 1025°F to 975°F, the blowers are turned off and full.power is applied to the heaters. A further drop of 50°F (to 925°F) causes the radiator doors to close. This is about 75°F above the temperature, 850°F, at which solids initially appear. : . The reactor and coolant drain systems pfovide additional protection against 'salt freeziné in the reactor or coblant systems piping in that the saits can be drained if the temperatures approach the freezing points. -39~ 3.2 Instruments 3.2.1 Reactor-Power Measurement ‘The reactor power is determined from the flow rate of coolant salt and its temperature difference across the fuel heat exchanger. The flow 18 measured with a Venturi meter and differential-préssUre cell in the coolant circuit " This equipment is located outside the reactor and can be easily maintained during reactor shutdowns. ' Reactor power will alsoc be indicated by the neutron level of the reactor, after the neutron indicator is calibrated by heat balances. 3.2.2 Fuel Inventory One of the most important measurements in a mobile fuel system is the fuel inventory measurement. Two pleces of infdrmation afe required: (1) the total quantity of salt in the system and (2) the concentration of U235 in the salt. The quantity of salt is determined from the mass of fuel in the circulating system as indicated by the liquid level in the pump bowl and by the mean temperature of the loop. To this must be added any material remaining in the drain tanks, which are we1ghed by pneumatlc load cells. The concentration of uranium in the salt must be determined by chemical analysis. Samples can be removed at any time from the sampler- enricher device described in Sec. 2.10. - 3.2.3 DNuclear Instruments The nucieaf‘instrumentation is comprised of two count-rate circults with U235ucoated chambers as the detectors and four high-level circuits equipped with ion chambers. All six chambers are inserted in access ‘thimbles which terminate insidetthe thermal shield around the reactor vessel. The arrangement is shown in Fig. 14. 3.2.4 Radiation Monitoring Rediation detectors are requiredv(l) to protect against the escape of radioactive liquids through the many service lines which penetrate the containment shell, (2) to measure the background levels of radiation {— SAFETY CHAMBER —={ )\ \ OUTER STEEL SHELL CONTAINMENT VESSEL {—-SAFETY CHAMBER /{‘I—COMF’ENSATED CHAMBER 4 REACTOR SHIELDING Sn—Be SOURCE TUBE Fig. 14. MSRE Neutron Chamber Arrangement, UNCLASSIFIED ORNL-LR-DWG 56878 2—-FISSION CHAMBERS { ~ COMPENSATED CHAMBER 1—SAFETY CHAMBER =41 through shielding adjacent to occupied areas, and (3) to monitor con- - tinuous samples of the atmosphere within the operating areas.and the stack discharge° . o | - | | The detectors which perform functions (l) take automatic safety 'action when preset safe limits are exceeded Those in categories (2) and (3) only alarm. All indicate remotely and record so that the . operator can immediately investigate the source of indicated radiation. 3.2.5 Pressure Measurements Pressure measurements at various points in the cover-gas and off- gas systems are required during operation and during salt transfers. Seal-welded pressure elements are used in the fuel and coolant systems. Other pressures are measured with conventional pressure devices located outside the container. They include the lubricating-oil systems, the water cooling systems, and the pressure differential between the con- tainment cell and the atmosphere. 3.2.6 Flow Measurements Tn addition to the coolant salt, flow measurements are required on the off-gas and lubricating-oil gas streams at the fuel pump, and on the sump lubricating-oil and coolsmt-water streams. Seal-welded dif- ferential-pressure cells are installed to measure the pressure drops across these orifices and capillaries. 3.2.7 Temperature Measurements Temperature'determinations throughout the reactor plant are made by Chromel-Alumel thermocouples sheathed in Tnconel. The data will be used for control, for operational information, or for initiating safety . actions. Temperature points of interest in the reactor system include all salt containers and piping, reactor-vessel and pump-bowl areas that might have local heating, freeze flanges and freeze valves that must be kept cool, and many locations in the off-gas system. Also, lubricating- oil ard cooling-water temperatures and the ambient temperature:. of the cell atmosphere must be monitored. - Over 500 thermocouples will be used to collect the desired information. | 3.2.8 LiquidJLevel Measurements The level of salt in the primary pump is measured with a differ- ential transformer actuated by a float in the pump 5ow1. The level of salt in the drain tanks is determined by weigh cells vhich are designed into the tank support system. The levels of other liquids, such ‘as the 1ubrica£ing oll and the water in varlous cooling systems, are determined by conventional floats or static pressure-head measure - ments because these devices are located outslide the container. _u3_. L. REACTOR PHYSICS DATA A l1list of reactor physics data for the currently proposed MSRE core design (in the clean, hot condition) is given in Table 9. The critical mass, system inventory, temperature coefficient, neutron fluxes, and power-density estimates were obtained by using the IBM-TOk one-dimensional maltigroup diffusion-theory program GNU3 end the 34- group cross-section lfl.'brt—:i.rylL prepared for use in the phermal-bfeeder- reactor evaluation program.5 The poison-tube worth was calculated withAtheitwb4group, fiwo-dimensional diffusioh-theory program PDQ,6 using'an effective extrapoiation distance at.fihefibeundaries of the con- trol regionsiend two-group constants preduced by the GNU calculations elsevhere. Table 9. Reactor Data for Clean Hot Condition, All Tubes Empty Fuel volume fraction in core 'l 1 0.225 ' Core critical mass, kg U= o 15.6 Circulating-system inventory, kg U235 | My Temperature coefficient of reactivity, (8k/k)/°F -9 x 107 Fuel - ' : - -3 x 10~ Graphite | -6 x 10~ Mean neutron lifetime, sec | 3 x 107" Effective delayed—neutron fraction - 0.0048 Total poison-tube worth (four tubes), % Bk/k - ke Ffaction'of power generated in core ‘ 0.962 Per cent thermal fissions - . 86.6 Total power, Mw : ‘ 10 Mean core power density, w/em® . , 3.86 Peak core power density, w/cm® | 10.5 Mean salt pofier density, w/em® o 17.2. Peak salt power density, w/cm® 46.7 Peak thermal flux, neutrons/cm3-sec 8.1 x 1013 Mean thermal flux, neutrons/cm®-sec | 2.6 x 1013 (8c/k)/ (/M) | - 0.23 e 5. THE REACTOR COMPLEMENT; 5:1 Building The Molten'SaltiReactor Experiment will be conducted in ORNL Building 7505,‘which originally.was constructed for the-Aircraflt Reactor Experiment (AREj and iater wa.s extefisively modified to'house the Aircraft Reactor Test (ART). The additional revisiéns required for the MSRE are described here. | | A plan view of the building is shown’in.Fig- 15. The reactor contaimment cell, the drain-tank pit, and the coolant;eqfiipment pit are located in the south end of the structu:ej pits’for fuel storage and contaminated-equipment handling are in the nofth end. This por- tion (the high bay) of the building is isolated from the other por- tions by a ventilation system and is designated a "contamination zone'" because it is likely to become slightly contaminatéd during reactor maintenance operations. On the west Wall.is‘éOnstructed a shielded, remote-maintenance control room from which hoists and manipulators may be controlled. Access-to’the high bay is at the - northeast corner through.a change»room where workers can chahge to "contamination" clothing before entering and can be surveyed for contamlnatlon before leaving. East of the high bay on the ground level are the reactor control room and the office area. On the basement level are installed auxiliary reactor instrumentation and building service equipment. At the south end of the building is the fan room, where the two large cooling fans are mounted to discharge over the coolant-salt radiator and to the stack outside. The building has two extensions on‘the west side, where an emergency diesel power statlon and electrical switch gear are housed. 5.2 Containment The cofitainment philosophy which has been applied to the MSRE - requires that a mifiimum of two barriers be provided as protection against the escape of radiOdctivity. The first barrier is the reactor UNCLASSIFIED ORNL—LR—DWG 56879 OPERATING - : /CREW CHIEF W INSTALLATION [{NSTALLATION / | OPERATING . CRAFT ENGINEER ENGINEER CREW : : HEALTH o FOREMAN _ ENGINEER -, INSTRUMENTATION SHOP | e / : LOCKERS ' REACTOR CONTROL ROOM S CHANGE b . _ - i ROOM. | | ZONE 11 T 5 A, ’ ] A B | o REACTOR | | ) : ] I ’ —i [l cCONTAINMENT ) - \ CELL ) : N — R0 | - —— C D DRAIN | . —| | RADIATOR = CONTAMINATED ! /N, | . TANK PIT LI PIT 3 ' 1 LA | I / I_‘ . . MAINTENANCE ~-- ‘ (MAINTENANCE | VENT SHOP . ri| CONTROL HOUSE - |__room _ , {EL.B62 ' . Lot N #5—49_» ...... FLOOR O GO SE ANNULUS FILLED WITH WATER AND AGGREGATE /—\V“"\ -Fig. 16. Shielding ond Sealing Membrdne for Top of Cell. T~~~ SHIELDING RING - 24-f1-DIA REACTOR CONTAINMENT CELL. 'Lfl' UNCLASSIFIED ORNL-LR-DOWG 56884 I\ : ' . i CONTROLLED VENTILATION ZONE Ofw STACK CHANGE HOUSE MAINTENANCE CONTAINMENT CELL SHOP STORAGE REACTOR (VESSEL DECONTAMINATION AREA FUEL STORAGE AND TRANSFER AREA DRAIN TANK CONTAINMENT PIT (WATER \_THERMAL SHIELD ANNULUS Fig. 17. North-South Sectional Elevation—-Bldg. 7503 -49- Since the drain-tank cell is connected to the reactor cell, both cells will be pressure- and leak—testédjat the same time. When~con;? struction is completed but before equipment is installed, a hydrostatic pressure of 45 psig will be applied. Later, after the reactor is. ready for operation, pneumatic pressure will be used to determine -the leak - rate and to lodéte the leaks. This test will be repeated-each time. the cells are opened and closed for maintenance to be certain that - the 1%/24 hr allowable leak rate is not exceedegd. Furthermore,_during- operation the.cells will bé kept at a negative pressure of'2upsigi(13f psia);; and the leak rate will be indicated continuously by. the rate’ of pressure rise. 5.2.2 Other Cells As mefitioned previously, pits for fuel storage, loading and un- loading, used-equipment storage, and decontamination are also located in the high—bay area. The storage tank is provided with a secondary container in the‘form'of another tank which surrounds the storage tank and is designed for complete containment. The amount of activity in the other cellsAis low by comparison to that in the reactor ceil, and it is considered sufficient to protect against the escape of activity by providing a strong draft through openings and by the use of absolute filtérs to remove particuléte activity before the air is diséharged to the stack. Approximately 8000 cfm of air is-available for the venti- lation of these cells. 5.2.3 Penetrations ‘Each of the many service lines which penetrate the walls of the secohdary containers is equipped with a seal or a closing device to prevent the escape of radicactive fluids. | All electrical and thermocouple wires are encased in metal tubing and insulated with a dense packing of magnesiufi oxide. Leakage tests. on this type of cable have indicated leak-tightnesé at pressures as high as 1500 psig. _ | Water, oil, and air lines are designed with_solenoid, pnéfimatic, or spring-loaded valves which may be actuated to close in case of back- flow or the detection of radiocactivity. Radiation monitors will be -50- loaded sufficiently near each line to detect and actuéte-the closures before dangerous radiation levels are attained. Each of :the separate fluid service systems.is completely closed. The leakage of radioactive material into one of these systems will constitute a contained hazard rather than a release of activity. _ The coolant-salt lines are not equipped with closure devices. As previously mentioned, the coolant-salt pressure is kept greater than the fuel-salt pressure, and in the event of a heat-exchanger tube failfire, the coolant salt-will be pushed into the fuel. Shduld the differential pressure disappear, the reactor will be drained. Although the-coolant;salt cell is not leak;tight, containmént protection provided by a fldw of air maintains theficell at a negative pressure greater thgn 0.1l in. Héo, snd the air is monitored for radioactivity. 5.3 Shielding The shielding arrangement for the réactor equipmént is shown in Fig. 1T. The reactor is shielded around the sides and on the top by a 16-in. iron-water (50% Fe - 50% HpO by volume) laminated thermal shield. An INOR-8 casting located in the outlet dome of the reactor vessel, in order to reduce the dose to the fuel circulating pump, is also effective in reducing the dose from the reactor above the reactor cell. ‘ | ' The reactor-cell shielding consists of T ft of ordinary concrete covering the cell and aggregate and water in the 3-ft annulus between the reactor containment vessel and the cell wall. An additional 2 ft of ordinary concreté; which constitutes part of the cell wall, encloses the cell except for the section which is adjacent to the radiator room. The reactor-cell top shielding and sealing mémbrane were previously. " ‘described- in Sec. 5.2.1. The drain—tank cell is covered by .6 ft of ordinary concrete sandwiching a sealing membrane in the same manner as that described for the reactor container in Sec. 5.2.1 (see Fig. 16). | The penthouse (the portion of the radiator ceil extending abhove the main operating floér) has 2 ft of ordinary concrete shielding to provide shielding from the activated coolant during operation. The - 5] = radiation here results from F1%(n,q)N'® with a 7.4t-sec half life and from F'2(n,7)F2° with an ll-sec Half life and permits entry into the radiator room a few minutes following shutdown. . The fuel storage cell is covered by a L-ft-thick ordinary concrete plug. This cell, the hot storage cell, the decontanmination cell, and the fuel transfer cell are separated by B-ft-thick'ordinary-copcrete shadow-shield walls. - The estimated dose rate in a small area directly over the reactor through the 7 ft Qf ordinary-concrete shielding plugs is ~90 mr/hr, reducing to lower values in other locations over the cell. Should the actual dose rate.berthis_high, concrete -blocks will be stacked on the shielding to reduce the dose. In no'contifiuously occupied area. will the dose rate exceed 1 mr/hr. 5.4 Arrangement of Equilpment The arrangement of the reactor components in these containment areas 1s presented in Figs. 18 and 19. The layout of the fuel‘loop in the 24-ft-diam containment cell achieves piping stresses well below the ASME-code allowable values and provides satisfactory accessibility for maintenance. The reactor is anchored, but the fuel circulating pump above 1t is supported by spring hangeré. The primary heat exchanger is free to move horizontally and vertically; the inlet is aligned with the pump discharge nozzle, and the outlet with the reactor inlet volute. The components are joined by 5-in.-ID tubing (0.165~in.-thick wall), | with freeze flanges to permit removal. ' ' The fuel drain line connects the reactor to the drain tanks located in the rectangular container adjacent to and south of the_reattbr pifi. The two drain tanks and the flush tank are also arranged for ma@ntenance from overhead., The elevation of the tanks provides a head of 5 to 15 ft of salt to drain the reactor. Freeze valves are positioned between each tank to control the routing of the salts. The line extending:thpqugh the north wall connects these tanks with the storage tank. ‘ UNCLASSIFIED " ORNL—LR—DWG 52040 ' —=F : CLEAR WATER SHIELDING o NUCLEAR . L M — Ll L — } | | r | } REACTOR AND PRIMARY PUMP | | I l : | i PIPE SLEEVES | | } I = } -~ FILL AND DRAIN TANK NUMBER 3 || B i | ' = l 15 | ! 1 HOT STORAGE PITS } : ! l: L— CELL I ' 1l N X N ! 1 } DRAIN TANK NEUTRON SOURCE TUBE/ i ” NUMBER 2 — : WA —i | I : | | [ / I ' 4 k I t [ : g SECONDARY BuMP STACK AND AIR DUCT Y NUM L —=——= | FLusH TaNK] 4 P J_L CABLE AND PIPE Ti Y s , S - n n 1 1 \\ /’I 0 { \°1';" i / iL.—AIR DUCT ! ) i PLENUM—"h by ,f :: : [ " :I / I /! (L Ii 1 I ,l EXISTING FANS C:___r i i |1 | s BELLMOUTH INLET ' ES . ' = ! Fre BLOWER HOUSE @ | ' (o] L?_'i ./ - | | Ql~ :: a wy I fil /= TRACK ! L :i E_f i i | L] *, [ ..a'g- 7ft Oin. CONCRETE SHIELDING STEEL -BARRIER \ _ 4 FLUSH TANK WATER SHIELD \ MOTOR PRIMARY PUMP 8-in. SUCTION HEAT EXCHANGER — REACTOR—1L 1'%-in. DRAIN AND FILL LINE N\ " EXISTING STEEL-—\ TOP OF PENTHOUSE~ UNCLASSIFIED ORNL-LR-DWG 504104 _MOTOR SECONDARY PUMP BESREh TANK NO. 2 LOW POINT OF TANK ¢ TEST CELL 1 AIR PLENUM WALLS .1 Y5-in. DRAIN LINE PIT RADIATOR FILL AND DRAIN TANK NO. 3 Fig.19. Arrangement of Equipment: Elevation. ..Eg.. -5 Loading and unloading equipment is located in the pit directly east of the storage—tank cell. Provision is made for a shiéldedlcarrier on top of the concrete blocks covering this cell; manipulators through . the blocks accomplish the loading and unloading of salt.. Outside the contalnment areas in the south end of the bullding is the coolant_equlpment. The salt-to-air radiator is mounted ;n the coolant air duct. Two 5-in.-OD (0.165-in.-thick wall) pipes comnecting the radistor and the-heat exchanger enter the reactor container through bellows-equipped seals. The coolanf cell is nof completely sealed, because the N*© activity of the coolant salt does not éonstitute a hazard and_because-the coolant—salfi préssure is kept greater than the fuel-salt pressure to prevent the inleakage of fuel. 5«5 Maintenance _ The MSRE 1s designed for replacement maintenance instead of repair in situ. All parts of the fuel system are located so that they'can be placed by remote techniques,-and this arrangement also permits semi-. direct maintenance. Remote maintenance is accomplished from & shielded control room from which s General Mills manipulator can be directed to do work within the cell. After the shielding blocks are removed, the manipulator with lighting and television cameras is positioned on a track on top of the cell and above the equipment to be removed. The arm of the manipulator operates the*toois necessary to disconnect the equipment. The control-~ room operator then moves the manipulator out of the way and brings the high-bay bridge hoist into'position to 1ift the component and move it to the storage pit in the north end of the building. The hoist can then be used to replace the lower shielding plugs. 4' In the semidirect approach, & mobile shield is positioned over one block in the lower layer of conerete shielding blocks. The motorized shield is opened to permit remofial of the block which coveré the failed component, and is then closed, By using long-handled tools through openings in the shield, an operator standing on top of the shield dis- connects the component. Viewing is through lead-glass windows and 55 periscopes, The operator then retires to the shielded control room, opens the motorized.shield, and‘remotely controls the high-bay holst to 1ift the component and move it to the storage ceil.-.~ Figure 20 illuefrates the use of the remote manipulato; and shows the posifiion of the'shielded control room, whieh is efiuipped with lead-glass windows as'weli as teievision‘receivefs. Both these methods have been demonstrated afi ORNL.to be feasible means of maintenance. Additional experience under conditions of high radioacfiVity is necessary'before either method could be chosen for larger reactors. Since the removal of components necessarily requires breaching the containment cell and opening pipes in the fuel system, protection agalnst the escape of radioactive geses and particles must be given pafticular attentien. Meintenance procedures require that the fuel be drained, the pipes flushed with clean salt, and the system cooled before repair work is begun. When the secondary container is opened to permit work with the manipulator.or the'motbrized‘shield, the secondary container ie venfiilated at a rete of 10 to 15,000 cfm. If it should be néceSsary to remove all the’shieldifig blocks, a velocity of at least 30 ft/fiin could be maintained downward through the openirg. 'However,'the normal fiaximum opening will be only 100 ft?,'and the velocity of air will be 100 to 150 ft/min. When the reector pipes are epened, nitrOgen_Wili be purged into the pipes until temporary closures can be fastened. Closures are also attached to the-flanges on tfle failed component to prevent the escape of ectivitj dufing removai. Similar techniques have beefi-developed on HRE—2 and haje'proved satisfaetory; The coolant-system pit may be entered for direct maintenance as soon as the system is cooled. The aciivity'in the salt decays by'e factor of 10° in 2 min, andeho other sigfiificant centamination is expected. -56- _ UNCLASSIFIED ORNL-LR-DWG 56882 1 T T T 1 = n . N - ) o ‘ ] | LK "; & REMOTE CONTROL ROOM FOR HOISTS AND MANIPULATORS \ m : CRANE BAY % 1 T £ ‘ REMOTE CONTROL ROOM AND SERVICE AREA MANIPULATOR ST TV EY = hl,: n=4= ! ilLJ? m =M= m =l Il SEAL - WELDED ' MEMBRANE-/ m=n=q CELL THERMAL =y SHIELD m= [l — — CONTROL TUNNEL/"; =M= ANNULUS FILLED WITH WATER-AGGREGATE MIXTURE EYi == (HSI=0 REACTOR CONTAINMENT 1 MOTOR FUEL PUMP INSUL ATION | =11 == ZHEN = i = Zuy 9 Y = =7 l’sm =m W=y = 7 Fig. 20. East-West Sectional Elevation—Building 7503. -57- 6. CONSTRUCTION, STARTUP, AND OPERATION 6.1 Construction - Although no special construction‘practices will be employed in assembling the MSRE, many special precautions will be taken to ensure a high-quality, clean, and leak;tight assembly. A detailed specifica-. tion, requiring quality control better than existing commercial codes,’ has been prepared'for each cdmponent of the éysfem (seézAppendix C). Nondestructivé inspection techniques such as ulfirasofiic.testing,.dye; penetrant inspection, x-ray examinatién, and helium leak testing are employed at each fabricator's plant, under the Supervisidn of ORNL inspectors.. Assembly at the reactor éite will be similarly examined. After completion the separate systems will be leak—teéted using rate- of-pressure fise and isotopic-tracer techniques. The'leékage rate from the entire fuel-containing system must be measured to be less than 1 cc/day. _ _ After construction is completed, a period of several weeks will be occupied by remote-maintenance practice. 6.2 Flush-Salt Test Tt is planned to demonstrate the mechanical performance of the system by a several-month period of'testing with a flush salt in the ffiel system. Each piece of equipment will be ‘examined to determine whefher it pérforms as designed; insofar as this can be determined without nuclear heat generation. The flush salt will also serve to scavenge oxygen and to remove other i@purities. Another important benefit of the flush-salt test will be the development of the operéting skills necessary for satisfactory control of the system variables. 6.3 Startup After the flush-salt tests are concluded, the neutron source will be installed and counting rates will be determined on each of the two count-rate circuits before the salt is drained. Then the drain tanks ..58_ will be loaded with fuel salt containing sbout three-fourths of the UZ%S “estimated for criticality at the minimum temperature (1000°F). The fuel salt will be slowly pressurized into.‘the reactor, which has previously been heated to l2OQ°F. Only one-half of the available poison will be inserted during this initial loading and for later fuel additions. The partial insertion will allow experimental flexibility such as withdrawal to test for criticality after loading. If criticality is not attained after any addition, the system temperature will be gradually lowered to 1000°F if necessary, and counting rates will again be determined. This procedure will be followed after each fuel addition until the reactor is critical. Further fuel additions will increase the critical tempera- ture and provide information on the over-all temperature coefficient. Quantities greater than 120 g of U235 w111 be added to the drain tanks and transferred to the reactor; quantities of 120 g or less can be added at the bowl of the fuel circulating pump through the sampler-enricher mechanism. 6.4 Approach to Power After a series of zero-power experiments to determine the nuclear characteristics of the system, the pbwer will be raised in increments of a few hundred kilowatts over a period of several weeks until full power is reached. At each successively higher power level, information will be collected and fuel éamples will be analyzed for studies of xenon and fuel permeation of the graphite, fuel stability, power coefficient, radiolytic-gas handling, and power stability. 6.5 Operations Personnel The operation of the MSRE facility will be the responsibility of the Reactor Division Operations Department. The previous assignments of personnel in this group include the construction, startup, and operation of four other experimental reactors: the Low-Intensity Test Reactor, Homogeneous Reactor Experiments. 1 and 2, and thé Aircraft Reactor Experiment. -59- The experiment will be conducted on a three-shift basis, employing four operatlng shifts and a day staff for- reactor analys1s and planning. Each of the four shifts will be headed by a senior-level superv1sor for - the flrst few months. A Junlor engineer and three or four nontechnlcal operators many with the prev1ously mentioned reactor experience, will complete the shift organization. | . The Reactor Analys1s Group will be composed of four to six engineers with a broad experience in fluid-fuel reactors. Tts function will be principally to plan, supervise, and analyze the experimental program. - Over the years:this organizatioh has developed training,loperatihg,. and maintenance pfactices which especially contribute to ekperimenfal- reacpor safety. The same methods and policies will'be applied to the Molten-Salt Reactor Experiment. -60_ T. HAZARDS ANATYSIS The hazards inherent in the MSRE are considered on the basis of _ possible damage, first to the primary container and second to the sec-. ondary container. Finally, assuming thgt activity escapes‘from the secondary container, the danger is considered to personnel at the site and in the surrounding territory. T.1 Damage to the Primary Container The possibilities for rupture of the primary'container were investigated in several categories: (1) reactivity excursions, (2) melting of walls, (3) failure by excessive stresses, (4) corrosion. T.1.1 Reactivity Excursions Excess reactivity can result from the following.unusual circum- stances. (a) Startfip Accident. The MSRE is started normally by transferring hot (>1000°F) fuel into the preheated (>1000°F) circulating system. ‘The normal fuel concentration is sufficient to make the reactor chain- reacting at 1200°F when'at full power with the full éffect of the power coefficient and xenon poisoning. When these effects are not counter- acting excess reactivity--that 1s, at startup--the poison must be suf- ficlent to hold the reactor subcritical. If; by some instance, the poison were ndt inserted when fuel is pushed intc the core, the reactor would begin to generate power unexpectedly with the core only partly full,‘ In the worst situation the core would continue to £ill, the - reactor would continue to generate heat, and the reactor temperature wofild rise to a final temperature of 1L0O0O°F. Althpugh such a tempera- ture rise is undesirable, it should not damage the reactor. This accident will be'analyzed in m?re detail on the reactor simu- lator before the final design of the conirol circuits.. It is planned to protect against the startup accident in several wajs, For routine startups, the control circuits will require that all the poison be inserted before f£illing can be begun. A large number of thermocouples _61 - distributed on-the reactor vessel, throughout.the heat removal system, and on the drain tank must indicate at least 1000°F. The rate at which fuel can be transferred from the drain tanks to the reactor vessel will be limited so that the loading time will be approximately one hour. The reactor will be filled in several steps, with sufficient delay between steps. for the neutron multiplication to be determined. The reactor can easily be drained if eriticality is approached unexpectedly, because the drain line is also the fill line. - When the system has been filled to the operating level, there will be a time delay before the pump cen be started so that the drain line can be frozen and the temperature distribution in the reactor system can be checked. Any lafge temperature differences between the reactor ; and the heat‘removal system will be reduced by natural-convection circu- lation during this delay time. ‘ A variation of the startup accident would involve filling the - reactor and withdrawing poison to make the reactor critical at the operating temperature of 1200°F before starting the pump. Assume that the heat removal system is operating so that all the fuel in the heat. exchanger is cooled to 850°F., If the circulation pump could be started to cause the 850°F fuel slug to traverse the core at the average circu- latiqn rate, the reactivity would increase at the average rate of 0.15% 2Mk/k per second for 7 sec. Actually thermal-convection circulation begins when the fuel in the heat exchanger is cooled so that the reactor gradually rises in power to satisfy the demand. According to reactor simulator studies, thermal-conveetion circulation is sufficient to extract 9.4 Mw with a temperature rise of 200°F across the reactor. The studies do not indicate a likelihood of damage from the cold- -slug accident. ' Nevertheless the fuel pump will be started et.reduced speed, and the rate at which the speed can be increased will be regulated to limit the rate at which reactivity can be added by introduction of cold fluid. | - (o) Graphite Problems. Four potentiasl reactivity problems are associated fiith~the3presence of bare graphite: ' | o 1. compatibility with the fuel salt, 2. fuel penetration into graphite voids, 62- 3. 1irradiation-induced shrinkage, 4, xenon penetration into graphite voids. If the salt and graphite were chemically incompatible at any tem- perature above the liquidus of the salt, there would be concern sbout reactions which might result in uranium'separation. Since no reaction has been observed in several thousand hours of loop tests or in labora- tory studies, there 1s considerable confidence that none occurs, and that graphite and salt can be judged completely compatible. An associated pfdblem is that of penetrétion of fuel into the T% of accessible volds in the graphite. Many out-of-pile experiments have been done on the wetting and permeation of graphite by molten salts. The tests indicate that graphite is not wet by the MSRE fuel salt and that permeation of MSRE grades of graphite does not exceed 0.5 vol % at pressures as high as 150 psi. One in-pile test has been done and it indicated that radiation effects do not change this behavior. Both in-pile and out-of-pile testing will be continued to determine whether any condifiions that can be produced in the reactor will cause wetting and appreciable penetration of the graphite by fuel. It is believed that the tests will continue to show no significant penetration of salt into the pores of the graphite. However, in spite of this infor- mation, operation of the reactor will be monitored fdr long-term effects in the large mass of graphite at high radiation levels. Permeation greater‘than 0.5 vol % should occur slowly, and the accompanying rise in reactivity would be slow. This would be indicated by a gfadual rise in critical temperé.ture° The increase in reactivity can‘5é easily counteracted by the control system, by omitting fuel addi#ions to com- pensate for burnup, or by adding lithium or thorium tolthe fuel as a poison. The question of graphite compatibility is discussed nore fully in Appendix F. ‘ ‘The only potentially hazardous situation that cculd result from several per cent of fuel soaking into the graphite porés exists during maintenance. The decay of fission broducts in the graphite could raise the central graphite temperature to about 2000°F in about 200 hr after shutdown. If the core vessel were opened in air with the gréphite at 2000°F, some burning could occur, the quantity depending on the amount 63- of oxygen available. Undoubtedly the evolution of CO and COp from burning would carry fission products into the reactor cell. - It is unlikely that more than 1% of the graphite would burn or more than 0.06% of the fission - products would escape the reactor core before thesreaction could be stopped by closing the vessel. The rapid ventilation of the secondary container "~ would deliver the activity to the filters; only noble gases would be.ree leased to the stack; so the biological hazard would not be significant. Protection against this mishap. is conceived to be:.. (1) thorough cooling of the graphite with flush salt prior to beginning maintenance work, (2) developing practices for rapid closure of lines which might cause a "chimey" effect through the hot core, (3) insistence on good ventilation and monitoring so .that the maintenance work can be halted and closures can be made to’ stop the escape of activity, and (4) a purge system of nitrogen or helium which provides a blanket of gas in the contalner and rednces the likelihood of the entrance of air. iIrradiationwinduced shrinkage of the graphite was studied to determine its effect on neactivity.‘ Because the shrinkage results from neutron bom- bsrdment, the flux variations across the reactor produce dimensional changes which vary with position. This results in stresses within the separate graphite pieces, and bowing as well as axial and transverse shrinkage. The resultant stresses may be considerably relieved by creep and annealing, but even without these mechanisms, the.graphite should not begin to form cracks in less than two years' exposure at 10 Mw. As shrinkage gradually occurs, the space will £i11 with fuel salt. This_ process is so slow that the reactivity increase (which will result if the spaces fill with fuel salt) will be almost exactly counterbalanced by the buildup of long-lived fission product poisons. If there is.a change in reactivity, it is predlcted to be slightly negative. Because of the very slow rate of change, compensation can be easiiy accomplished by adjustment of the rate at which fuel is added to the system to compensate for burnup. | | | ' The description of the core graphite assembly (Sec. 2.3) included a mention of molybdenum bands which restrain the deformed graphite. Originally the arrangement of fuel and- graphite was such that failure of the molybdenum bands would cause e‘reactivity excunsion, but with the present design, any reactivity shift would be negative and not -6h- significant (0.1% Ak/k). The molybdenum could be omitted from the present design, but 1is being retained as an experiment because the metal might be used in future reactors. ' Thus it appears that neither breakage of the bands nor slow shrink- age of the graphite bars will be readily detectable while the reactor is operating with the original charge'of fuel. However, & possibly hazardous situation would arise if either change took place and the initial fuel charge was replaced with new fuel. Without the fission product poison of the old fuel, the Same concentration of uranium in the new fuel would re- sult in a reactivity approximately 4% greater than the original. This is but one reason why a reloading of the reactor, should it ever be neces- sary, will be handled as carefully as the original critical experiment. The precautions for that experiment (see Sec. 6.3) should protect against acclidents in future loadings. Xenon penetration into the graphite voids will occur by transfer from the circulating fuel and/or from any fuel which has penetrated the graphite. For 0.5% fuel penetration thé total xenon effect at 10 Mw is estimated to be 1.3% Ak/k. At its peak after power reduction the poisoning will amount to ~4% Ak/k. Xenon will be removed by the stripping system as long as the fuel is circulated. No excess reactivity will be provided to over- come xenon buildup on reduction of bower. 7.1.2 Fuel Separation One of the few weaknesses of the fuel composition selected for this experiment is its vulnerability to large amounts of oxygen as gas or in compounds. The ZrF4 component of the fuel is for the purpose of reacting fiith‘any oxygen. and thus'preventing the precipitation of U0z. Extensive laboratory tests have shown that eo;long as the ratio of'ZrF4/UF4 is 3 or greater,_Zan iz always precipitated in preference to U0s. The actual ratio of ZrF,/UFs4 in the fuel is 5, which gives a good margin of safety -to maintain the'fuel within its known limitsvof stability. Approximately 2.5 £t3 of water or 7000 ft2 of air (ST?) would have to react with the salt to precipitate the excess zirconiumd These amounts are considered to be very large in view of the care being used to prevent the system from becoming contaminated. The periodic sampling of the fuel should - reveal the presence of centaminents long before this level is reached. More informstion on the fuel chemistry is presented in Appendix A. -657 In spite of the apparently good protection by ZrF4, and in spite of -the extraordinary.efforts to keep the system free of oxygen and water vapor, the consequences of U0z separation must be‘examined.- After pre- cipitation, U0- would tend to remain suspended and circulate with the fuel. If it were uniformly distributed, it would be indistinguishable, nuclearwise, from the normal fuel. However, ‘after a while it would be- gin to collect in low velocity areas or at the points of lowest tempera- ture, and the most likely location 1s the reactor-vessel plenum under the graphite. At this location, 2.5 kg of UQ- deposited per square foot of surface would incresse the temperature only 100°F. | , The worst possibility 18 to assume the sudden transfer of the sepa- rated material into the core. The resulting excursion would depend among other factors, on the quantity of U 235 involved; so it is worth- vhile to estimate the limits of detectable fuel loss. | Assuming that a 25°F reduction in critical temperature could be easily detected, the equivalent loss of U3 yould be 160 g from the core or 625 g from the entire fuel system. If this meterial collected in the reactor-vessel plenum it might all be returned to the core at a rate which would be limited by the T-sec fluid transit time through the core. The system temperature would rise temporarily by 100 tollSOPF but no damage shaild result. The control system would normally eliminate most of the temperature transient. . Frequent analysis of the fuel and calculation of the U235 inventory will be enother check on uranium separation, but the limits of detection by this method are approximstely +3% of the fuel inventory, which amounts to 1500 g. Since Zr0O- appears in the salt before U0z and 1s -easily - recognizable,‘an inspection of the frequent salt samples will provide a warning vhen oxygen enters the system. 7 1. 3 Flow Stoppage Several nonroutine situations, ranging from probable to nearly in- credible, that involve flow stoppage in the feed or coolant circuits have been analyzed on an analog computer. ‘The necessary protective actions and potential hazards are discussed below ' - (a) Fuel-Circulation-Pump Failure. Failure of the fuel circulation pump is highly credible because it could result from an electrical failure in the circuit or pump motor, or from a mechanical defect, such as a - ~66- bearing failure. Instantaneous introduction of the delayed neutrons normall& genérated'out of the core amounts to O,lfi%fiék/ko_ All this amount is not effective because of the gradual stoppage of the pump and the exponential decay of the precursors; the expected temperature rise is less than 150°F. Failure of the pump causes automatic insertion¥of poison. Convectilve circuiation is adequate to remove the afterheat, but if the reactor'tem@erature continues to increase, the galt systems will be drained. | . (b) Coolant-Pump Failure. Failure of the coolant pump is also highly credible and for the same reasons. Agaln the pcison is auto- matically inserted, and dréinage of the salt in both systems 1s con- venient protection. (¢) Simultaneous Pump Failures. Simltaneous failures of pumps in the fuel and the coolant loops is also credfble, because a power out- age or a burnout of a main power bus would stop both motors. Automatic action of thé poison is the same as.'before° A power outage automatically starts the diesel-generators (within 15 sec), and operation could be resufied. For other conditions which result in pump stoppages of long duration, thermal-convection flow in the two systems is: sufficient to handle the afterheat, and both loops can be drained. (d) Flow Stoppage in Fuel toop. Flow in the fuel loop might be stopped by plugging somewhere in the circuit; this 1is incfedible as an instantaneous event and not very probasble even ag a gradual occurrence. In the case of gradual plugging, as indicated by the temperature drop across the heat exchanger, the reactor could be shut down and drained rbutinely. In the unlikely event of instantemeous flow stoppage, the _ situation would be similar to that already mentioned for fuel-circulation- pump failure. The increase in neutron flux would dutomatically cause the polson to be inserted, shutting down the reactor. The reactor would be drained, because no cooling would take place and the afterheat would cause the reactor temperature to rise. ' In the situations considered above, there would be no” concern if the fuel could be drained. Thus the most hazardous situation is a plugged drain line combined with Interference with heat dissipation 67~ to the radiator. Any one of four possible‘circumStancesmcould~produce‘: these conditions: fuel circuit plugged or partiaslly drained, the cool- ant cifcuit»plugged or partially or completely drained. The power productlon from afterheat within the core and the bulk mean temperature of the fuel, graphite, and INOR-8 associated with the core vessel are shown in Fig. 21 as a function of time after reactor shutdown. With the reactor heaters off the temperature would rise to 1500°F in 16 hr, to a maximum of 1800°F in 105 hr and then would slowly decline. The reactor pfobably would be damaged and would have to be replaced if held ~ at 1800°F for several hundred hours. Methods of providing cooling to maintain the temperature below 1500°F in an_emefgency are being con- sidered for incorporation in the final design. 7.1.4 Control-System Failure The“control system is not a safety system and is not required to protect the reactor against calamity. Although relisbility will be an impbrtant»criterion in designing the control system, the consequence of complete failure was exemined. With the small amount of excess reac- tivity present, sudden removal of all the control poison would result in a final temperature near 1400°F. It is unlikely that the thermal stresses produced 5y this sudden rise in temperature would cause seriofis,damage to the assembly, or that salt would be spilled into the reactor cell. 7.1.5 Drain-Tank Hazards Two possibly hazardous problems are encountered when the drain tank is filied with fuel salt from the operating reactor. These prdblefis are: afterheat and potentially-critical fuel configurations. If the fuel is drained a short time after shutdown, provisions for removal of after- heat are necessary. Without heat removal, the temperature of the salt would rise approximately 500°F in a 3-hr period starting 1/2 hr after shutdowh, assuming the reactor had been ppérating af full powér for 1000 hr prior to shutdown. This condition is avoided by pfoviding a 100-kw heat removal system (see Sec. 2.7.1) to keep the bulk mean tem- perature of the salt below 1400°F during storage in the tank T POWER GENERATION (kw) 150 UNCLASSIFIED ORNL-LR-DWG 56883 TIME AFTER SHUTDOWN (hr) 100 / < TEMPERATURE 4 POWER GENERATION — 20 30 40 50 60 70 80 90 100 10 1800 {700 1600 1500 1400 1300 Fig. 21. Afterheat Power Generation and Temperature Rise of Core Vessel vs. Shutdown Time. BULK MEAN TEMPERATURE (°F) _'89- -69- Water was selected as the coolant because of the relative simplicity of the associated cooling system.~-The water and the salt are never in contact with a common wall, as is evident from Fig. 9. Water is fed through the center tube, and the steam forms in the annulus. The circuit .is completely closed and requires no pumps. In an emergency, the steam can be vented up the stack and fresh water added to the system. An emergency reservoir is installed to provide cooling for 6 hr. Several configurations of fuel inside the fuel drain: tank were investigated for the possibility of some configuration having an effec- tive multiplication constant greater than unity The first configuration investigated was the flooding of the cell with water, which is a possible measure in case all other cooling of the drain tanks fails. The water would then act as a neutron reflector around the drain tank. The k oo for this situation is O. 852. (The Kope Without the water is 0.826. ) Flooding, therefore, does not present 8 criticality hazard. ' - The second problem considered was the precipitation of the uranium by an oxidizing agent. If air should accidentally enter the drain tank through the helium blanket system or an air leak in the drain-tank | vessel and the salt should maintain contact with sufficient oxidizing agent, the uranium, thorium, and zirconium would be precipitated. The precipitate would form a semisolid mixture with the salt, and the LiF and BeF» would exist as a liquid above the semisolid. The k. pp calcu- lated for this configuration was only 0.185. Precipitation of the uranium, therefore, does not create a criticality hazard. 7.1.6 Other Poeeibilities for Primary-Container Damage There are several other accidents in which the integrity of the primary containment might suffer. (a) Freeze-Valve Damage. The freezing and thawing of the freeze valves could conceivably result in a rupture of the piping. This is specifically minimized in the design by making the freeze-valve section as short as possible so that there is a small danger of bursting as a ‘result of-efifirepment of liquid between the ends of a plug. Because -T0= the salt expands-as it thaws, special precautions are taken to apply the heating so that expansion space is always avallable. A single freeze valve has been frozen and thawed more than 100 times without apparent damage. (b) Freeze Flanges. The flanged joints in the fuel circulating loop are also possibilities for failure. Freeze flanges were selected as the simplest and most reliable Jjoint aveilaeble. The strength of the bolting and the flange compression menbers are considerably in excess of the strength necessary to maintain s tight joint. However, an analysis of the original flange design8 suggested several improvements. The recommendations of this study were incorporated in a new design which will be thoroughly tested prior to use in the reactor. The old flariges fiere cycled more than 100 times without any evidehce of failure, and the new type is expected to be better. (c) Excessive Wall Temperatures. Overheating of pipe and vessel " walls might occur from the external electric heaters or from internal ganma heating. The heater elements have a melting poini several hun- dred degrees above that of the INOR-8, but it is not considered likely that the INOR-8 could be melted by external heating as long as salt is present inside the pipes. If salt were not present, the pipe wall might melt before the heater element, but in this csse a small fraction of the activity would be released. | | During reactor operation various components are exposed to high . gamma fluxes which result in gamma heating of the components. If this heating should produce large temperature grasdients, excessive thermal stresses may arise. The effects of gamma heating on the core vessel and the grid structure vere investigated because these structures are in regions of the highest gamma flux. | The gamme heating of the core vessel results in a temperature difference of only 1.3°F across the vessel walls and produces a calcu- lated thermal stress of 300 psi, which is not serious. The support grid structure experiences a.temperature difference of 3.8°F across each grid. The resulting thermal stress is calculated to be 850 psi. | -1~ Gamms. and beta heating of the top of the pump bowl results in a thermal. stress of 17,000 psi at a temperature of approximately 1000°F. A static pressure stress of approximately 4000 psi exists at the same point, resulting in a combined stress of approx1mately 21 000 psi. The meximum allowsble static stress at this temperature is 16,000 psi}. However, the ASME Unfired Pressure Vessel Code allows the combined ] static and thermsl stress to be as much as 1.5 times the maximm allow- sble static stress (24,000 psi at this temperature). Hence, excessive thermal stresses do not exist at the junction in'qnestion. ' ' Thus a wall failnre as a result of beta-gamms heating does not appear reasonsble., | Another possibility for melting of a primazybcontainer wall is in the fission—product adsorption beds in the off-gas system. There have | been'1nstances9 where carbon beds became ignited in the presence of : oxygen and consumed a portion of the charcoal in the beds. This acci- dent is much less likely in the MSRE because of the special efforts to exclude’oxygen-from.any part of the reacfor system. Furthermore, the high temperatures necessary for ignition could not normally cceur ) because the beds are submerged in a pool of cooling water. In the. unlikely event that ignition does occur, the resulting high bed tem- peratures will alarm, and the inlet and exit valves will be closed. The blanket of CO- resulting from the fire will extinguish the fire. As final protection the beds are enclosed in a secondary container. () Excessive Stresses. In a normal thermal cycle the témpera- ture‘of‘the reactor varies between TO°F and 1300°F. With such & range there are possibillties for excessive stresses as the piping expands o and contracts. Particular attention has been given to providing a flexible layout. Analysis of the extreme conditions indicate that the maximum stress caused'by expansion or contraction is only 7050 psi. Instiumentation is provided to observe the normal rate of heatup or cooldown, which will not exceed 100°F/hr. | | (e) Corrosion. Another posaiblllty for failure of the primary container is by corrosion. As reported in detail in Appendix A, the corrosion rates experienced W1th the INOR-8 alloy have been very low - (less than 1 mil/yr) for periods as long as 15,000 hr. All available evidence indicates that corrosion is not likely to be a cause of piping failfireé. Numerous in-pile capsules and two L0O-hr in-pile circulating corro- sion tests have been examined for evidence that corrosion under irradi- ation was different from that out of pile. Particular attention was paid to the possible effects of free fluoride. No evidence was found that indicates that high irradiation altered the normal corrosion pattern. 7.1.7 Detection of Salt Spillage The escape of activity from the primary container would be detected by radiation monitors. If the spillage were into the secondary con- tainer, the activity would be indicated by-monitors on a system which cohtinfiously samples the cell atmosphere at several locations. Leakage into a service line (e.g., the cooling water) would be detected by monitors attached to the line just outside the cell. In eithér_case, the action of the monitors would be to stop power removal and insert poison. The salt would be drained unless the leak were in the drain tanks. (.2 Rupture of the Secondary Container Assuming that radioactifie material has escaped the primary contain- ment and has spilled into the secondary containment, the next concern is preventing the escape of the activity from this second barrier. Two means by which the secondary container walls might be ruptured are by missiles and excessive internal pressure. 7.2.1 Missile Damage It has not been possible to devise a situation in which damage by missiles appeared likely. The maximum pressure expected in thé reactor system is less than 100 psig, and the INOR-8 material is very ductile at the normal operating temperature. No very large pressure excursion can be envisioned without assuming that the large inlet and exit lines are ~73= both frozen.. In any caée, the vessel and component sections gre rela- tively thin.. Furthermore, the vessel is completely -surrounded by the steel thermal shield, which is good protection against the possibility. of missiles tearing through the wall of the container. 7.2.2 Excessive Pressure (a) Salt‘Spillage.: The spillage of the salt at a high temperature does'have pqséibilities‘for raising the ceil_pressure to high values. A rapid spill into the cell of all the salt in both the fuel and coolant systems would heat the Qell atmosphere sufficiently to produce a 2.fiApsig final pressure. If the fuel were released as a fine spray,-the‘maximum ' pressure would be 16.L psig. The worst situation would be the simultaneous release of the salt and the inleakage of the correct amount of water to allow the generation of steam without subsequent cooling from additional inleakage of water. Taking into account the large heat capacity of the secondary container- and reascnable heat transfer coefficients, this worst accident wouid-pro— duce a peak pressure of 39 psig, which is less than the U5-psig test pressure of the container. ‘ The details of these calculations are presented in Appendix E. (b) Oil-Line Rupture. The fuel-pump lubrication system contains a maximum of 28 gal of oil, which, in the event of an oil-line rupture, could come into contact with the hot pump bowl and reactor vessel. _With an atmosphere containing thé normal 21% of oxygen in the cell,‘the oil. ~would burn, producing an excessive pressure 1n the cell, or would form an explosive mixture which might later be 1gnited. To ensure against containment damage by these possibilities, the . oxygen content of the cell is kept below 5% by dilution with nitrogen.‘ Nitfogen'will be fed into the cell'continuously to maintain the oxygen content below this level. 7.2.3 Acts of Nature (a) Earthquake. Information on the frequency and severity of earth- quakes in East Tennessee has been obtained both from-Lynch (letter from J. Lynch to M. Mann, Nov. 3, 1948, quoted in A Report on the Safety 4{ . 20 -/ STAINVILLE KN O X \ wmascor T ] FOUNTAIN CITY ] WINDROCK \ BRICEVILLE ¥ GOBEY, FORK MT. SER s ON v O WARTBURG g peTROS N A N CLINTON & {+] OLIVER e °MAYL‘.‘ND M O R G A I(?JOALFIELD ) ; SR[NG o WHITTLE SP \ CUMBERLAND . W KNOXVILLE CROSSVILLE \ /‘9 ////‘ e : \ . HARRIMAN /’4’7\"" CRAB ORCHARD S. HARRIMAN »EMORY GAPoO g AROCKWO0D ; A NEUBERT o ~ KingsTon HFIR4 <] MARTEL ROCKFORD LENOIR CIT “ \ i o QD ° / ! 1RING CITY ’ o oS\v‘-'EETWATEFl oMADISONVILLE MONROE -~ Q D ) o MOUNT VERNON Q) | \/\ © TELLICO PLAINS' < O , / e B4 Fig.22. O MAP OF CITIES AND COUNTIES SURROUNDING OAK RIDGE AREA 10 0 10 20 30 MILES = = T S e o s e e | o 0 § E g, o S o - “‘__) w W " 1 : : NI Oe } 2 - [ ] X g “ ’ D '9 : . M N-18,800 DIESEL ¥u‘sg TAN K. x GENERATOL 0 H_\_J,fi,_fi_/ 1 N- 18,700 | JL1 1750 lSWITEH uoufii/flHE. /"‘ & N-18,000 ‘ CoMPLESSOn HLOWENL WOLUSE-S CHALCOAL PIT- ] / N-l&,soo TSNEW P - T BLO N-18,%00) N-18,200 N-18,100 _PLoT PLAN SCALE "= (00" NOT CLASS|\FIEDR PLOT PLAN MOLTEN SALT REACTOR EXPERIMENT BUILPING 7503 | -80= Appendix A CHEMISTRY AND CORROSION Tuel Composition and Stability The MSRE fuel 1s a mixture of molten fluerides with the composition LiF-BeFo-ZrF4-ThF4~UF, (70-23-5-1-1,mole %). The latest experiments | ' show that all three quadrivaelent cations behave similarly and that 15% of these cations can be contained in a liquid solution of LiF and BeFo at 40°C (824°F). Research is being continued in order to have a com- plete phase diegram for the system. ‘ i The purpose of ZrF4 in the fuel 1is to prevent the uranium from pre- cipitating as U02 in the presence of small amounts of an oxidizing agent. When solid BeOo is added to a molten mixture of LiF, BeF,, ZrF4, and UFy with a ZrF./UF, ratio of 2, UOs is precipitated and no ZrO, is found. When the ratio ZrF4/UFs is increased to 3, ZrO, is precipitated and no UO» is found. This indicates that the reaction UFs + ZrOp 2 ZrFy + UOp has an equilibrium constent between 2 and 3; that is,- ‘ - (zrF, 1 (U0, ] ZeE,] K = . - = . ~ 2.5 [UF), 1lzr0, | [uqJ ~ (the activities of Uo,, and Zr0,, both being solids, are taken as one) - Therefore, to prevent precipltatlon.of vo,, a ratio ZrFL/UEh of not less than 3 is necessary. The actual ratio in the fuel 1s 5, which gives a good safety margin, and maintaips the fuel well within its known.limits of stebility. o | ~ The vapor pressure of the fuel salt at operating temperatures is about 10 -2 mm Hg, as shown in Fig. A.l. The liquidus of the fuel salt is 440°C (828°F), At this tempera- ture a solld phase appears having LiF, BeF,, and ZrF), in as-yet-unknown'. proportions. At 431°C (800°F), an edditional solid, 2LiF:BeF,, forms. The next known phase appears at 425°C (797°F). This is TLiF-6(U,Th)FL, PRESSURE {(mm Hg) 1072 -81- UNCLASSIFIED ORNL-LR-DWG 56885 —-——————l———-—-—-—-—-—iatm.———-—-"—-— — T —— l0g,q #= 32— 15,450 —Tk_ —-6.039 logy, 7; P=PRESSURE {mm of Hg) 7= TEMPERATURE {°Kelvin) 1000 2000 3000 4000 5000 6000 TEMPERATURE (°K) l A I | | | J { | i I 1000 2000 3000° 4000 5000 6000 - TEMPERATURE (°C) | L I l ! | ! } | | P 1000 3000 5000 7000 9000 11,0600 TEMPERATURE (°F) Fig. A.A. Estimated Vapor Pressure of Fuel 3alt. -82- which is 15 mole % U, 85 mole % Th, Thus it may be noted that there is no tendency for uranium to concentrate at the freeze flanges or freeze valves. | The compositions of the fuel andlcoolant salts are matched so that it is impossible to freeze the ffiel salt by removing heat through the coolant salt. The coolant-salt liquidus is 450°C (842°F). Below this point about 96% of the salt forms 2LiF-BeF,, which would not circulate. This is 18°F sbove the liquidus of the fuel. The phase diagram of the coolant salt is shown in Fig. A.2. Corrosion of INCR-8 by Fuel Numerous corrosion tests have been comyleted‘with fuel mixtures of the type to be utilized in the MSRE. Results of 37 INOR-8 thermal- convection loops, 17 6f which operated in excess of one year,; show complete compatibility between INOR-8 and the beryllium-based fluoride systems,ll“l3 Experiments conducted in INOR-8 forced-convection loops for one year or more similarly show low corrosion rates in fluoride mixtures of this type.ll'l3 The operating conditions of these experi- ments are shown below: Forced Thermal Variable Convectior Loops Coavection Loops Fluid-metal ' o o interface temp. 1300°F 1350°F Fluid temperature ° o gradient 20C°F 170°F Flow rate ~ 2 gpm ~ T fom Metallographic examinations of INOR-8 surfaces following salt exposure in these loop experiments reveal no corrosion effects in time periods up to 5000 hr. At times ionger than 5000 hr & thin (less than 1/2 mil) continuous surface layer develops at the salt-metal interfsce. A quantitative measurement of the corrosion rate occurring in an INOR-8 forced-convection system containing I..:I.}.i‘»Be__F‘E-uUFLF (62=37=1 mole %) A TEMPERATURE (°C) UNCLASSIFIED ORNL-LR-DWG 56886 900 800 700 600 Lif + LIQUID \ N / BeFy+ LIQUID 400} 2LiF-Ber\\ e + ~LLiQuiD \/ b LiF +2LiF -BeF, o . : W 2LiF -BeF, +BeF2 (HIGH QUARTZ TYPE) 300} "3 [3V] L o : 2LiF -BeF2 & LiF -BeF, +BeF, (HIGH QUARTZ TYPE) L ' & ] | 560 | LiF-BeF2 3 LiF -BeF, + BeF> (LOW QUARTZ TYPE) - - LiF 10 40 50 60 70 80 90 BeF, - BeF2 (mole %) A.2. Phase Diagram of Coolant Salt. ..Eg_ -8)4- was carried out by means of carefully fieighed and measured inserts located at the point of maximm salt temperature (1300°F). %13 mme inserts, removed after test intervals of 5000, 10,000 and 15,000 hr, reveal relatively small weight losses, as shown below: Timé (br) 2Wéight Loss . ' mg/cm mg/cn” /mo 5,000 1.8 0.26 10,000 2.1 0.15 15,000 1.7 0.08 The weight losses do not increase measurably after the first 5000 hr. No changes in the wall thickness of the inserts are detected based on before~ and after-test dimensions. In some loops where there is evidence of contamination by water or some other oxidizing agent, greater attack is found in the hot regions, with roughening and pitting of the surface to depths of 1-1/2 mils or more. In the cold-leg regions, magnetic metal crystals loosely adherent to the cold-leg wall are found, composed pfedaminantly of nickel and containing only-minor amounts of chrdmium and iron. ‘Several brazing materiasls have been developed for possible use in the heat exchanger. Three nickel-base alloys, one gold~-base alloy, and pure copper were tested in thermal-convection loops at l300°F without showing any signs of attack after 10,000 hr of operation when they were used to join INOR-8. . Corrosion Reactions The main corrosion mechanism is selective leaching of chromium, not because of physical solubility of chromium metel in molten fluorides, but by chemical reaction of this metal with oxidizing agents present in. the melt or in the original metal surface. Typical impurities produce corrosion by the following reactions: @ + s * 5 NlFé + Cr - CrFé + Ni FeF, + Or - CrF, + Fe | 2FeF3 + 30r - 3CrF, + 2Fe 2CrF, + COr - 3CrF, Oxide films on the metal walls can react with the fuel constituents (ZrFL or-UFh) to yield structural metal fluorides: 2Ni0 + ZrF4> - 2NiF +: Zr0 2Fe203 + 3ZrFL N hFeFé + 37Zr0 2Cr 0 + ZrF, - u4CrF, + Zr0 These metal fluorides are then available for reaction with chromium as shown above. It is therefore necessary that both the melt and the structural metal be of high purity. Even in this case, a possible cor;051on reac- tion is 2UF4 + Cr & CrF. + 2UF.. 2 3° Sampling systems designed to provide periodic analyses of salts “during corrosion tests are utilized in conjunction. with several forced- " convection loop experiments.llF Samples taken over a 20,000-hr period in loops operated under temperature conditions listed above show only slight increases in the chromium concentration of the salt during test. ~In an experiment containing the mixture LiF-BeF JUFh-ThFh (62=36.5=0,5=1 2 mole %) the chromium concentration increased from an initial level of 400 ppm to 500 ppm during the first 1000 hr of operation and remained at approximately the latter value during the remainder of the test.‘ Contamination of the Molten Fuel by Moisture or Air In the case of moisture contamination, the possible reactlons are: H 0 + OLiF - LiO0 + CHF 2 2 2HF + Cr ~» CrF, + Hy ZrF) + 210 - Zro,, + LLiF -86- These reactions are complete and very rapild, causing both corrosion of INOR-8 and precipitation of Zr0, in the fuel. | Contamination by oxygen of the air has a worse effect, although the reactions are not so fast as above: 0. + 2UF, -~ 2UOF, - - - - (strong oxidamt) 2 UOF, + Ni ~ NO 4+ UR, 2Ni0 + + ZrF, - Zr0, + Z2NiF | | NiF, + Cr - CrF2 + Ni+¢ 2 2 The last two reactions are relatively slow but'they cause nickel transfer from hot to cold regions. | The Zroziis not dense enough to settle and stays in the fuel as.a slurry. Although its particles are hard, erosion in the pump blades is not evident. A test program is in progress to evaluate the effects of contamina- tion on the corrosion behavior of fused fluoridé mixtures, and to ascertain the limits of the various contaminants which cen be tolerated without seriously increasing the corrosiveaness of the fuel salt. The contaminants under studj ificlude HF, metal oxides, and oil vapors. Results of this program will be applied to specifications of the cover-gas purity as well as to salt purity requirements. | | Corrosion by the Coclanfi - The coolant, being a mixture of LiF and BeF 09 does not present prob- lems of precipitation of U02 or Zr02. Although the coolant is very semsitive to moisture and air, the oxides are 150 times more scluble in the coolant than in the fuel. Some possible reactions are: 0 + 2Ni - 2Ni0o . ' ' _' $ . NiO + BeFé BeO ¥ + NiF, NiFé + Cr « CrF2 + TNi -87- This causes selective leaching of chromium, precipitation of BeO, and deposition of nickel. in the cold regions of -the loop, although the reaction NiF, + Cr « Nij+ CrF, is not as féfiperafiure,sensitive as the UFA oxidation reaction. , Since the probability of contamination in the coolant circuit is higher than in the fuel eircuit, caution must be taken to avoid undue corrosion. ~88- Appendix B HOT SPOT ANALYSIS An ana;ysis was made to estimate the temperature that the MSRE grephite and fuel may attain if one of the fuel passages becomes blocked so that mo flow occurs. The estimate is based on a camiparison between the temperatures resulting from a unidirectional heat-flow case and those given by a relaxatiqn-solution techniqué of twomdifiensiénal heat flow from one face to the other three faces of a square:rod.. The relexation solution is found to be 36%, or about one-third, of that for the uni- directional case. | The following cases are superposed: a. Temperature rise in an infinite slab of fuel with & uniform volume heat .source. b. Temperature drop for éonduction of this heat from the fuel across a graphite slab. ‘ c. Temperature drop caused by heat génerated.within the graphite. d. Film drop between the graphite wall and the bulk mean tempera- ture of the fuel in the adjacént open channel due to the heat inflow and its own volume heat source. The expansion of the fuel'réduges the power generation rate slightly. This adjustment is made at the end of the calculations. | The following conditions and properties are used in the calcula- tions: Reactor power 10 Mwt Fission-product decay heat outside core 4% Fraction of core power in graphite ) 7% Peak-to-average power ratio 2.8 Thermal conductivity of fuel (Kf) 2.75 Btu/hr-ft-°F Thermal conductivity of graphite (Kc) 12 Btu/hr-ft-°F Half fuel-channel width (xf) 0.20 in. or 0.0167 ft Craphite block thickness (xc) 1.60 in. or 0.1333 ft Bulk mean temperature of fuel in 1210°F adjacent channel Core size ,-89- Fraction of core volumefiin fuel - 0.225 Symbols | , q” Heat generation rate, Btu/hr-ft5 q” Heat flow rate, Btu/hr-ft | nmi' Temperature difference,'°F K Thermal conduct1v1ty, Btu/hr- ft-°F X Thlckness or distance, ft Subscripts f Fuel N bf Fuel in blocked channel ff Fuel in flowing ohannel c Carbon -or éranhifie A 7 Wall of flowing fuel channel W 4.5-ft diam, 5.5-ft height The temperature drop across a slab with uniform heat generation and cooling on one. side is: The bulk mean temperature of the fuel is: 2 AT = £ AT. m ) The tefiperature drop across & slab with Uniform heat flow is: /” ' Ar = X K (1) (2) (3) The film drop for laminar flow in flat chennels having uniform.heat generation is: /4 q The heat generated within a slab and flowing across the face is: X (17_ @+a” w/a” x) - 14) 35 (4) =90~ -qg} = qg; = 4.3 x 106 Btu./hr-ft5 5 V7 : 6 2 Q) = 0.095 x 10~ Btu/hr-ft° . For part (&): AIII - q X?‘ ‘= 1"03 X 106 X (00016112 = 220°F Féi péit (b): ‘. q” X : 6 S | AT = 0.36 x "t o 0i36 x 4.3 x 10" x 0.0167.x 0.133 _ 287°F. c S Kc . 12 4 For part (g): | | X 0.095 x 10° x (0.133)2 o ATC = 0.36 X ——2?{-:—- = 10.36 X 5 x 1o = | 25‘F_.- For part (d): qM,AX?f , 17(1 + q” ‘w/q”’ X ')wlh AT . - fIf , : £ff" £f Kf ’ [ 35 ‘] ) Id4 q, is taken to be one-third of the heat generated-ih one-half of the blocked fuel channel plus four-thirds of that generated in the graphite rod normally flowiiig through thé;fééé. V7 7z 7’ qT Xf + _l_l- q(.‘, X, W 3 3 Tk 6 6 o . %3 x10 x0.0167 , 0.095 x 10" x 0.133 24,100 + 4,200 = 28,300 . -0l - 6 x 0.0167 = 72,000 . Q” XKoo = k.3 x 10 1 +q)/a¥ Xpe =1 + 28,300/72,000 = 1.39 . AT '=.-_ ]'"'5 X 106 X 0.01672 <_‘_|_7 x 1.39 - 1k > ff 2.75 35 = Lho x 0.274 = 121°F The bulk mean temperature of the fuel above the wall is: .2 .2 . 1R Amm = 3 Ambf = 3 pd 220‘¢ 1&7 F With the fuel density changing -1.25 X lO_h/qF, héat generation in the blocked fuel channel is-‘about 93.5% of the base gsed in these calcula- tions. Making this adjustment, the temperatures become as shown: Temperature in the adjacent fuel channel, °F 1210 Temperature rise in fuel, °F - . _113 | Graphite.wall temperature, °F ; | | 1323 B Temperature rise in g;afihite, °F "f 4 : 292 Top graphite wall temfiérature, °F | - 1615 Temperature rise in fuel channel, °F ‘ 206 Peak fuel temperature, °F | | - agel These temperatures are attainable only in the center of the core and should not damagefeither'theffuel or the graphite. _92 - " Appendix C Number JS-80-123 Date 1/24/61 Revised Page 1l of 11 JOB SPECIFICATION REACTOR DIVISION OAK RIDGE NATIONAL IABORATORY UNION CARBIDE NUCLEAR COMPANY Division of Union Carbide Corporation ORNL, Oak Ridge, Tennessee Subject: Specification for Primary Drain and Fill Tanks, Primary Flush Tank, Secondary Drain Tank and Fuel Storage Tank for Molten Salt Reactor Experiment am— — — = — — 1. SCOPE This specification covers the requirements for materials, fabrication, inspection and testing of the primary drain and fill tanks, primary flush tank, secondary drain tank and fuel storage tank for the Molten Salt Reactor Experiment. : 2. APPLICABLE SPECIFICATIONS) CODES, DRAWINGS, AND QTHER PUBLICATIONS 2.1 The latest revisions of the following dccuments shall form e part of this specification to the exvent stated in subsequent sections ASME Boiler and Pressure Vessel Code, Sections VIIT and IX ASME Code Case Interpretations 1270N-2 and 1273N;3 ORNL Specification MET-RM- 4 for INOR-8 Welding Fittings, Shapes, etc ; ORNL;Spec1fication MET-RM-B163 for INOR-8 Tubing ORNL Specification MET-RM-B167 for INOR-8 Pipe ORNL Specification MET -RM-B304 for INOR-8 Weld Filler Material ORNL Specification MET-RM-B334k for INOR-8 Plate ORNLlSpecifIcation MET-RM-2 for INOR-8 Forgings ORNL Specification MET-NDT-E165 for Liquld Penetrant Inspection ORNL Specification MET-WR-2 for INOR-8 Welding Requirements - 93.. Number JS-80-123 Date 1/24/61 Revised Page 2 of 11 ‘ORNL Specifications P.S.-23, P.S.=25, P. S.-26 for INOR-8 Welding Procedures (for information only) ORNL Specifications QI'S-23, QTS-25, QTS-26 for Welder Qualification Tests (for information only) 2.2, Drawings The following Company's fabrication drawings form a part of thiS“ specification - Assembly D-FF-.-Ahoh'j'j Primary Drain and Fill Tank DéFFgAhOh56 Primery Drain and Fill Tank - Steam Dome - Assembly and Details D-FF-A4OUST Primary Drain and Fill Tank - Assembly and Details D-FF-A4O458 Primary Drain and Fill Tank Bayonet Exchanger ~ Assembly and Details ' - : D-FF-ALOL59 Primary Drain and Fill Tank Bayonet Exchanger Bracing ' D-FF-AUOLEO Primary Drain and Fill Tank - Cooling Water Header Assembly and Details D-FF-AkOLEL Secondary Drain Tank - Assembly and Details D-FF-A4OL62 Primary Flush Tank and Fuel Storage Tank - ; Assembly and Details 3. REQUIREMENTS 3.1 Design The tanks to be furnished under this specification shall be fabricated in accordance with the Company‘'s designs, and as shown on the Company's fabrication drawings accompanying and forming a part of this specification, except that the Seller may adopt his own standards for weld-end preparation, provided they are submitted to and approved by the Company in writing prior to start of fabrication. The tanks will contain molten fluoride. salts. ’ ' ' 3.1.1 3.1.2, Number JS-80-123 Date 1/24/61 Revilsed Page 3 of 11 Primary Drain and Fill Tanks Two (2) primary system drain and £ill tanks, similar in. design, with the exception of certain nozzle locations, will be required. Each fill and drain tank consists of a lower tank, which will contain the molten fluoride salt, and an ‘upper steam dome. Heat will be removed from the molten salt in the lower tank by introducing water into concentric thimbles - which penetrate the top head of the lower tank. The steam generated in the thimbles will be collected in the upper steam dome through tubing and flexible connectors. Details of the tank support designs bhave not been finalized. The Seller shall be responsible for furnishing tank support brackets and steam dome supporting steel after designs are completed and drawings furnished to the Seller. The following design criteria are included for the Seller's information: Tank désign pressure 50 psig Tank design temperature '1300°F Cooling system capability 100 kw at 1300°F Cooling water temperature 100°F Steam dome design pressure | 50 psig Steam dome design temperature 300°F Primary Flush Tank One (1) primary system flush tank will be required. This tank will contain the molten. salt necessary for flushing the primary piping system after the radiocactive salts have been drained from the.systen. ' Design conditions are as follows: Design temperature 13009F Design pressure . 50 psig 3.1.3 3.1.4 -95 Number Js-80-123 Date 1/24/61 Reviged ' Page 4 of 11 Secondary Drain Tank ‘One (1) secondary -drain tank will be required to contain the molten salt drained from the secondary piping system. "Design conditions are as follows: Design temperature | : 13009F Design pressure _ 50 psig Fuel Storage Tank one (1) fuel storage tank will be required. Design conditions are as Tollows: Design temperature 1300°F Design pressure | 50 psig 3.2 Materials " 3.2.1 All tanks, tubing, piping, etc., to be furnished under this Temp. Op specification shall be fabricated,from INOR-8, a nickel- molybdenum-chromium material, unless otherwlse specified in the Parts Lists which appear on the Company's fabrication drawings. ‘ | Design data for INOR-8 are given in the following table: Modulus of Mean Coeff. Allowable Elagticity - of Expansion Thermal - Stress psi ' in./in. OFx10~6 Conductivity psi 10° psi TO%F 0 T _ Btu/ft2-hr-OF /£t 24,000 314 1T 24,000 30.7 | 8.1 22,800 30.1 21,700 29.5 6.45 9.0 20, 800 28.9 ' 20,000 28,5 6.76 9.9 19,300 28.0 Number JS=-80-123 Date 1/25/61 Revised Page 5 of 11 Modulus of Mean Coeff. . Allowable Elasticity of Expansion Thermal Temp. °F Stress - psi ' in./in. Opx10-6 Conductivity T psi 100 psi 0% to T Btu/ft2-hr-SF/ft 800 - 18,700 37.7 - 7.09 | | 10.8 900 18,150 27.2 1000 16,000 | 26.8 o T.43 11.7 1050 13,250 g6,6 1100 9,600 26.4 1150 6,800 - 26.3 1200° 4,950 26.1 7.81 12.6 1250° 3,600 25.8 1300 2,750 25 .4 1350 . 2,0% ah.g | | | 100~ 1,600 2L, Y | - 8:16 | ‘13.5 Density, 0.3L7 lb/in3 Specific heat, 0.095 Btu/1lb °F 3 3.2.2 All material received by the Seller for use in fabrication . .of the:equipment to be furnished under. this specification shall be inspected by the Seller for damage during ship- ment. All material found to be defective shall be- rejected and reported to the Company. The methods of inspection shall be approved in writing by the Company. 3.3 Fabrication 3 3 l All tanks shall be fabricated in accordance with the applicable sections of the ASME Boller and Pressure Vessel Code, Sections VIIT and IX, including Code Case Interpretations 12TON-2 and 1273N-3 for primary nuclear vessels, except that materials of construction and allowable stresses shall Number JS5-80-123 Date 1/24/61 Revised ‘ Page = 6 of 1L be in accordance with this specification. In addition, the supplementary requirements of this specifieation'Shall be met. Where conflicts or inconsistencies occur, the requirements of this specification shall govern. Code stamping will not be required. . 3.3.2 .Forming of INOR-8 Materials All procedures to be used by the Seller for forming of INOR-8 materials shall be submitted in writing for the approval of the Company prior to start of fabrication.. All materials after‘belng formed.by any method ‘such as bending, drawing, or swaging shall be subaected to liquid penetrant inspection of all surfaces in accordance with . specification No.. MET-NDT-E165. Any type of crack, fissure, fold, or other injurious defect shall be cause for re jection of the part. Removal or repair of injurious defects shall be permitted only after written approval of the Company. - - Formed heads shall be made in accordance with ORNL spe01fi- cation No. MET-RM-6, . 3.3.3 Heat Treatment The Seller shall furnish the Company with written heat -treatment procedures which he proposes to adopt during the fabrication of the tanks. These procedures shall be ~ approved In writing by the Company prior to start of tank fabrication. A record of each heat treatment shall be made and shall form a: part of the fabrication and inspection report. 3.3.4 Welding All welding of INOR-8 material shall conform to the Com-- pany's Welding Specification MET-WR-2, attached hereto. Welding of INOR-8 material may be performed in accordance with the Company's Procedure Specifications PS- 23, PS-25, and PS-26. at the Seller's discretion, however, the Company assumes no responsibility for the adequacy of these speci- fications to meet the requirement of this specification. All welding procedures used by the Seller in weldingiINOR-B materials shall be submitted to and approved in writing by the Company prior to start of faebrication. -The procedures 3.3.5 3.3.6 3.3.7 Number JS-80-123 Date 1/24/61 Revised . Page T _of 11 and welders shall ‘be- qualified in accordance with ORNL Specification MET-WR-E Cleaning Tmmediately following any operation that imposes any unclean condition and before assenbly, all parts and subassenmblies shall be cleaned free of all oxides, grease, oil, filings, dust or any other foreign material. All internal and external surfaces shgll have a bright finish. Precuations shall be taken to insure that all parts and subassemblies are kept in a clean conditiom throughout fabrlcat ot of tce subject tanks. It is especilally important that oxide scale shall be removed from all parts that camnot be reached for direct ‘inspection and cleaning after assembly. Discoloration of su*glcal gauze after wiping metal surfaces shall be used as a check for cleanliness;7 Compounds containing sulphur, lead or mercury shall not be permitted to come into contact with uurfacem of INOR~8 material. Workmanship The Seller:shall fabricate all tanks and appurtenances in a menner consistent with the standards of high gquality workmanship. Inferior quality of workmanskip, azs determined by visual inspection by the Compeny's inspector, shall.bé couse for rejection of‘the work . The quality of workmanship as approved in the Welder's qualifi- cation tests shall be maintzined throughout performance of all work on the subject tanks. Inferior workmanship as determined by testing and inspection of the welds in accordance with this specification shall be cause for rejection of the work and re- gqualification of the welder. Identification The Seller shall affix an INOR-8 nameplate to the outside shell of each tank. The following data shall be stamped or engraved on each nameplate: Fabricator's Name Specification JS=80-123 Year Completed Design Pressure Design Temperature Hydrostatic Test Pressure A nameplate giving similar data shell also be affixed to the steam dome of each primary drain and fill tank. -99- o Number JS-80-123 Date’ 1/24/61 'Revised ‘Page 6 of 11 . INSPECTION AND TEST REQUIREMENTS L.1 h.2 4.3 L.b Tn addition to welds, each forging shall be- liquid penetrant inspected in accordance with ORNL Spec1f1cation MET-NDT-E165 'Inspection of welds shall be in accordance with ORNL Welding _ Specification MET-WR 2. The Seller shall arrange for the Company's representative to - have access to such parts of all plants as are concerned with the supply, manufacture, and assembly of parts for the subject tanks when requested, including Seller's own plants and those of his suppliers.. Where reference is made to the Purchaser in the ORNL specifications for INOR-8, it shall be interpreted to mean Company . The Seller shall notify the Company at least three (3) working days in advance of . the start of tests and inspections so the Company representative may be present. The tests and inspections | _referred to include: L.5 h.6 (1) Welding procedur'e qualification. (3.3.4) (2) Welder qualifications. (3.3.k) | (3) Any repairs of defects. (4.5) (4) Hydrostatic test. (L4.6) | -(5)‘ Leak tests.' (4.7, 4.8) (6) Any other tests. (4.9) (T) - Preparation for shipment. (5) '"No waiver of inspection observation or any requirement of this spec1fication will be made unless confirmed in writing by the Company . Repalrs necessitated by defects in material or workmanship shall not be made without full knowledge and approval of the Company. The tanks shall be subjected to hydrostatic tests as follows: The lower tank sections of the primary drain and f£ill tanks shall be subjected'to a hydrostatic test of 655 psig. The secondary drain tank, the primary system flush tank and the fuel storage tank shall each be subjected to a hydrostatic test of 655 psig. L7 L.8 h.9 k.10 =100- Number JS-80-123 Date . 1/24/61 Revised - Page 9 of 11 The primary drain and fill tank steam dome and interconnecting piping to and including the tank thimbles shall be given a hydrostatic test of 80 psig. The hydrostatic test pressures shall be held for one hour, during vbich time all surfaces and joints shall be visibly inspected for leéaks. Repairs shall be made only after notifying end receiving the approval of the Company. After the hydrostatic test, a helium mass spectrometer leak test shall be applied to the lower tank sections of the primary drain and f£ill tenks, and to the primary flush tank, fuel storage tank and secondary drain tank. The tanks shall be tested for leakage to the inside by bagging each tank in a plastic bag filled with helium and evacusting the tank. Each tank shall be tested separately. The leak detector shall be demonstrated under test conditions ~to be sensitive to 1 x 10-8 §TD cc/sec of helium when using 8 standard leak of 1 x 10~9 STD cc/sec connected to the most remote portion of the tank. Indicated 1§qkage‘into each tank under test shall not be greater than 1 x 10-° STD ce/sez. The leak test shall be run for a minimum of 30 minutes on each tank,or for sufficient time to detect the standard lesk, whichever 1s greater. If, in the cpinion of the Company, the background reading of the leak detector has changed sufficiently during the above test to create a doubt regfirding the absolute lesk-tigbtness of the tanks, the test shall he repeated. : The presence of the Company representative is required during all helium leak testing. After the hydrostatic test a halogen leak detection test shall be applied to the upper tank sections, thimbles and inter-connecting tubing to the primary drain and fill tanks. The volumes to be tested shall be pressurized with a mixture of st least 25% freon in air, and the exterior surfaces surveyed with a halogen leak detector, using a technique demonstrated under test conditions, to be capable of detecting a lesk of 1 x 102 STD ce/sec of the pressurizing mixture. Leaks giving indications greater than 1 x 1072 STD cc/sec shall be cause for rejectlon. ' The Seller shail make and report such other tests and inspections as are necessary to satisfy himself of the integrity and good condition of the tanks. - E If at any stage of teéting or inspection, physilcal failfire, deformation or mechanical damage occurs or is observed, it shall be deemed as failure of the tank to meet these specifications. (- 5. 6. .11 <101~ o Number JS-80-123 Date 1/24/61 Revised o Page 10 of 11 A test fallure which requires repair and/or corrective measures to be made will automatically necessitate repetition of the - previous inspections and/or tests on the part requiring repair. PREPARATION FOR DELIVERY 5.1 NOTES - 6.1 Preservation and Packaging Prior to shipment, all tank openings shall be closed with gas-tight closures and the tanks shall be evacuated and charged o to 25 psig with welding quality helium, argon or dry nitrogen gas. A valved pressure gauge shall be furnished and shall be securely affixed to one of the closures in each tank in such a manner as to permit reading of the internal gas pressure. Each tank shall be securely mounted on a skid and suitably blocked and strapped to prevent shifting and/or damage while in transit. Engineering Information 6 1.1 Design Approval Within five weeks after receipt of order, “the Seller shall furnish the following Engineering Information for approval by the Company " No. of Copies ‘ Descrlption T Leak Test Procedures . T o Welding Procedures .7 ‘Heafi Treatment Proceoures T Record of Welder s Qualification Tests T INOR 8 Inventory System After receipt of Engineering Information; the Company will - require a minimum period of four (4) working days for its review. TFabrication of work shall not be started and de- livery of equipment and material shall not be authorized until written approval of the Engineering Information has been extended by. the Company. : 6.2 6.3 6.4 =102 - . , Number JS-80-123 Date '~ _1/24/61 Revised = Page I of 11 The Seller shall furnish four (h) certified coples of the Engineering Informetion within four (4) weeks after receipt of approval. Seller's Data Certified éopies of Seller's Data shall be furnished in the | quantities specified prior to shipment of equipment. . No. of . Copies ' Description 1 Radiographs 1 Heat Treating Charts 'h Leak Test Reports A Liquid Penetrant Test Reports L Weld Identification b INOR-8 Inventory Report Y Supplementary Shop Dr&wings L - ASME Form U-1 Manufacturing, Inspection and Test Schedule Within four (4) weeks after receipt of order, the Seller shall submit to the Company the manufacturing, inspection and test schedule for all material and equlpment furnished under this specification. TNOR-8 Material Inventory Throughout fabrication of the tanks, the Seller shall maintain a system of materlal identification and control sufficient to establish the complete identity and history of all INOR-8 material, including weld filler material. The system shall be approved in writing by the - Company prior Yo shipment of material to the Seller. 4 7N -103- - Appendix D COMPONENT DEVELOPMENT PROGRAM IN SUPPORT. OF THE MSRE The reliasble performance of components;and-auxiliaries used in the circulation ofimoltenwsalts has been established in;oyer QOO,QQO hr of.. accumulated loop operations, and forms. the basis,forAtne specification of components for the MSRE. | ; e ' As an added insurance of reliability and safety, prototypes of critical MSRE- components will be operated out—of-pile under conditions resembling‘those of the reactor. Facilities to be used for this testing include salt systems of various sizes and different degrees‘of'complex- ity and model tests in which hydraulic and mechanical processes are :analyzed Core Flow A1/5- scale plastic model of the reactor has been operated with : .water as the fluid. Fluid velocity distribution in the entrance plenum in the region next to the cylindrical wall and in the lower plenum has been experimentally determined to be satisfactory_for purposes of B cooling those regions..\On the basis of measurements made in'tfie 1/5- scale model, the des1gns of the MSRE core and of a full-scale model were establiShed The full scale model. Wlll be operated w1th 1250 gpm of glycerol solution to reproduce the expected Reynolds nuflber of the reactor salt. .The adequacy of flow distribution in every portion of the core-vessel assembly will be proved or required modifications will be devised and demonstrated The possible interactions of fuel salt and full-size core graphite stringers will be investigated in an B-in -diameter vessel, part of the Engineering Test Loop (a facility for testing many MSRE prototype components .and 0perating procedures) =10k~ Fuel Circulation Pump The conceptual design of the MSRE pump is similar to that of the - pump developed for the Alrcraft Reactor Experiment. These pumps have beén virtually trouble—free in out-of—pile use. rThe prototype will be tested with vater and with molten salt. - Hydraulic performance will be investigated in the water tests, later the design of “the experiment will be modified as reqnired for the stripping of dissolved gases in the pump bowl. rThe'investigation of gas kinetics in.a pump bowl has been started with COo being used as a tracer. B . Hot tests with molten salt will provide-a{final test of the-pump geometry derived from the water tests. Krypton-85 tracer will be used to develop an-effective means of purging fission gases from the pump and of excludirng them from the motor. The pump support arrangement designed for the MSRE will be utilized in the bot test stand. The MSRE pumps will be tested at operating temperstures in this equipment prior to installation at the reactor. ‘ ‘ Freeze Flanges Freeze flanges have been & part of moltennsalt engineering devel— opment for several years. They have been found satisfactory in a large circulating salt system, the Remote-Maintenance Demonstration Facility. At was shown that the flanges could withstand repeated thermal cycling and could be broken and reassembled. ' . Two flange pairs were cycled between room temperature and 1300 F for more than lOQ_times without failure. Salt leakage from a freeze flange has not been experienced to date, slthough leaksge of helium through the secondary gas buffer seal is not uncommon. ' The design of the flanges has been improved to provide grester strength toward axial loads, to reduce the thermal stress, and to improve the tightness of the buffer seal. Flanges of the improved . type will be tested under simulated reactor service conditions to establish their reliability. -105- Heat Exchangers Heat exchangers as complex as the MSRE exchanger were fabricated of . Inconel for the ARE and the ART. The fabricability of INOR-8 inmto tubes, and tube-tubesheet assemblies has been demonstrated.. The MSRE heat ex- changer and radiator assemblies, ee now designed, do not require further development. Freeze Valves Plugs of frozen salt havegbeen used in many loops to isolate circu- lating molten salt from the environment. Two prototypes of "valves" to be used in the MSRE drain lines have been frozen and thawed 100 times without damage or incident. . Ability of the valve to fall safe on loss of heating or 1oss of cooling was demonstrated. | | Testing of prototype Preeze valves will be continued as part of the operation of other loop and component tests. Sampler-Enricher Part of the sampler-enricher mechanism has been mocked up, and mechanical components and instruments are being added to the mockup as . the deSign evolves. The basic mechanical parts thus far are functioning reliably. | o ‘ | A complete sampler-enricher mockup w1ll be fabricated when draw1ngs i of the MSRE device are finished. After the mechanical debugging of - this prototype,‘its ability to obtain representative samples without - contaminating the system with oxygen, or the environment with activity, will be demonstrated on the Engineering Test Loop Heaters Prototype MSRE heaters for pipes and vessels are being fabricated for testing on mocknps and loops. .The heaters will be sflbjected to life tests, and their temperature distribution and heat loss will be measured. Possible_damage to pipe walls by overheating will be investigated also. ~106- Poison Tube ‘A mockup of the MSRE poison tube will be fabricated and tested to establish its reliability, response times, and control characteristics. Gas-Handling System Helium cleanup traps, based on the technology developed for.gas- cooled reactors, will be tested under flow conditions. At the same time, methods of chemical analysis to detect small quantlties of contaminating oxygen will be developed and demonstrated Maintenance The practicality of the two general methods of maintenance designed into the MSRE has been thoroughly demonstrated during the past 3 years. Entirely remote maintenance, with the aid of stereo-television and a General Mills manipulator; was demonstrated on & special salt system (Remote-Maintenance Demonstration Facility) of about the size and degree | of,complexity'of the MSRE. After this facility was operated with molten salt, the pump, the dummy core, the heat exchanger, and heateérs were removed and replaced. ,The:system was shown to be operable following the replacements. : f ' Semidirect maintenance with long-handled tools operated through ,portable shielding has been used successfully on a number of operations at Hbmogeneons Reactor Experiment No. 2, a reactor of the general size and activity level of the MSRE. BSome of the operations performed with these techniques were repair of the core vessel and replacement of pumps, valves, ‘and a filter. Although the feasibility of maintenance is regarded as having been established, additional development and practice is required to work out detailed procedures. Specific maintenance problems are being solved with the use of- appropriate mockups in the Remote=Maintenance Demon- . stration Facility. As an aid to the designer of maintenance procedures, - a 1/12-scale model of the reactor system and maintenance area is being built. This model will provide insurance that every item in the MSRE can be repaired or replaced after operations have started. _107- Appendix E 'ACCIDENTS INVOLVING RELEASE OF MOLTEN SALT INSIDE THE REACTOR CELL Several incidents involving the release of the hot molten fuel salt and coolant salt into the reactor cell were studied. The first postu- lated accident involved an instantaneous drop of the molten salt'into the reactor cell. The second accident 1s a rapid dispersion into the cell, such that the air within is quickly heated. The third accident is similar to the first except that the molten salt falls into water in the - bottom of the cell. o i S Many simplifying assumptions are made, but all are conservative; therefore the calculated. results of these accidents are more severe than would be experienced in an actual accident. o The calculations for the. three postulated accidents 1ndicate that the third, invwhich water leaked into the reactor cell, is the most severe. However, an automatic sump pump ‘removes all inleakage of water. Furthermore, . any presence‘of7water is immediately alarmed to the operator, who may then tske corrective action if the pump fails to work. Air Only in Cell Let it be assumed that- both fuel and coolant systems rupture and all fuel and coolant fall into the bottom of the reactor cell For sim- plicity in calculations it is assumed that the bottom ‘of - ‘the reactor cell is flat. The follOWing parameters are used in the- analySis of such an accident: . : a. The total volume 6f the fuel and coolant mixture is 85 £, b. The mixture covers'the bottom of the cell ih a layer 0.8 ft thick. , . ¢. The temperature ofgthe mixture is;l500°F iflitially. ] -d. The temperature of the air in the cell is*100°F initially. e. The volume of the reactor cell is 11, 55O'ft3. i The total heat transfer coefficient is- calculated to be = 6.3 Btu/hr-°F-£t°. =-108- The produced decay heat, q(t),'is chosen to be 7% of total power at 1l sec after shutdown. Q o q@ (L sec) = (0.07)(20 Mw)(3.%15 x 10 Q'O it 2.39 x 10° Btu/hr . The decay heat per unit surface 6 q - 2P x 20 Btu/hr , 107 ft 2.2k Btu/hr--ft2 o ] The Way-Wigner equation for decay heat is: . -002 a(t) = q t . Writing (5) with t in hours and assuming t = O at 1 sec after shutdown, a(t) = q (3600 t + 1)'0'2 . EEEPR - ~0.2 a(t) = g (3600 €)% . The parficular equation for this system is = q(t) = hte 2 where 6 = teyperature‘in °F and ¢ QITcppfl- , (0.468)(133)(0.8) , - 49.8 Btu./°F-ft2 . 6 Btu/Mw-hr) , (1) (2) (3) (4) (5) (6) (7) (8) -109- 'The Laplacian transformation of Eg. (8) is: o) = —2 L) —2 . (9) s + (ht/c) C s + (ht/C) AR | Let. M =hfc = 0.126/nr Returning to the time domain, and‘using the convélution integral, B o) = 6™ + 2 [q(p) e"“"'] , | (10) t - - é)ie"')Lt + ;=./fi a(t) é-(t’T)k ar , £ J | o S 4.35 x 10° f 0.2 e-_xt"a (11) - i 49.8 i o ) | Yo Equation (11) is not readily integrablej. therefore let‘i%rbe approxifiatédiby'. | B | 6(t) = e v [ 6, + 87 .k Z ti-o-.e e At:} L (12) L .=1 The surface temperature of the molten salt is-initially at 1300°F, and the)shield water outside the reactor cell is initially at lOO°F. If it is assumed that after,thé accidental falling of the fgel and coolant these températures remain cdnstant, a meafilsteady-staté air fempefature may be compfited as follows. The heat flow out of the'salt t0 the_air in the steady-state condition equals the heat flow from the air through the steel shell to the water. This tacitly assumes that thé‘salt is an in- finite source whilé the water is an infinite sink. In equation form° the heat flow is written as follows: hsAS(tS - ta) At = Subscripts where t p : a hAt +hAt 8 8- 8 W WW -110:- (1k) thermal, conductivity, Btu/ft2-°F-hr , area, ft2 R | temperature differential . 'thw(ta -t) . (15) s, a, and v refer to salt, air, and water, respectively. R h A + h A g s W W i h s ja g ] Substitute 214°F . (16) 2 Btu/ft°-°F-hr , 0.86 , 107 £t° 2 2370 £t° 1300°F , 100°F . in (16): (17) Initially the temperature of the air = 100°F, P = 14.2 pSia.} The steady-state préssfire of the air within the reactor cell is: 560 (1b.2)(674) | | (18) 17.1 psia |, - | - (19) 2.4 psig. | - (20) =111~ Sudden and Minute Dispersion .of Fuel and Coolant Salt Within the Reactor Cell Let it be assumed thatpthe fuel and coolant circuits are ruptured resulting in a very sudden and minute dispersion‘of thé molten salt into the reactor cell.fl Let it also bevassumed that the molten salt and air within the cell constitute an insulated system and that the air then attains thermal equilibrium with the salt. ' In the steady state, the heat loss of the molten salt equals the heat gained by the air. ' : | (Tfuel B Tfinal) Ve cpf = (Tfinal B To) Ya cpa A (21) where Tfuel temperature of fuel = 1300°F ,,.j Tfinal = final steady state of fuel and glr s T0 = initial temperature of air = 100°F , W - weight of molten salt = (85) £t (133) 1b/ft’ = 11,320 1b cp' = specific heat of molten salt = 0.468 Btu/1b-°F ¢ : _ . o : | w, = weight of air = (2.378 x 10'5)(52.17)(11,550) = 884.0 1b, e, = specific heat of air = 0.24 Btu/lb . a A Substitute in (21): (1300 - Tfiqa;Q(ll,ieo)(0-468) = (Tpypgy - 100)(88)(0.24) , (22) 6,890,000 - 5300 Ty oo 212 Tpy oy - 21,200 Teinal = ©,911,200/5512 1250°F . | | | (23) <112 - The idéalized gas relationship is: - - > R~ where - Pi = initial pressure in vessel = 1h.2 psia at 100°F , T, = 100°F = S559°R ., . T2 = 1250°F = 1709°R ., .. Substitute in (3): 5 _ (1k.2)(1709) - T 2l P 559 N P ( ) = L43.4 psia o * (25) 28.7 psig . ' - (26) Assumé now that, after a steady-étate condition is reached within the pressfire vessel, the drain-tank cell volume is suddenly added to the reactor-cell volume. Initially, the reactor-cell steady-state pressure is P, = 28.7 psig, and T = 1250°F, and the drain-tank cell is at P, = 2 3 14.2 psig and T3 ‘= 100°F. -7/ ’ -/ = (M s . NN (TE Tfinal) ?f Cpf M (TQ J Tfinal) Ya cpa (Tfinal _15)_wae cpa;’(ET) (T2 Tfinal) (Wf.cp t Wy Cp ) (Tfinal_ T5) Ve Sp (28) , e £ ! 2 “a where .?Tgr = 1250°F , Cwe = 11,320 b, c. = 0.L68 , Pp . W = 884 1b , -113= T, = 100°F , w = (2378 x 157)(32.17)(7010) = 5% 2 ‘ : c = 0.24 Btu/lb Pa " Substitute in (8): ‘ (1250 - T% nal)[(11,320)(0 468) + (88&)(0 eui]= (Tfifiél - 100)(536)(0 2k) , | (29) (1250 -1, 0 )(5502) = (Tfi g1 - 100)(128.6) , . _ 6,880,000 + 12,860 - 6,892,860 . _ . ;ep Teinal =~ 565 _' S5 - R T < TOVF , A R - (0) Ty Lo : L ' . .2 ;Ei_' 559 ~ 1223 ’ | 'P3 = Blfl ésia , - 16.4 psig . | - (31) Water in the Reactor Cell In the third postulated accident, the ffiel system was considered ruptured and molten salt released into water already in the bottom of the containment shell, producing steam and increasing the pressure. The worst situation was determined to be thé“%fifiture'of the fuel line between the heat exchanger and the thermal shield, which would discharge about - 3 3 ft© of fuel in 1l sec, under pump pressure, with another 2k ft3 draining v -11k- out in 60 sec. Ruptures withln the thermal, shleld or in the drain line would supply more salt but at a slower rate Likew1se, a rupture of a 3 tube in the heat exchanger would allow most ;of the 30 £t~ of coolant salt to drain, but again at a slower rate, for the addltlonal salt. The heat d1ss1pat10n rate of the contalnment-vessel walls when condensing steam is 1.65 x; 106 Btu/mln, whrgh is the equivalent of a - hot-salt leak rate of 21 ft3/m1n Thus,’after convective'circulation is establlshed within the reactor cell, condensatlon beglns and the | pressure within the contalnment cell w1ll begln to decrease For. the calculations, it is assumed that there w1ll be no heat loss from the vessel for the first 350 sec, -and that it w1ll increase 11nearly from 0 to 1.65 x 106 Btu/mln 1n the next 30 sec, . . | | The pressure rise is. calculated on the,assumption of" the optlmum amount of water being present The max1mum pressure results when thére is Jjust sufficient water to be converted ; to saturated steam, surplus water absorbs heat, and superheating the steam produces less increase in pressure than the same; heat put into evaporatlng water Thus, if ' more water is present than” the optimum calculated below, ‘the initi&l" pressure rise would be less than calculated and the contlnued dralning of the salt would' evaporate more water, but the pressure would be® less than the peask of the initial pressure surge. L - The following is a table of symbols and definitions of the variables and fixed parameters: P, -Initial temperature of fuel = 1225°F. 1 T, Initial temperature of ‘water and air. lOO i Te Final temperature oquuel, vater, and‘alr, op p - Density of fuel = 154.5 1b/ft3 W, VWelght of fuel released 4170 1b f - . wa‘. Welght of air in the reactor contalnment shell = 834 1b L e | : wa‘ Weight of'air in drain-tank containment volume = 579 1b ‘o 1 ] _lls - v.OW. o+ W = total weight of air, 1b a a a 1 2 L Welght of water, 1b e Average specific heat of fuel over the temperature range £ | | = 0.544 Btu/1b-°F c, Average specific heat of water = 1.0 Btu/1b-°F | | W : ' e, Average specific heat of air = 0.24 Btu/1b-°F 8 o Mww - Number of moles of steam Mwa Number of moles of air hg' Enthalpy of steam at T, Btu/1b Pa Partial pressure of air PS | Partial'pressure of HéO The energy balance after 1 min is: (T, -32) w.c_ + (T -32)(w c +w c. ) L. £ pf} °© w pw 1 Py = (Tf 52) (v, e, Wy © ) +w_h_ - 0.41 x 10 (32) | £ "1 Pa The sum of partial pressures is: (Tf + 460) : P = P T+ 1%0) + P | o (33) also, =116~ P - P (34) Subspitutipg known values and simplifying, 2.39 x 100 - T, (2468) | o WW‘ = - 68 ) I . | . (35) g P = :"0}025561(Tf + 460) + P, o ” | (36) P = P [1+Ww ] . . o (37) The solutiohs to Hs. (35), (36), and’ (37) show the maximum pressure to be 39 psig, the temperature to be 260°F, and the optimum emount of water to be 1590 1b. | [ 3] PrYg W, -11%- ~ Appendix F GRAPHITE COMPATIBILITY WITH SALT Assessment of the general question of.compatibility of graphite with molten fluoride reactor fuels has required experimental study of several possible problems. The present status of this'experimentai.program is descrihed briefly in this appendix. ;Chemical Interactions Intercalation compounds. of graphite with a variety of . -pure chlorides, _bromides, and lodides are known. Graphite 1s severely damaged when such compounds .are formed; a 2-hr treatment with FeCl, at 300°C, for example, 3 reduces graphite to powder by formation of an intercalate.- , - The maJor constituents, 1iF, BeF Zth, h’ h’ of the MSRE fuel mixtures are known from a large number of experiments to form ne such com- pounds with: graphite., This situation is not changed when such materials as NiFg, FeFe, and CrF2, rides are added in apprec1able concentrations. . T alkaline earth fluorides, -and rare-earth fluo- A few of the possible- fission-product fluorides (MoF5, for example) might in the pure state, form intercalation compounds with.graphite. ,However, intercalation byicompounds which react readiiy when pure can be .firevented by dilution with nonreactive salts. The possible'intercalate formers among the fission-product fluorides will occur only.at concentra- tions belowflQ.Ol mole %f“ifi appears rery unlikely"that,compound formation can occur between the graphites and any constituent of the.molten fluoride solution. R | » " Some cesium and rubidium isotopes of a variety of half-lives will be formed in ‘the graphite through decay ‘of xenon and krypton isotopes which have diffused into the graphite. ' The moderator temperatures are such that some stability of such compounds’ as CSCEh or CsCB'must be ‘expected. Such compounds will tend to disrupt the graphite structure and would, if present in sufficient concentratiohs, probably disintegrate the moderator blocks. ' The' absolute amounts of these elements so-introduced into the graphite are 'so small that their efféct can hardly be important. -118- Chemical reaction of the fuel salts with some of the contaminants which desorb from the graphite must be expected. Mixtures containing 1iF, BeF ZrFu, ThFu, and UFh in concentrations typical of the MSRE 2 fuel redgt slowly if at all with CO,, CO, and 055 though reactions such as 002 + Fe - FeO + CO and Zth + 2Fe0 Ha'”2FeF2 + ZrO2 would introduce some oxide contamination into the fuel salt. Reaction of the salt mixture with HEO as by . 2 is quite rapid. It will probably prove impossible to remove all chemi- 2 h ) . ZrFu + 200 - LWHF + Zr0 'sorbed oxygen-bearing species from the moderator graphite before addi- tion of the MSRE fuel. However, the MSRE fuel mixture can apparently accommodate up to 1300 ppm of O in solution without precipitation of a solid oxide; this is more by a factor of 3 than that available (assuming complete desorption and reaction) from the graphite. Moreover, if the ZrFu/UFu ratio in the MSRE mixture exceeds 2, the first fiaterial.to pre- , which contains nmo UO,. Since the Zth/UFu ratio ifi?the MSRE fuel will be at least 5;Ithere seems to be no fear of. precipitation of U’O2 from the circulating fuel. cipate is triclinic -ZrO Permeation of Graphite by Fuel Mixture Graphite is apparently wefted by some fluorides in fhemolten _ state. Treatment of graphite with molten SnF2 at 300°C and at atmos- pheric pressure, for example, results in virtually complete penetration of the speéimen by the salt and in a continuous film of salt over the graphite surface. | - : The results of a considerable series ‘of experiments indicate, how- ever, that graphife is not wetted by mol£en fluoride mixtures containing LiF, BeFé;‘Zth, ThFu and UFh“ This behavior is not changed by addition of minor constituents suchas the fluprides:of structural metals. This 'behaviéf is essentially unchanged on treatment of the molten salt mixture +ll9; with anhydrous HF,or'stronglreducing agents such as Zr°. Superficial evidence of wetting is obtained if the graphite;salt system at‘700°q is exposed to the air; in that case the screen which forms on the salt ap- pears QO'promote wetting of the graphite surfaces. ©Since the graphite - is not wetted by the salt mixture under nonoxidizing conditiohs, the graphite tends to resist?pehetraticn by the fuel salt into its pore structure. : | ) ' o o ' Moderator graphite is, however, of much less than theoretical density. About 10% of the volume of a specimen of moderator graphite consists of' a network of interconnecting capillary pessages of a variety of sizes. “Even though the graphite is not wetted by ‘the salt, therefore, the salt can be forced into the graphite pores by application of external pressure. - The amount ofrfienetration to be expected is a function of the external pressure, the interfacie1~tension of the liquid-solid interface, and the poreisize spectrum of the'gréphite. The penetration observed for a given.specimen should be independent of treatment time. Permeétion.by fused fluorides has been studied with a large. number of graphite’specimens which were degassed under conditions that could. be matched in the reactor and were subsequently exposed to LiF-BeF, -ThFu—UFh mixtures at” 1500 F and at 95 and 150 psig of argon pressure. Low-density and qulte—permeable graphltes such as AGOT are permeated to the extent of nearly l5°vol % at elther pressure. Of a total of 31 grades of graphite tésted, four grades (B-1, S-4LB, GT-123-82, and CS-112-8) show salt permeation of less than 0.5% of the bulk graphite volume under 150 psig, while four others (CT-150, CEY-1350, CT-158, and CEY-G) also show less than 0.5% permeation under ‘95 psig. No effect of treatment time was observed in these tests. . Compatibility in. Long-Term Forced-Flow Test In.an*efigineering téstof'the compatibility cf*graphitefiwith‘molten salts, 31 spécimens of a”speéiai'graphite (National Carbon Co. GT-123-82) ‘were exposed at 1300°F in & flowing stream of LiF- BeF »~UF), - (62 -37-1 mole %) for one year in a forced-c1rculat10n loop of. INOR 8. A flow .rate of 1.1 gpm was maintained ‘over the spec1mens with an - effectlve pumping . 120~ head of 10 péi.‘ An additional pressure of 3 psig was maintained by pressurizing the helium cover gas, so that the total pressure on the salt-graphite system was 13 psig. The graphite specimens (of two sizes: 11 in. long by 1/2 or 3/8 in. in diameter) were degasséd for 24 hr at 1100°F by evacuation of the test loop and were flooded with argon be- fore the loop was charged with salt. After the year of operation the.specimens were recovered for examination. The graphite was clearly not wetted by the fluoride melt; the specimens drained clean except for a few tiny spherical particles of salt which loosely adhered to the specimens. Dimensional changes. for the rods averaged less than 0.5 mil on the diameter; this figure 1is close to the probable error of the measurements. No weight gains were observed. Weight losses varied from negligible to 0.05% and averaged 0.02%; these losses may represent desorption of residual gases | from the samples or may, perhaps, be evidence of slight erosion. Analysis of the graphite for uranium indicated an average of 15 ppm. The graphite was in excellent condition; it is clear that exposure under these conditions is not deleterious to the system. In-Pile Testing ‘The results of an extensive program of out-of-pile testing are generally quite reassuring.. The in-pile testing has disclosed no evi- dence which contradicts the out-of-pile tests, but the in-pile tests have been too few to be reassuring. | | E Two graphite crucibles (each 1.5 in. long, 0.10 in. in inside dia- meter, with 0.025 in. wall thickness) of a high-density graphite from National Carbon Company (similar to GT—123-82) were irradiated ih the MIR at 1250°F for 1610 and 1492 hr, respectively, while charged with LiF-BeFEwUFu (62-57—1 mole %) mixture containing fully enriched U255. The crucibles were plugged top and bottom with caps of the same graphite material,afld were enclosed in containers of Inconel. Irradiations to integrated dosages of 1520 and 1375_kw-hr/cm§ were given to the cap- sules. Postirradiation sectioning and inspection of the specimens re- vealed no evidence of damage to the graphite nor any evidence that the X -121- graphite structure had been perméated by the salt. It may be concluded that no gross damage to the MSRE graphite will occur, at least during short-term irradiation. Two éttempts at a considerably more sophisticated'experiment*gave- only partial success. Graphite specimens, enclosed in flexible cap- sules fabricated from INOR-8 bellows and filled with LiF-BeF,-ThF -UF) mixture, were exposed for 1600 hr at 1300°F in the MIR. - The fléxible capsules were immersed in a bath of molten sodium which served as the heat removal sink and also transmitted a pressure of lOO psig to the graphlte -salt system. Only one of elght capsules so exposed;suryiyedf the numerous thermal cycles, with sudden freezing and thawing of the 'salt, imposed by the MIR operation. The ome surviving capsule con- tained a specimen of S-UA graphite. After this exposure, in which the power density in the fluoride fuel was about 200 w/cc, the external appearance-of the graphite was relatively-good_and the physical dimen- sions had changed relatively little. From the increase in weight it appeared that 0.71 vol % of the specimen was permeated with salt; this is to be compared with out-of-pile tests in which S-4A was permeated to 1.0% at 150 psia. | Subsequent sectioning of the specimen and a combination of auto- radiography, micro core drilling and subsequent analysis by counting techniques, -and metallographic examination showed relatively high con- " centrations of fuel near the external surface of the graphite and along an internal chord of the specimen. Except in these surface re- gions and along this band of apparenfly‘high-porosity’graphite within the specimen, the graphite interior is relétively clean; in‘:general, the fission species which can be identified in the interior are alkali metals presumably deposited from noble-gas precursors which permeated - the moderator. | . It would be unwise to conclude too much from this single. specimen- (which is far from the best available graphite), and more studies are clearly needed. It appears, however, that no marked differenceS’be-‘- tween the in-pile and out-of-pile behavior have yet been encountered. l. 2. ~122- REFERENCES W. B. Cottrell, ed., ORNL-140T (Nov. 24, 1952)(Secret). W. B. Cottrell et al., The Aircraft Reactor Test Hazards Summary Report, ORNL-1835 (Jan. 19, 1955). C. L. Davis, J. M. Bookston, and B. E. Smith, GNU IT - A Multigroup One-Dimensional Diffusion Program for the IBM-TOL, General Motors Report GMR-101 (1957). | Cc. W. Nestor, Jr., Multigroup Neutron Cross Sections ORNL CF-60-3- 35 and revision (March 1960). o L. G. Alexander et al., Preliminary Report on Thermal Breeder Reactor Evaluation, ORNL CF-60-T-1 (July 1960). G. G. Bilodeau et al., PDQ - An IBM-T0l Code to Solve the Two- Dimensional'Few-Group Neutron-Diffusion Equations, WAFD-TM~TO (1957). L. F. Parsly, MSRE Drain Tank - Heat Removal Studies, ORNL CF-60-9-55 (Sept. 1960) Sturm-Krouse, Inc., Analyses and Design Suggesticns for Freeze Flange Assembllies for MSRE, Nov. 30, 1960. S. E. Beall et al., HRP Quar. Prog. Rep. CGet. 31, 1958, ORNL-265Lk, . T. ' fi A Meteorological Survey of the Oak Ridge Area, OR0O-99. Met. Ann. Prog. Rep. Oct. 10, 1958, ORNL-2632, p 89-92 (Classified). Met. Ann. Prog. Rep. Sept. 1, 1959,’0RNL=2839, é 150-153. . Met. Amn. Prog. Rep. July 1, 1960, ORNL-2983, p 197-199. 14, MSR Quar. Prog. Rep. July 31, 1959, ORNL-2799, p 88. O @O~ O\ Wi 10. H. G. MacPherson ~123= EXTERNAL TRANSMITTAL AUTHORIZED ORNL CF-61-2-46 INTERNAL DISTRIBUTION Manly L. G. Alexander - W. D, S. E. Beall 37. E. R. Mann C. E. Bettis 38. 'W. B. 'McDonald E. S; Bettis 39. C. K. McGlothlan F. F. Bla.nkenship 40. R. L. Moore A. L. Boch . J, C. Moyers S. E. Bolt k2, C. W. Nestor ' W. L. Breazeale 3. T. E. Northup R. B. Briggs L4y, L. F, Parsly F. R. Bruce 45. H. R. Payne R. A. Charpie 46. R. E. Ramsey W. H. Cook 47. W. D. Reel G. A, Cristy 48, M. Richardson J. L. Crowley 49, R. T. Santoro F. L., Culler. 50. H. W. Savage E. P. Epler . 51. D. Scott W. K. Ergen 52. M. J. Skinner. W. H. Ford 53. A. N. Smith - A, P, Fraas 54, P. G. Smith C. H. Gabbard 55, I. Spiewak W. R. Gall 56. J. A. Swartout R. B. Gallaher 57. A. Taboadsa W. R.- Grimes 58. J. R. Tallackson . A. G. Grindell 59.. W. C. Ulrich P. N. Haubenreich 60. A. M. Weinberg J. W. Hill - 6l. J. H. Westsik E. C. Hise 62, L. V. Wilson P. P, Holz 63. C. 'E'o";Win't;erS L. N. Howell 64, C. H. Wodtke W. H. Jordan 65-66. R. D. Library P. R. Kasten 67-68. ORNL - Y-12 Technical Library, . R, J. Kedl - \ Document Reference Section B. W. Kinyon 69-94. Laboratory Records Department J. A. Lane ' 95. Laboratory Records, ORNL R.C. 96-98. Central Research Library T EXTERNAL DISTRIBUTTION 99-135. H. M. Roth, AEC, ORO